Turbopump options for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Bissell, W. R.; Gunn, S. V.

1992-07-01

Several turbopump options for delivering liquid nitrogen to nuclear thermal rocket (NTR) engines were evaluated and compared. Axial and centrifugal flow pumps were optimized, with and without boost pumps, utilizing current design criteria within the latest turbopump technology limits. Two possible NTR design points were used, a modest pump pressure rise of 1,743 psia and a relatively higher pump pressure rise of 4,480 psia. Both engines utilized the expander cycle to maximize engine performance for the long duration mission. Pump suction performance was evaluated. Turbopumps with conventional cavitating inducers were compared with zero NPSH (saturated liquid in the tanks) pumps over a range of tank saturation pressures, with and without boost pumps. Results indicate that zero NSPH pumps at high tank vapor pressures, 60 psia, are very similar to those with the finite NPSHs. At low vapor pressures efficiencies fall and turbine pressure ratios increase leading to decreased engine chamber pressures and or increased pump pressure discharges and attendant high-pressure component weights. It may be concluded that zero tank NSPH capabilities can be obtained with little penalty to the engine systems but boost pumps are needed if tank vapor pressure drops below 30 psia. Axial pumps have slight advantages in weight and chamber pressure capability while centrifugal pumps have a greater operating range.

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

DOE PAGES

Bartel, N.; Chen, M.; Utgikar, V. P.; ...

2015-04-04

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

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

Bartel, N.; Chen, M.; Utgikar, V. P.

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

Non-thermal atmospheric pressure plasma activates lactate in Ringer’s solution for anti-tumor effects

NASA Astrophysics Data System (ADS)

Tanaka, Hiromasa; Nakamura, Kae; Mizuno, Masaaki; Ishikawa, Kenji; Takeda, Keigo; Kajiyama, Hiroaki; Utsumi, Fumi; Kikkawa, Fumitaka; Hori, Masaru

2016-11-01

Non-thermal atmospheric pressure plasma is a novel approach for wound healing, blood coagulation, and cancer therapy. A recent discovery in the field of plasma medicine is that non-thermal atmospheric pressure plasma not only directly but also indirectly affects cells via plasma-treated liquids. This discovery has led to the use of non-thermal atmospheric pressure plasma as a novel chemotherapy. We refer to these plasma-treated liquids as plasma-activated liquids. We chose Ringer’s solutions to produce plasma-activated liquids for clinical applications. In vitro and in vivo experiments demonstrated that plasma-activated Ringer’s lactate solution has anti-tumor effects, but of the four components in Ringer’s lactate solution, only lactate exhibited anti-tumor effects through activation by non-thermal plasma. Nuclear magnetic resonance analyses indicate that plasma irradiation generates acetyl and pyruvic acid-like groups in Ringer’s lactate solution. Overall, these results suggest that plasma-activated Ringer’s lactate solution is promising for chemotherapy.

75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

...The Nuclear Regulatory Commission (NRC) is amending its regulations to provide alternate fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. This final rule provides alternate PTS requirements based on updated analysis methods. This action is desirable because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action reduces regulatory burden for those PWR licensees who expect to exceed the existing requirements before the expiration of their licenses, while maintaining adequate safety, and may choose to comply with the final rule as an alternative to complying with the existing requirements.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

Prospective scenarios of nuclear energy evolution over the 21. century

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

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

A Review of Large-Scale Fracture Experiments Relevant to Pressure Vessel Integrity Under Pressurized Thermal Shock Conditions

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

Pugh, C.E.

2001-01-29

Numerous large-scale fracture experiments have been performed over the past thirty years to advance fracture mechanics methodologies applicable to thick-wall pressure vessels. This report first identifies major factors important to nuclear reactor pressure vessel (RPV) integrity under pressurized thermal shock (PTS) conditions. It then covers 20 key experiments that have contributed to identifying fracture behavior of RPVs and to validating applicable assessment methodologies. The experiments are categorized according to four types of specimens: (1) cylindrical specimens, (2) pressurized vessels, (3) large plate specimens, and (4) thick beam specimens. These experiments were performed in laboratories in six different countries. This reportmore » serves as a summary of those experiments, and provides a guide to references for detailed information.« less

Application of nuclear pumped laser to an optical self-powered neutron detector

NASA Astrophysics Data System (ADS)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

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.

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

2007-01-01

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Multiphysics Thrust Chamber Modeling for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Cheng, Gary; Chen, Yen-Sen

2006-01-01

The objective of this effort is to develop an efficient and accurate thermo-fluid computational methodology to predict environments for a solid-core, nuclear thermal engine thrust chamber. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation. A two-pronged approach is employed in this effort: A detailed thermo-fluid analysis on a multi-channel flow element for mid-section corrosion investigation; and a global modeling of the thrust chamber to understand the effect of heat transfer on thrust performance. Preliminary results on both aspects are presented.

Horizontal steam generator thermal-hydraulics

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

Ubra, O.; Doubek, M.

1995-09-01

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

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

DOEpatents

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Layer, E.H. Jr.; Peet, C.S.

1962-01-23

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

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

Edwards, Tyler A.; Edwards, Michael J.

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

Performance assessment of low pressure nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Gerrish, H. P., Jr.; Doughty, G. E.

1993-01-01

A low pressure nuclear thermal propulsion (LPNTP) system, which takes advantage of hydrogen dissociation/recombination, was proposed as a means of increasing engine specific impulse (Isp). The effect of hydrogen dissociation/recombination on LPNTP Isp is examined. A two-dimensional computer model was used to show that the optimum chamber pressure is approximately 100 psia (at a chamber temperature of 3,000 K), with an Isp approximately 15 s higher than at 1,000 psia. At high chamber temperatures and low chamber pressures, the increase in Isp is due to both lower average molecular weights caused by dissociation and added kinetic energy from monatomic hydrogen recombination. Monatomic hydrogen recombination increases the Isp more then hydrogen dissociation. Variations in the mole fraction of monatomic hydrogen are similar to variations in static pressure along the axial nozzle position. Most recombination occurs close to the nozzle throat. Practical variations in nozzle geometry have minimal impact on recombination. Other models which can simulate a wider range of nozzle designs should be used in the future. The uncertainty of the hydrogen kinetic reaction rates at high temperatures (approximately 3,000 K) affects the accuracy of the analysis and should be verified with simple bench tests.

Nuclear pumped laser II

NASA Technical Reports Server (NTRS)

Deyoung, R. J.; Lee, J. H.; Pinkston, W. T.

1977-01-01

The first direct nuclear pumped laser using the He-2-(n,p) H-3 reaction is reported. Lasing took place on the 1.79 microns Ar I transition in a mixture of He-3-Ar at approximately 600 Torr total pressure. It was found that the electrically pulsed afterglow He-Ar laser had the same concentration profile as the nuclear pumped laser. As a result, nuclear lasing was also achieved in He-3-Xe (2.027 micron) and He-3-Kr (2.52 micron). Scaling of laser output with both thermal flux and total pressure as well as minority concentration has been completed. A peak output (He-3-Ar) of 3.7 watts has been achieved at a total pressure of 4 atm. Direct nuclear pumping of He-3-Ne has also been achieved. Nuclear pumping of a He-3-NF3 mixture was attempted, lasing in FI at approximately 7000 A, without success, although the potential lasing transitions appeared in spontaneous emission. Both NF3 and 238UF6 appear to quench spontaneous emission when they constitute more than 1% of the gas mixture.

Multiphysics Thermal-Fluid Design Analysis of a Non-Nuclear Tester for Hot-Hydrogen Materials and Component Development

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Foote, John; Litchford, Ron

2006-01-01

The objective of this effort is to perform design analyses for a non-nuclear hot-hydrogen materials tester, as a first step towards developing efficient and accurate multiphysics, thermo-fluid computational methodology to predict environments for hypothetical solid-core, nuclear thermal engine thrust chamber design and analysis. The computational methodology is based on a multidimensional, finite-volume, turbulent, chemically reacting, thermally radiating, unstructured-grid, and pressure-based formulation. The multiphysics invoked in this study include hydrogen dissociation kinetics and thermodynamics, turbulent flow, convective, and thermal radiative heat transfers. The goals of the design analyses are to maintain maximum hot-hydrogen jet impingement energy and to minimize chamber wall heating. The results of analyses on three test fixture configurations and the rationale for final selection are presented. The interrogation of physics revealed that reactions of hydrogen dissociation and recombination are highly correlated with local temperature and are necessary for accurate prediction of the hot-hydrogen jet temperature.

Performance assessment of low pressure nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harrold P., Jr.; Doughty, Glen E.

1993-01-01

An increase in Isp for nuclear thermal propulsion systems is desirable for reducing the propellant requirements and cost of future applications, such as the Mars Transfer Vehicle. Several previous design studies have suggested that the Isp could be increased substantially with hydrogen dissociation/recombination. Hydrogen molecules (H2), at high temperatures and low pressures, will dissociate to monatomic hydrogen (H). The reverse process (i.e., formation of H2 from H) is exothermic. The exothermic energy in a nozzle increases the kinetic energy and therefore, increases the Isp. The low pressure nuclear thermal propulsion system (LPNTP) system is expected to maximize the hydrogen dissociation/recombination and Isp by operating at high chamber temperatures and low chamber pressures. The process involves hydrogen flow through a high temperature, low pressure fission reactor, and out a nozzle. The high temperature (approximately 3000 K) of the hydrogen in the reactor is limited by the temperature limits of the reactor material. The minimum chamber pressure is about 1 atm because lower pressures decrease the engines thrust to weight ratio below acceptable limits. This study assumes that hydrogen leaves the reactor and enters the nozzle at the 3000 K equilibrium dissociation level. Hydrogen dissociation in the reactor does not affect LPNTP performance like dissociation in traditional chemical propulsion systems, because energy from the reactor resupplies energy lost due to hydrogen dissociation. Recombination takes place in the nozzle due primarily to a drop in temperature as the Mach number increases. However, as the Mach number increases beyond the nozzle throat, the static pressure and density of the flow decreases and minimizes the recombination. The ideal LPNTP Isp at 3000 K and 10 psia is 1160 seconds due to the added energy from fast recombination rates. The actual Isp depends on the finite kinetic reaction rates which affect the amount of monatomic hydrogen recombination before the flow exits the nozzle. A LPNTP system has other technical issues (e.g. flow instability and two-phase flow) besides hydrogen dissociation/recombination which affect the systems practicality. In this study, only the effects of hydrogen dissociation/recombination are examined.

Analysis of Material Sample Heated by Impinging Hot Hydrogen Jet in a Non-Nuclear Tester

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Foote, John; Litchford, Ron

2006-01-01

A computational conjugate heat transfer methodology was developed and anchored with data obtained from a hot-hydrogen jet heated, non-nuclear materials tester, as a first step towards developing an efficient and accurate multiphysics, thermo-fluid computational methodology to predict environments for hypothetical solid-core, nuclear thermal engine thrust chamber. The computational methodology is based on a multidimensional, finite-volume, turbulent, chemically reacting, thermally radiating, unstructured-grid, and pressure-based formulation. The multiphysics invoked in this study include hydrogen dissociation kinetics and thermodynamics, turbulent flow, convective and thermal radiative, and conjugate heat transfers. Predicted hot hydrogen jet and material surface temperatures were compared with those of measurement. Predicted solid temperatures were compared with those obtained with a standard heat transfer code. The interrogation of physics revealed that reactions of hydrogen dissociation and recombination are highly correlated with local temperature and are necessary for accurate prediction of the hot-hydrogen jet temperature.

Assessment of safety-relevant aspects of Kraftwerk Union's 200-MW(thermal) nuclear district heating plant concept

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

Erlenwein, P.; Frisch, W.; Kafka, P.

Nuclear reactors of 200- to 400-MW(thermal) power for district heating are the subject of increasing interest, and several specific designs are under discussion today. In the Federal Republic of Germany (FRG), the Kraftwerk Union AG has presented a 200-MW(thermal) heating reactor concept. The main safety issues of this design are assessed. In this design, the primary system is fully integrated into the reactor pressure vessel (RPV), which is tightly enclosed by the containment. The low process parameters like pressure, temperature, and power density and the high ratio of coolant volume to thermal power allow the design of simple safety features.more » This is supported by the preference of passive over active components. A special feature is a newly designed hydraulic control and rod drive mechanism, which is also integrated into the RPV. Within the safety assessment an overview of the relevant FRG safety rules and guidelines, developed mainly for large, electricity-generating power plants, is given. Included is a discussion of the extent to which these licensing rules can be applied to the concept of heating reactors.« less

Evaluation of coupling approaches for thermomechanical simulations

DOE PAGES

Novascone, S. R.; Spencer, B. W.; Hales, J. D.; ...

2015-08-10

Many problems of interest, particularly in the nuclear engineering field, involve coupling between the thermal and mechanical response of an engineered system. The strength of the two-way feedback between the thermal and mechanical solution fields can vary significantly depending on the problem. Contact problems exhibit a particularly high degree of two-way feedback between those fields. This paper describes and demonstrates the application of a flexible simulation environment that permits the solution of coupled physics problems using either a tightly coupled approach or a loosely coupled approach. In the tight coupling approach, Newton iterations include the coupling effects between all physics,more » while in the loosely coupled approach, the individual physics models are solved independently, and fixed-point iterations are performed until the coupled system is converged. These approaches are applied to simple demonstration problems and to realistic nuclear engineering applications. The demonstration problems consist of single and multi-domain thermomechanics with and without thermal and mechanical contact. Simulations of a reactor pressure vessel under pressurized thermal shock conditions and a simulation of light water reactor fuel are also presented. Here, problems that include thermal and mechanical contact, such as the contact between the fuel and cladding in the fuel simulation, exhibit much stronger two-way feedback between the thermal and mechanical solutions, and as a result, are better solved using a tight coupling strategy.« 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.

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

NASA Astrophysics Data System (ADS)

Porcheron, Emmanuel; Lemaitre, Pascal; Nuboer, Amandine; Vendel, Jacques

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surété Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Integrating CO₂ storage with geothermal resources for dispatchable renewable electricity

DOE PAGES

Buscheck, Thomas A.; Bielicki, Jeffrey M.; Chen, Mingjie; ...

2014-12-31

We present an approach that uses the huge fluid and thermal storage capacity of the subsurface, together with geologic CO₂ storage, to harvest, store, and dispatch energy from subsurface (geothermal) and surface (solar, nuclear, fossil) thermal resources, as well as energy from electrical grids. Captured CO₂ is injected into saline aquifers to store pressure, generate artesian flow of brine, and provide an additional working fluid for efficient heat extraction and power conversion. Concentric rings of injection and production wells are used to create a hydraulic divide to store pressure, CO₂, and thermal energy. Such storage can take excess power frommore » the grid and excess/waste thermal energy, and dispatch that energy when it is demanded, enabling increased penetration of variable renewables. Stored CO₂ functions as a cushion gas to provide enormous pressure-storage capacity and displaces large quantities of brine, which can be desalinated and/or treated for a variety of beneficial uses.« less

Digital computer program for nuclear reactor design water properties (LWBR Development Program)

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

Lynn, L.L.

1967-07-01

An edit program MO899 for the tabulation of thermodynamic and transport properties of liquid and vapor water, frequently used in design calculations for pressurized water nuclear reactors, is described. The data tabulated are obtained from a FORTRAN IV subroutine named HOH. Values of enthalpy, specific volume, viscosity, and thermal conductivity are given for the following ranges: pressure from one bar (14.5 psia) to 175 bars (2538 psia) and temperature from as much as 320 deg C (608 deg F) below saturation up to 500 deg C (932 deg F) above saturation. (NSA 21: 38472)

Liquid metal pump for nuclear reactors

DOEpatents

Allen, H.G.; Maloney, J.R.

1975-10-01

A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank.

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Targets for the production of radioisotopes and method of assembly

DOEpatents

Quinby, Thomas C.

1976-01-01

A target for preparation of radioisotopes by nuclear bombardment, and a method for its assembly are provided. A metallic sample to be bombarded is enclosed within a metallic support structure and the resulting target subjected to heat and pressure to effect diffusion bonds therebetween. The bonded target is capable of withstanding prolonged exposure to nuclear bombardment without thermal damage to the sample.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Analytical study of nozzle performance for nuclear thermal rockets

NASA Technical Reports Server (NTRS)

Davidian, Kenneth O.; Kacynski, Kenneth J.

1991-01-01

Nuclear propulsion has been identified as one of the key technologies needed for human exploration of the Moon and Mars. The Nuclear Thermal Rocket (NTR) uses a nuclear reactor to heat hydrogen to a high temperature followed by expansion through a conventional convergent-divergent nozzle. A parametric study of NTR nozzles was performed using the Rocket Engine Design Expert System (REDES) at the NASA Lewis Research Center. The REDES used the JANNAF standard rigorous methodology to determine nozzle performance over a range of chamber temperatures, chamber pressures, thrust levels, and different nozzle configurations. A design condition was set by fixing the propulsion system exit radius at five meters and throat radius was varied to achieve a target thrust level. An adiabatic wall was assumed for the nozzle, and its length was assumed to be 80 percent of a 15 degree cone. The results conclude that although the performance of the NTR, based on infinite reaction rates, looks promising at low chamber pressures, finite rate chemical reactions will cause the actual performance to be considerably lower. Parameters which have a major influence on the delivered specific impulse value include the chamber temperature and the chamber pressures in the high thrust domain. Other parameters, such as 2-D and boundary layer effects, kinetic rates, and number of nozzles, affect the deliverable performance of an NTR nozzle to a lesser degree. For a single nozzle, maximum performance of 930 seconds and 1030 seconds occur at chamber temperatures of 2700 and 3100 K, respectively.

Heat transfer deterioration in tubes caused by bulk flow acceleration due to thermal and frictional influences

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

Jackson, J. D.

2012-07-01

Severe deterioration of forced convection heat transfer can be encountered with compressible fluids flowing through strongly heated tubes of relatively small bore as the flow accelerates and turbulence is reduced because of the fluid density falling (as the temperature rises and the pressure falls due to thermal and frictional influence). The model presented here throws new light on how the dependence of density on both temperature and pressure can affect turbulence and heat transfer and it explains why the empirical equations currently available for calculating effectiveness of forced convection heat transfer under conditions of strong non-uniformity of fluid properties sometimesmore » fail to reproduce observed behaviour. It provides a criterion for establishing the conditions under which such deterioration of heat transfer might be encountered and enables heat transfer coefficients to be determined when such deterioration occurs. The analysis presented here is for a gaseous fluid at normal pressure subjected strong non-uniformity of fluid properties by the application of large temperature differences. Thus the model leads to equations which describe deterioration of heat transfer in terms of familiar parameters such as Mach number, Reynolds number and Prandtl number. It is applicable to thermal power plant systems such as rocket engines, gas turbines and high temperature gas-cooled nuclear reactors. However, the ideas involved apply equally well to fluids at supercritical pressure. Impairment of heat transfer under such conditions has become a matter of growing interest with the active consideration now being given to advanced water-cooled nuclear reactors designed to operate at pressures above the critical value. (authors)« less

A modernized high-pressure heater protection system for nuclear and thermal power stations

NASA Astrophysics Data System (ADS)

Svyatkin, F. A.; Trifonov, N. N.; Ukhanova, M. G.; Tren'kin, V. B.; Koltunov, V. A.; Borovkov, A. I.; Klyavin, O. I.

2013-09-01

Experience gained from operation of high-pressure heaters and their protection systems serving to exclude ingress of water into the turbine is analyzed. A formula for determining the time for which the high-pressure heater shell steam space is filled when a rupture of tubes in it occurs is analyzed, and conclusions regarding the high-pressure heater design most advisable from this point of view are drawn. A typical structure of protection from increase of water level in the shell of high-pressure heaters used in domestically produced turbines for thermal and nuclear power stations is described, and examples illustrating this structure are given. Shortcomings of components used in the existing protection systems that may lead to an accident at the power station are considered. A modernized protection system intended to exclude the above-mentioned shortcomings was developed at the NPO Central Boiler-Turbine Institute and ZioMAR Engineering Company, and the design solutions used in this system are described. A mathematical model of the protection system's main elements (the admission and check valves) has been developed with participation of specialists from the St. Petersburg Polytechnic University, and a numerical investigation of these elements is carried out. The design version of surge tanks developed by specialists of the Central Boiler-Turbine Institute for excluding false operation of the high-pressure heater protection system is proposed.

NERVA-Derived Nuclear Thermal Propulsion Dual Mode Operation

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Hundal, Rolv

1994-07-01

Generation of electrical power using the nuclear heat source of a NERVA-derived nuclear thermal rocket engine is presented. A 111,200 N thrust engine defined in a study for NASA-LeRC in FY92 is the reference engine for a three-engine vehicle for which a 50 kWe capacity is required. Processes are described for energy extraction from the reactor and for converting the energy to electricity. The tie tubes which support the reactor fuel elements are the source of thermal energy. The study focuses on process systems using Stirling cycle energy conversion operating at 980 K and an alternate potassium-Rankine system operating at 1,140 K. Considerations are given of the effect of the power production on turbopump operation, ZrH moderator dissociation, creep strain in the tie tubes, hydrogen permeation through the containment materials, requirements for a backup battery system, and the effects of potential design changes on reactor size and criticality. Nuclear considerations include changing tie tube materials to TZM, changing the moderator to low vapor-pressure yttrium hydride, and changing the fuel form from graphite matrix to a carbon-carbide composite.

Thermal and flow analysis of the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% effort of Title 1)

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

Steinke, R.G.; Mueller, C.; Knight, T.D.

1998-03-01

The computational fluid dynamics code CFX4.2 was used to evaluate steady-state thermal-hydraulic conditions in the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% of Title 1). Thirteen facility cases were evaluated with varying temperature dependence, drywell-array heat-source magnitude and distribution, location of the inlet tower, and no-flow curtains in the drywell-array vault. Four cases of a detailed model of the inlet-tower top fixture were evaluated to show the effect of the canopy-cruciform fixture design on the air pressure and flow distributions.

Thermally-driven Coupled THM Processes in Shales

NASA Astrophysics Data System (ADS)

Rutqvist, J.

2017-12-01

Temperature changes can trigger strongly coupled thermal-hydrological-mechanical (THM) processes in shales that are important to a number of subsurface energy applications, including geologic nuclear waste disposal and hydrocarbon extraction. These coupled processes include (1) direct pore-volume couplings, by thermal expansion of trapped pore-fluid that triggers instantaneous two-way couplings between pore fluid pressure and mechanical deformation, and (2) indirect couplings in terms of property changes, such as changes in mechanical stiffness, strength, and permeability. Direct pore-volume couplings have been studied in situ during borehole heating experiments in shale (or clay stone) formations at Mont Terri and Bure underground research laboratories in Switzerland and France. Typically, the temperature changes are accompanied with a rapid increase in pore pressure followed by a slower decrease towards initial (pre-heating) pore pressure. Coupled THM modeling of these heater tests shows that the pore pressure increases because the thermal expansion coefficient of the fluid is much higher than that of the porous clay stone. Such thermal pressurization induces fluid flow away from the pressurized area towards areas of lower pressure. The rate of pressure increase and magnitude of peak pressure depends on the rate of heating, pore-compressibility, and permeability of the shale. Modeling as well as laboratory experiments have shown that if the pore pressure increase is sufficiently large it could lead to fracturing of the shale or shear slip along pre-existing bedding planes. Another set of data and observations have been collected associated with studies related to concentrated heating and cooling of oil-shales and shale-gas formations. Heating may be used to enhance production from tight oil-shale, whereas thermal stimulation has been attempted for enhanced shale-gas extraction. Laboratory experiments on shale have shown that strength and elastic deformation modulus decreases with temperature while the rate creep deformations increase with temperature. Such temperature dependency also affects the well stability and zonal sealing across shale layers.

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Non-nuclear Testing of Reactor Systems in the Early Flight Fission Test Facilities (EFF-TF)

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2004-01-01

The Early Flight Fission-Test Facility (EFF-TF) can assist in the &sign and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are "non-nuclear" in nature (e.g. system s nuclear operations are understood). For many systems. thermal simulators can he used to closely mimic fission heat deposition. Axial power profile, radial power profile. and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other NASA centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004.

A Nuclear Ramjet Flyer for Exploration of Jovian Atmosphere

NASA Astrophysics Data System (ADS)

Maise, G.; Powell, J.; Paniagua, J.; Lecat, R.

2001-01-01

We investigated the design, operation, and data gathering possibilities of a nuclear-powered ramjet flyer in the Jovian atmosphere. The MITEE nuclear rocket engine can be modified to operate as a ramjet in planetary atmospheres. (Note: MITEE is a compact, ultra-light-weight thermal nuclear rocket which uses hydrogen as the propellant.) To operate as a ramjet, MITEE requires a suitable inlet and diffuser to substitute for the propellant that is pumped from the supply tanks in a nuclear rocket engine. Such a ramjet would fly in the upper Jovian atmosphere, mapping in detail temperatures, pressures, compositions, lightning activity, and wind speeds in the highly turbulent equatorial zone and the Great Red Spot. The nuclear ramjet could operate for months because: (1) the Jovian atmosphere has unlimited propellant, (2) the MITEE nuclear reactor is a (nearly) unlimited power source, and (3) with few moving parts, mechanical wear should be minimal. This paper presents a conceptual design of a ramjet flyer and its nuclear engine. The flyer incorporates a swept-wing design with instruments located in the twin wing-tip pods (away from the radiation source and readily shielded, if necessary). The vehicle is 2 m long with a 2 m wingspan. Its mass is 220 kg, and its nominal flight Mach number is 1.5. Based on combined neutronic and thermal/hydraulic analyses, we calculated that the ambient pressure range over which the flyer can operate to be from about 0.04 to 4 (terrestrial) atmospheres. This altitude range encompasses the three uppermost cloud layers in the Jovian atmosphere: (1) the entire uppermost visible NH3 ice cloud layer (where lightning has been observed), (2) the entire NH4HS ice cloud layer, and (3) the upper portion of the H2O ice cloud layer.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William

2013-01-01

A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

Discovery of massive star formation quenching by non-thermal effects in the centre of NGC 1097

NASA Astrophysics Data System (ADS)

Tabatabaei, F. S.; Minguez, P.; Prieto, M. A.; Fernández-Ontiveros, J. A.

2018-01-01

Observations show that massive star formation quenches first at the centres of galaxies. To understand quenching mechanisms, we investigate the thermal and non-thermal energy balance in the central kpc of NGC 1097—a prototypical galaxy undergoing quenching—and present a systematic study of the nuclear star formation efficiency and its dependencies. This region is dominated by the non-thermal pressure from the magnetic field, cosmic rays and turbulence. A comparison of the mass-to-magnetic flux ratio of the molecular clouds shows that most of them are magnetically critical or supported against the gravitational collapse needed to form the cores of massive stars. Moreover, the star formation efficiency of the clouds drops with the magnetic field strength. Such an anti-correlation holds with neither the turbulent nor the thermal pressure. Hence, a progressive build up of the magnetic field results in high-mass stars forming inefficiently, and this may be the cause of the low-mass stellar population in the bulges of galaxies.

Multiphysics Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Cheng, Gary; Chen, Yen-Sen

2006-01-01

The objective of this effort is to develop an efficient and accurate thermo-fluid computational methodology to predict environments for a hypothetical solid-core, nuclear thermal engine thrust chamber. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics methodology. Formulations for heat transfer in solids and porous media were implemented and anchored. A two-pronged approach was employed in this effort: A detailed thermo-fluid analysis on a multi-channel flow element for mid-section corrosion investigation; and a global modeling of the thrust chamber to understand the effect of hydrogen dissociation and recombination on heat transfer and thrust performance. The formulations and preliminary results on both aspects are presented.

Liquid metal hydrogen barriers

DOEpatents

Grover, George M.; Frank, Thurman G.; Keddy, Edward S.

1976-01-01

Hydrogen barriers which comprise liquid metals in which the solubility of hydrogen is low and which have good thermal conductivities at operating temperatures of interest. Such barriers are useful in nuclear fuel elements containing a metal hydride moderator which has a substantial hydrogen dissociation pressure at reactor operating temperatures.

Heated-Pressure-Ball Monopropellant Rocket Engine

NASA Technical Reports Server (NTRS)

Greene, William D.

2005-01-01

A recent technology disclosure presents a concept for a monopropellant thermal spacecraft thruster that would feature both the simplicity of a typical prior pressure-fed propellant supply system and the smaller mass and relative compactness of a typical prior pump-fed system. The source of heat for this thruster would likely be a nuclear- fission reactor. The propellant would be a cryogenic fluid (a liquefied low-molecular-weight gas) stored in a tank at a low pressure. The propellant would flow from the tank, through a feedline, into three thick-walled spherical tanks, denoted pressure balls, that would be thermally connected to the reactor. Valves upstream and downstream of the pressure balls would be operated in a three-phase cycle in which propellant would flow into one pressure ball while the fluid underwent pressurization through heating in another ball and pressurized propellant was discharged from the remaining ball into the reactor. After flowing through the reactor, wherein it would be further heated, the propellant would be discharged through an exhaust nozzle to generate thrust. A fraction of the pressurized gas from the pressure balls would be diverted to maintain the desired pressure in the tank.

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.

Heterogonous Nanofluids for Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Alammar, Khalid

2014-09-01

Nuclear reactions can be associated with high heat energy release. Extracting such energy efficiently requires the use of high-rate heat exchangers. Conventional heat transfer fluids, such as water and oils are limited in their thermal conductivity, and hence nanofluids have been introduced lately to overcome such limitation. By suspending metal nanoparticles with high thermal conductivity in conventional heat transfer fluids, thermal conductivity of the resulting homogeneous nanofluid is increased. Heterogeneous nanofluids offer yet more potential for heat transfer enhancement. By stratifying nanoparticles within the boundary layer, thermal conductivity is increased where temperature gradients are highest, thereby increasing overall heat transfer of a flowing fluid. In order to test the merit of this novel technique, a numerical study of a laminar pipe flow of a heterogeneous nanofluid was conducted. Effect of Iron-Oxide distribution on flow and heat transfer characteristics was investigated. With Iron-Oxide volume concentration of 0.009 in water, up to 50% local heat transfer enhancement was predicted for the heterogeneous compared to homogeneous nanofluids. Increasing the Reynolds number is shown to increase enhancement while having negligible effect on pressure drop. Using permanent magnets attached externally to the pipe, an experimental investigation conducted at MIT nuclear reactor laboratory for similar flow characteristics of a heterogeneous nanofluid have shown upto 160% enhancement in heat transfer. Such results show that heterogeneous nanofluids are promising for augmenting heat transfer rates in nuclear power heat exchanger systems.

Thermal decomposition hazard evaluation of hydroxylamine nitrate.

PubMed

Wei, Chunyang; Rogers, William J; Mannan, M Sam

2006-03-17

Hydroxylamine nitrate (HAN) is an important member of the hydroxylamine family and it is a liquid propellant when combined with alkylammonium nitrate fuel in an aqueous solution. Low concentrations of HAN are used primarily in the nuclear industry as a reductant in nuclear material processing and for decontamination of equipment. Also, HAN has been involved in several incidents because of its instability and autocatalytic decomposition behavior. This paper presents calorimetric measurement for the thermal decomposition of 24 mass% HAN/water. Gas phase enthalpy of formation of HAN is calculated using both semi-empirical methods with MOPAC and high-level quantum chemical methods of Gaussian 03. CHETAH is used to estimate the energy release potential of HAN. A Reactive System Screening Tool (RSST) and an Automatic Pressure Tracking Adiabatic Calorimeter (APTAC) are used to characterize thermal decomposition of HAN and to provide guidance about safe conditions for handling and storing of HAN.

Development of RF plasma simulations of in-reactor tests of small models of the nuclear light bulb fuel region

NASA Technical Reports Server (NTRS)

Roman, W. C.; Jaminet, J. F.

1972-01-01

Experiments were conducted to develop test configurations and technology necessary to simulate the thermal environment and fuel region expected to exist in in-reactor tests of small models of nuclear light bulb configurations. Particular emphasis was directed at rf plasma tests of approximately full-scale models of an in-reactor cell suitable for tests in Los Alamos Scientific Laboratory's Nuclear Furnace. The in-reactor tests will involve vortex-stabilized fissioning uranium plasmas of approximately 200-kW power, 500-atm pressure and equivalent black-body radiating temperatures between 3220 and 3510 K.

An Analysis of the Thermal Stability of Conventional and Alternative Aviation Fuels

NASA Astrophysics Data System (ADS)

Young, Neell

An experimental apparatus was used to examine the thermal stability of conventional and alternative aviation fuels. The apparatus is a simplified but controllable representation of an aircraft fuel system consisting of a preheating section and a test section. The preheating section simulates the fuel conditions as it acts as a coolant on board of the aircraft while the test section simulates the conditions of the fuel injection nozzles. The apparatus measures the accumulated deposit by taking the pressure drop data across the heated test section. After thermal stressing, the pressure drop data is verified by a carbon burnoff apparatus. The fuel chemical composition is evaluated by nuclear magnetic resonance spectroscopy. Experimental results are presented and discussed in this thesis for four different types of aviation fuels to show the relationship between fuel chemical composition and coking propensity. The experiments show that fuels with aromatic content tend to produce more deposits and the alternative fuels are potentially more thermally stable than their conventional counterparts.

Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study

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

Pareige, P.; Russell, K.F.; Stoller, R.E.

1998-03-01

Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentrationmore » in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.« less

Multiphysics Computational Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Cheng, Gary; Chen, Yen-Sen

2007-01-01

The objective of this effort is to develop an efficient and accurate computational heat transfer methodology to predict thermal, fluid, and hydrogen environments for a hypothetical solid-core, nuclear thermal engine - the Small Engine. In addition, the effects of power profile and hydrogen conversion on heat transfer efficiency and thrust performance were also investigated. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics platform, while formulations of conjugate heat transfer were implemented to describe the heat transfer from solid to hydrogen inside the solid-core reactor. The computational domain covers the entire thrust chamber so that the afore-mentioned heat transfer effects impact the thrust performance directly. The result shows that the computed core-exit gas temperature, specific impulse, and core pressure drop agree well with those of design data for the Small Engine. Finite-rate chemistry is very important in predicting the proper energy balance as naturally occurring hydrogen decomposition is endothermic. Locally strong hydrogen conversion associated with centralized power profile gives poor heat transfer efficiency and lower thrust performance. On the other hand, uniform hydrogen conversion associated with a more uniform radial power profile achieves higher heat transfer efficiency, and higher thrust performance.

Pressurizer tank upper support

DOEpatents

Baker, T.H.; Ott, H.L.

1994-01-11

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. 10 figures.

SASS-1--SUBASSEMBLY STRESS SURVEY CODE

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

Friedrich, C.M.

1960-01-01

SASS-1, an IBM-704 FORTRAN code, calculates pressure, thermal, and combined stresses in a nuclear reactor core subassembly. In addition to cross- section stresses, the code calculates axial shear stresses needed to keep plane cross sections plane under axial variations of temperature. The input and output nomenclature, arrangement, and formats are described. (B.O.G.)

Effects of Beryllium and Compaction Pressure on the Thermal Diffusivity of Uranium Dioxide Fuel Pellets

NASA Astrophysics Data System (ADS)

Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Ferreira, R. A. N.

2017-09-01

In nuclear reactors, the performance of uranium dioxide (UO2) fuel is strongly dependent on the thermal conductivity, which directly affects the fuel pellet temperature, the fission gas release and the fuel rod mechanical behavior during reactor operation. The use of additives to improve UO2 fuel performance has been investigated, and beryllium oxide (BeO) appears as a suitable additive because of its high thermal conductivity and excellent chemical compatibility with UO2. In this paper, UO2-BeO pellets were manufactured by mechanical mixing, pressing and sintering processes varying the BeO contents and compaction pressures. Pellets with BeO contents of 2 wt%, 3 wt%, 5 wt% and 7 wt% BeO were pressed at 400 MPa, 500 MPa and 600 MPa. The laser flash method was applied to determine the thermal diffusivity, and the results showed that the thermal diffusivity tends to increase with BeO content. Comparing thermal diffusivity results of UO2 with UO2-BeO pellets, it was observed that there was an increase in thermal diffusivity of at least 18 % for the UO2-2 wt% BeO pellet pressed at 400 MPa. The maximum relative expanded uncertainty (coverage factor k = 2) of the thermal diffusivity measurements was estimated to be 9 %.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients

NASA Astrophysics Data System (ADS)

Gallais, L.; Burla, R.; Martin, F.; Richaud, J. C.; Volle, G.; Pontillon, M.; Capdevila, H.; Pontillon, Y.

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients.

PubMed

Gallais, L; Burla, R; Martin, F; Richaud, J C; Volle, G; Pontillon, M; Capdevila, H; Pontillon, Y

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

NASA Astrophysics Data System (ADS)

Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.

2017-12-01

The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

3D J-Integral Capability in Grizzly

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

Spencer, Benjamin; Backman, Marie; Chakraborty, Pritam

2014-09-01

This report summarizes work done to develop a capability to evaluate fracture contour J-Integrals in 3D in the Grizzly code. In the current fiscal year, a previously-developed 2D implementation of a J-Integral evaluation capability has been extended to work in 3D, and to include terms due both to mechanically-induced strains and due to gradients in thermal strains. This capability has been verified against a benchmark solution on a model of a curved crack front in 3D. The thermal term in this integral has been verified against a benchmark problem with a thermal gradient. These developments are part of a largermore » effort to develop Grizzly as a tool that can be used to predict the evolution of aging processes in nuclear power plant systems, structures, and components, and assess their capacity after being subjected to those aging processes. The capabilities described here have been developed to enable evaluations of Mode- stress intensity factors on axis-aligned flaws in reactor pressure vessels. These can be compared with the fracture toughness of the material to determine whether a pre-existing flaw would begin to propagate during a pos- tulated pressurized thermal shock accident. This report includes a demonstration calculation to show how Grizzly is used to perform a deterministic assessment of such a flaw propagation in a degraded reactor pressure vessel under pressurized thermal shock conditions. The stress intensity is calculated from J, and the toughness is computed using the fracture master curve and the degraded ductile to brittle transition temperature.« less

Thermal modeling of a vertical dry storage cask for used nuclear fuel

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

Li, Jie; Liu, Yung Y.

2016-05-01

Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 x 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. Themore » effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask.« less

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

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

Durbin, Samuel G.; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask, in part by increasing the efficiency of internal conduction pathways, and also by increasing the internalmore » convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above- and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.« less

Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

NASA Astrophysics Data System (ADS)

Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Matsukawa, Y.; Nishiyama, Y.; Onizawa, K.

2012-06-01

The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the δ-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

High Temperature Nanocomposites For Nuclear Thermal Propulsion and In-Space Fabrication by Hyperbaric Pressure Laser Chemical Vapor Deposition

NASA Astrophysics Data System (ADS)

Maxwell, J. L.; Webb, N. D.; Espinoza, M.; Cook, S.; Houts, M.; Kim, T.

Nuclear Thermal Propulsion (NTP) is an indispensable technology for the manned exploration of the solar system. By using Hyperbaric Pressure Laser Chemical Vapor Deposition (HP-LCVD), the authors propose to design and build a promising next-generation fuel element composed of uranium carbide UC embedded in a latticed matrix of highly refractory Ta4HfC5 for an NTP rocket capable of sustaining temperatures up to 4000 K, enabling an Isp of up to 1250 s. Furthermore, HP-LCVD technology can also be harnessed to enable 3D rapid prototyping of a variety of materials including metals, ceramics and composites, opening up the possibility of in-space fabrication of components, replacement parts, difficult-to-launch solar sails and panels and a variety of other space structures. Additionally, rapid prototyping with HP-LCVD makes a feasible "live off the land" strategy of interplanetary and interstellar exploration ­ the precursors commonly used in the technology are found, often in abundance, on other solar system bodies either as readily harvestable gas (e.g. methane) or as a raw material that could be converted into a suitable precursor (e.g. iron oxide into ferrocene on Mars).

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

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

Design and Analysis of Boiler Pressure Vessels based on IBR codes

NASA Astrophysics Data System (ADS)

Balakrishnan, B.; Kanimozhi, B.

2017-05-01

Pressure vessels components are widely used in the thermal and nuclear power plants for generating steam using the philosophy of heat transfer. In Thermal power plant, Coal is burnt inside the boiler furnace for generating the heat. The amount of heat produced through the combustion of pulverized coal is used in changing the phase transfer (i.e. Water into Super-Heated Steam) in the Pressure Parts Component. Pressure vessels are designed as per the Standards and Codes of the country, where the boiler is to be installed. One of the Standards followed in designing Pressure Parts is ASME (American Society of Mechanical Engineers). The mandatory requirements of ASME code must be satisfied by the manufacturer. In our project case, A Shell/pipe which has been manufactured using ASME code has an issue during the drilling of hole. The Actual Size of the drilled holes must be, as per the drawing, but due to error, the size has been differentiate from approved design calculation (i.e. the diameter size has been exceeded). In order to rectify this error, we have included an additional reinforcement pad to the drilled and modified the design of header in accordance with the code requirements.

The Air Blast Wave from a Nuclear Explosion

NASA Astrophysics Data System (ADS)

Reines, Frederick

The sudden, large scale release of energy in the explosion of a nuclear bomb in air gives rise, in addition to nuclear emanations such as neutrons and gamma rays, to an extremely hot, rapidly expanding mass of air.** The rapidly expanding air mass has an initial temperature in the vicinity of a few hundred thousand degrees and for this reason it glows in its early stages with an intensity of many suns. It is important that the energy density in this initial "ball of fire" is of the order of 3 × 103 times that found in a detonating piece of TNT and hence that the initial stages of the large scale air motion produced by a nuclear explosion has no counterpart in an ordinary. H. E. explosion. Further, the relatively low temperatures ˜2,000°C associated with the initial stages of an H. E. detonation implies that the thermal radiation which it emits is a relatively insignificant fraction of the total energy involves. This point is made more striking when it is remembered that the thermal energy emitted by a hot object varies directly with the temperature in the Rayleigh Jeans region appropriate to the present discussion. The expansion of the air mass heated by the nuclear reaction produces, in qualitatively the same manner as in an H.E. explosion or the bursting of a high pressure balloon, an intense sharp pressure pulse, a shock wave, in the atmosphere. As the pressure pulse spreads outward it weakens due to the combined effects of divergence and the thermodynamically irreversible nature of the shock wave. The air comprising such a pressure pulse or blast wave moves first radially outward and then back towards the center as the blast wave passes. Since a permanent outward displacement of an infinite mass of air would require unlimited energy, the net outward displacement of the air distant from an explosion must approach zero with increasing distance. As the distance from the explosion is diminished the net outward displacement due to irreversible shock heating of the air increases and in the limit of small distances and increasingly strong shocks the net outward displacement of the shocked air is equal to the maximum outward displacement. These statements are applicable for short times of the order of seconds following the explosion since the heated air l behind by the shock wave will rise. The pressures and air mass motions associated with the rise of the atomic cloud are relatively unimportant in the free air pressure ranges from 2-15 psi for bomb yields under 100 kilotons (KT)…

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements

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

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent tomore » the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.« less

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

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

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

Multifluid geo-energy systems: Using geologic CO 2 storage for geothermal energy production and grid-scale energy storage in sedimentary basins

DOE PAGES

Buscheck, Thomas A.; Bielicki, Jeffrey M.; Edmunds, Thomas A.; ...

2016-05-05

We present an approach that uses the huge fluid and thermal storage capacity of the subsurface, together with geologic carbon dioxide (CO 2) storage, to harvest, store, and dispatch energy from subsurface (geothermal) and surface (solar, nuclear, fossil) thermal resources, as well as excess energy on electric grids. Captured CO 2 is injected into saline aquifers to store pressure, generate artesian flow of brine, and provide a supplemental working fluid for efficient heat extraction and power conversion. Concentric rings of injection and production wells create a hydraulic mound to store pressure, CO 2, and thermal energy. This energy storage canmore » take excess power from the grid and excess/waste thermal energy, and dispatch that energy when it is demanded and thus enable higher penetration of variable renewable energy technologies (e.g., wind, solar). CO 2 stored in the subsurface functions as a cushion gas to provide enormous pressure-storage capacity and displace large quantities of brine, some of which can be treated for a variety of beneficial uses. Geothermal power and energy-storage applications may generate enough revenues to compensate for CO 2 capture costs. While our approach can use nitrogen (N 2), in addition to CO 2, as a supplemental fluid, and store thermal energy, this study focuses using CO 2 for geothermal energy production and grid-scale energy storage. We conduct a techno-economic assessment to determine the levelized cost of electricity of using this approach to generate geothermal power. We present a reservoir pressure-management strategy that diverts a small portion of the produced brine for beneficial consumptive use to reduce the pumping cost of fluid recirculation, while reducing the risk of seismicity, caprock fracture, and CO 2 leakage.« less

Thermal expansion in dispersion-bound molecular crystals

NASA Astrophysics Data System (ADS)

Ko, Hsin-Yu; DiStasio, Robert A.; Santra, Biswajit; Car, Roberto

2018-05-01

We explore how anharmonicity, nuclear quantum effects (NQE), many-body dispersion interactions, and Pauli repulsion influence thermal properties of dispersion-bound molecular crystals. Accounting for anharmonicity with ab initio molecular dynamics yields cell parameters accurate to within 2 % of experiment for a set of pyridinelike molecular crystals at finite temperatures and pressures. From the experimental thermal expansion curve, we find that pyridine-I has a Debye temperature just above its melting point, indicating sizable NQE across the entire crystalline range of stability. We find that NQE lead to a substantial volume increase in pyridine-I (≈40 % more than classical thermal expansion at 153 K) and attribute this to intermolecular Pauli repulsion promoted by intramolecular quantum fluctuations. When predicting delicate properties such as the thermal expansivity, we show that many-body dispersion interactions and more sophisticated density functional approximations improve the accuracy of the theoretical model.

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

Winterrose, M.; Lucas, M; Yue, A

Synchrotron x-ray diffraction (XRD) measurements, nuclear forward scattering (NFS) measurements, and density functional theory (DFT) calculations were performed on L12-ordered Pd3Fe. Measurements were performed at 300 K at pressures up to 33 GPa, and at 7 GPa at temperatures up to 650 K. The NFS revealed a collapse of the 57Fe magnetic moment between 8.9 and 12.3 GPa at 300 K, coinciding with a transition in bulk modulus found by XRD. Heating the sample under a pressure of 7 GPa showed negligible thermal expansion from 300 to 523 K, demonstrating Invar behavior. Zero-temperature DFT calculations identified a ferromagnetic ground statemore » and showed several antiferromagnetic states had comparable energies at pressures above 20 GPa.« less

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

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

Durbin, Samuel G.; Lindgren, Eric Richard

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing themore » internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.« less

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Qualification tests for {sup 192}Ir sealed sources

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

Iancso, Georgeta, E-mail: georgetaiancso@yahoo.com; Iliescu, Elena, E-mail: georgetaiancso@yahoo.com; Iancu, Rodica, E-mail: georgetaiancso@yahoo.com

This paper describes the results of qualification tests for {sup 192}Ir sealed sources, available in Testing and Nuclear Expertise Laboratory of National Institute for Physics and Nuclear Engineering 'Horia Hulubei' (I.F.I.N.-HH), Romania. These sources had to be produced in I.F.I.N.-HH and were tested in order to obtain the authorization from The National Commission for Nuclear Activities Control (CNCAN). The sources are used for gammagraphy procedures or in gammadefectoscopy equipments. Tests, measurement methods and equipments used, comply with CNCAN, AIEA and International Quality Standards and regulations. The qualification tests are: 1. Radiological tests and measurements: dose equivalent rate at 1 m;more » tightness; dose equivalent rate at the surface of the transport and storage container; external unfixed contamination of the container surface. 2. Mechanical and climatic tests: thermal shock; external pressure; mechanic shock; vibrations; boring; thermal conditions for storage and transportation. Passing all tests, it was obtained the Radiological Security Authorization for producing the {sup 192}Ir sealed sources. Now IFIN-HH can meet many demands for this sealed sources, as the only manufacturer in Romania.« less

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

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

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

Transport properties of LiF under strong compression: modeling using advanced electronic structure methods and classical molecular dynamics

NASA Astrophysics Data System (ADS)

Mattsson, Thomas R.; Jones, Reese; Ward, Donald; Spataru, Catalin; Shulenburger, Luke; Benedict, Lorin X.

2015-06-01

Window materials are ubiquitous in shock physics and with high energy density drivers capable of reaching multi-Mbar pressures the use of LiF is increasing. Velocimetry and temperature measurements of a sample through a window are both influenced by the assumed index of refraction and thermal conductivity, respectively. We report on calculations of index of refraction using the many-body theory GW and thermal ionic conductivity using linear response theory and model potentials. The results are expected to increase the accuracy of a broad range of high-pressure shock- and ramp compression experiments. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Company, for the U.S. Department of Energy's National Nuclear Security Administration under Contract DE-AC04-94AL85000.

Thermal Properties of Consolidated Granular Salt as a Backfill Material

NASA Astrophysics Data System (ADS)

Paneru, Laxmi P.; Bauer, Stephen J.; Stormont, John C.

2018-03-01

Granular salt has been proposed as backfill material in drifts and shafts of a nuclear waste disposal facility where it will serve to conduct heat away from the waste to the host rock. Creep closure of excavations in rock salt will consolidate (reduce the porosity of) the granular salt. This study involved measuring the thermal conductivity and specific heat of granular salt as a function of porosity and temperature to aid in understanding how thermal properties will change during granular salt consolidation accomplished at pressures and temperatures consistent with a nuclear waste disposal facility. Thermal properties of samples from laboratory-consolidated granular salt and in situ consolidated granular salt were measured using a transient plane source method at temperatures ranging from 50 to 250 °C. Additional measurements were taken on a single crystal of halite and dilated polycrystalline rock salt. Thermal conductivity of granular salt decreased with increases in temperature and porosity. Specific heat of granular salt at lower temperatures decreased with increasing porosity. At higher temperatures, porosity dependence was not apparent. The thermal conductivity and specific heat data were fit to empirical models and compared with results presented in the literature. At comparable densities, the thermal conductivities of granular salt samples consolidated hydrostatically in this study were greater than those measured previously on samples formed by quasi-static pressing. Petrographic studies of the consolidated salt indicate that the consolidation method influenced the nature of the porosity; these observations are used to explain the variation of measured thermal conductivities between the two consolidation methods. Thermal conductivity of dilated polycrystalline salt was lower than consolidated salt at comparable porosities. The pervasive crack network along grain boundaries in dilated salt impedes heat flow and results in a lower thermal conductivity compared to hydrostatically consolidated salt.

Structure-property relations of orthorhombic [(CH3)3NCH2COO]2(CuCl2)3 · 2H2 O

NASA Astrophysics Data System (ADS)

Haussühl, Eiken; Schreuer, Jürgen; Wiehl, Leonore; Paulsen, Natalia

2014-04-01

Large single crystals of orthorhombic [(CH3)3NCH2COO]2(CuCl2)3 · 2H2 O with dimensions up to 40×40×30 mm3 were grown from aqueous solutions. The elastic and piezoelastic coefficients were derived from ultrasonic resonance frequencies and their shifts upon variation of pressure, respectively, using the plate-resonance technique. Additionally, the coefficients of thermal expansion were determined between 95 K and 305 K by dilatometry. The elastic behaviour at ambient conditions is dominated by the 2-dimensional network of strong hydrogen bonds within the (001) plane leading to a corresponding pseudo-tetragonal anisotropy of the longitudinal elastic stiffness. The variation of elastic properties with pressure, however, as well as the thermal expansion shows strong deviations from the pseudo-tetragonal symmetry. These deviations are probably correlated with tilts of the elongated tri-nuclear betaine-CuCl2-water complexes. Neither the thermal expansion nor the specific heat capacity gives any hint on a phase transition in the investigated temperature range.

Tube failures in moisture separator-reheater tube bundles due to restrained thermal expansion

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

Heilker, W.J.; Cassell, D.S.

1983-01-01

In a nuclear power plant, moisture separator-reheater components (MSRs) are used to dry and superheat the exhaust steam from the high pressure turbine before admitting this steam to the low pressure turbines. MSRs have experienced numerous problems which have caused loss of plant thermal efficiency, poor unit availability and high maintenance costs. The most serious problem has been the progressive failure of the U-tubes, which has necessitated replacement of MSR tube bundles at several plants. This paper presents an explanation of the failure mode and identifies critical operational and geometric parameters as to their respective roles in the process. Detailedmore » thermal-hydraulic analytic modeling enables the calculation of tube wall temperatures along the length of each tube for selected power levels. These temperature data are input to finite element models of the tube bundle which yield interactive displacements, rotations and stresses. The results of these studies provide the rational basis for the tube failure mechanism, which is supported by data acquired from inspection of in-service MSRs.« less

PWR PRELIMINARY DESIGN FOR PL-3

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

Humphries, G. E.

1962-02-28

The pressurized water reactor preliminary design, the preferred design developed under Phase I of the PL-3 contract, is presented. Plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentatlon description, shielding description, auxiliary systems, power plant equipment, waste dispusal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information, and a training program outline are given. (auth)

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

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

farahani, A.A.; Corradini, M.L.

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for themore » foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.« less

Low Thrust, Deep Throttling, US/CIS Integrated NTRE

NASA Astrophysics Data System (ADS)

Culver, Donald W.; Kolganov, Vyacheslav; Rochow, Richard F.

1994-07-01

In 1993 our international team performed a follow-on ``Nuclear Thermal Rocket Engine (NTRE) Extended Life Feasibility Assessment'' study for the Nuclear Propulsion Office (NPO) at NASAs Lewis Research Center. The main purpose of this study was to complete the 1992 study matrix to assess NTRE designs at thrust levels of 22.5, 11.3, and 6.8 tonnes, using Commonwealth of Independent States (CIS) reactor technology. An additional Aerojet goal was to continue improving the NTRE concept we had generated. Deep throttling, mission performance optimized engine design parametrics, and reliability/cost enhancing engine system simplifications were studied, because they seem to be the last three basic design improvements sorely needed by post-NERVA NTRE. Deep throttling improves engine life by eliminating damaging thermal and mechanical shocks caused by after-cooling with pulsed coolant flow. Alternately, it improves mission performance with steady flow after-cooling by minimizing reactor over-cooling. Deep throttling also provides a practical transition from high pressures and powers of the high thrust power cycle to the low pressures and powers of our electric power generating mode. Two deep throttling designs are discussed; a workable system that was studied and a simplified system that is recommended for future study. Mission-optimized engine thrust/weight (T/W) and Isp predictions are included along with system flow schemes and concept sketches.

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

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

Williams, P. T.; Dickson, T. L.; Yin, S.

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include themore » NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.« less

Status update of the BWR cask simulator

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

Lindgren, Eric R.; Durbin, Samuel G.

2015-09-01

The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximummore » thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on heat load and the effect of simulated wind on a simplified below ground vent configuration.« less

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

Winterrose, M.L.; Lucas, M.S.; Yue, A.F.

Synchrotron x-ray diffraction (XRD) measurements, nuclear forward scattering (NFS) measurements, and density functional theory (DFT) calculations were performed on L1{sub 2}-ordered Pd{sub 3}Fe. Measurements were performed at 300 K at pressures up to 33 GPa, and at 7 GPa at temperatures up to 650 K. The NFS revealed a collapse of the Fe57 magnetic moment between 8.9 and 12.3 GPa at 300 K, coinciding with a transition in bulk modulus found by XRD. Heating the sample under a pressure of 7 GPa showed negligible thermal expansion from 300 to 523 K, demonstrating Invar behavior. Zero-temperature DFT calculations identified a ferromagneticmore » ground state and showed several antiferromagnetic states had comparable energies at pressures above 20 GPa.« less

Status Report and Research Plan for Cables Harvested from Crystal River Unit 3 Nuclear Generating Plant

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

Fifield, Leonard S.

Harvested cables from operating or decommissioned nuclear power plants present an important opportunity to validate models, understanding material aging behavior, and validate characterization techniques. Crystal River Unit 3 Nuclear Generating Plant is a pressurized water reactor that was licensed to operate from 1976 to 2013. Cable segments were harvested and made available to the Light Water Reactor Sustainability research program through the Electric Power Research Institute. Information on the locations and circuits within the reactor from whence the cable segments came, cable construction, sourcing and installation information, and photographs of the cable locations prior to harvesting were provided. The cablemore » variations provided represent six of the ten most common cable insulations in the nuclear industry and experienced service usage for periods from 15 to 42 years. Subsequently, these cables constitute a valuable asset for research to understand aging behavior and measurement of nuclear cables. Received cables harvested from Crystal River Unit 3 Nuclear Generating Plant consist of low voltage, insulated conductor surrounded by jackets in lengths from 24 to 100 feet each. Cable materials will primarily be used to investigate aging under simultaneous thermal and gamma radiation exposure. Each cable insulation and jacket material will be characterized in its as-received condition, including determination of the temperatures associated with endothermic transitions in the material using differential scanning calorimetry and dynamic mechanical analysis. Temperatures for additional thermal exposure aging will be selected following the thermal analysis to avoid transitions in accelerated laboratory aging that do not occur in field conditions. Aging temperatures above thermal transitions may also be targeted to investigate the potential for artifacts in lifetime prediction from rapid accelerated aging. Total gamma doses and dose rates targeted for each material will be determined based on filling gaps in prior work, known limits of material classes and resource constraints. Experimental plans will be developed in the context of existing data for the insulation and jacket materials available in published Department of Energy and Electric Power Research Institute reports toward addressing identified knowledge gaps.« less

Thermal expansion in dispersion-bound molecular crystals

DOE PAGES

Ko, Hsin -Yu; DiStasio, Robert A.; Santra, Biswajit; ...

2018-05-18

In this paper, we explore how anharmonicity, nuclear quantum effects (NQE), many-body dispersion interactions, and Pauli repulsion influence thermal properties of dispersion-bound molecular crystals. Accounting for anharmonicity with ab initio molecular dynamics yields cell parameters accurate to within 2% of experiment for a set of pyridinelike molecular crystals at finite temperatures and pressures. From the experimental thermal expansion curve, we find that pyridine-I has a Debye temperature just above its melting point, indicating sizable NQE across the entire crystalline range of stability. We find that NQE lead to a substantial volume increase in pyridine-I (≈ 40% more than classical thermalmore » expansion at 153 K) and attribute this to intermolecular Pauli repulsion promoted by intramolecular quantum fluctuations. Finally, when predicting delicate properties such as the thermal expansivity, we show that many-body dispersion interactions and more sophisticated density functional approximations improve the accuracy of the theoretical model.« less

Thermal expansion in dispersion-bound molecular crystals

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

Ko, Hsin -Yu; DiStasio, Robert A.; Santra, Biswajit

In this paper, we explore how anharmonicity, nuclear quantum effects (NQE), many-body dispersion interactions, and Pauli repulsion influence thermal properties of dispersion-bound molecular crystals. Accounting for anharmonicity with ab initio molecular dynamics yields cell parameters accurate to within 2% of experiment for a set of pyridinelike molecular crystals at finite temperatures and pressures. From the experimental thermal expansion curve, we find that pyridine-I has a Debye temperature just above its melting point, indicating sizable NQE across the entire crystalline range of stability. We find that NQE lead to a substantial volume increase in pyridine-I (≈ 40% more than classical thermalmore » expansion at 153 K) and attribute this to intermolecular Pauli repulsion promoted by intramolecular quantum fluctuations. Finally, when predicting delicate properties such as the thermal expansivity, we show that many-body dispersion interactions and more sophisticated density functional approximations improve the accuracy of the theoretical model.« less

Combined Pressure and Thermal Window System for Space Vehicles

NASA Technical Reports Server (NTRS)

Svartstrom, Kirk Nils (Inventor)

2015-01-01

A window system for a vehicle comprising a pressure and thermal window pane, a seal system, and a retainer system. The pressure and thermal window pane may be configured to provide desired pressure protection and desired thermal protection when exposed to an environment around the vehicle during operation of the vehicle. The pressure and thermal window pane may have a desired ductility. The seal system may be configured to contact the pressure and thermal window pane to seal the pressure and thermal window pane. The retainer system may be configured to hold the seal system and the pressure and thermal window pane.

Combined Pressure and Thermal Window System for Space Vehicles

NASA Technical Reports Server (NTRS)

Svartstrom, Kirk Nils (Inventor)

2017-01-01

A window system for a vehicle comprising a pressure and thermal window pane, a seal system, and a retainer system. The pressure and thermal window pane may be configured to provide desired pressure protection and desired thermal protection when exposed to an environment around the vehicle during operation of the vehicle. The pressure and thermal window pane may have a desired ductility. The seal system may be configured to contact the pressure and thermal window pane to seal the pressure and thermal window pane. The retainer system may be configured to hold the seal system and the pressure and thermal window pane.

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

NASA Astrophysics Data System (ADS)

Misenheimer, Corey Thomas

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

Portable double-sided pulsed laser heating system for time-resolved geoscience and materials science applications.

PubMed

Aprilis, G; Strohm, C; Kupenko, I; Linhardt, S; Laskin, A; Vasiukov, D M; Cerantola, V; Koemets, E G; McCammon, C; Kurnosov, A; Chumakov, A I; Rüffer, R; Dubrovinskaia, N; Dubrovinsky, L

2017-08-01

A portable double-sided pulsed laser heating system for diamond anvil cells has been developed that is able to stably produce laser pulses as short as a few microseconds with repetition frequencies up to 100 kHz. In situ temperature determination is possible by collecting and fitting the thermal radiation spectrum for a specific wavelength range (particularly, between 650 nm and 850 nm) to the Planck radiation function. Surface temperature information can also be time-resolved by using a gated detector that is synchronized with the laser pulse modulation and space-resolved with the implementation of a multi-point thermal radiation collection technique. The system can be easily coupled with equipment at synchrotron facilities, particularly for nuclear resonance spectroscopy experiments. Examples of applications include investigations of high-pressure high-temperature behavior of iron oxides, both in house and at the European Synchrotron Radiation Facility using the synchrotron Mössbauer source and nuclear inelastic scattering.

Thermohydrological conditions and silica redistribution near high-level nuclear wastes emplaced in saturated geological formations

NASA Astrophysics Data System (ADS)

Verma, A.; Pruess, K.

1988-02-01

Evaluation of the thermohydrological conditions near high-level nuclear waste packages is needed for the design of the waste canister and for overall repository design and performance assessment. Most available studies in this area have assumed that the hydrologic properties of the host rock are not changed in response to the thermal, mechanical, or chemical effects caused by waste emplacement. However, the ramifications of this simplifying assumption have not been substantiated. We have studied dissolution and precipitation of silica in liquid-saturated hydrothermal flow systems, including changes in formation porosity and permeability. Using numerical simulation, we compare predictions of thermohydrological conditions with and without inclusion of silica redistribution effects. Two cases were studied, namely, a canister-scale problem, and a repository-wide thermal convection problem and different pore models were employed for the permeable medium (fractures with uniform or nonuniform cross sections). We find that silica redistribution in water-saturated conditions does not have a sizeable effect on host rock and canister temperatures, pore pressures, or flow velocities.

Effects of CO2 injection and Kerogen Maturation on Low-Field Nuclear Magnetic Resonance Response

NASA Astrophysics Data System (ADS)

Prasad, M.; Livo, K.

2017-12-01

Low-field Nuclear Magnetic Resonance (NMR) is commonly used in petrophysical analysis of petroleum reservoir rocks. NMR experiments record the relaxation and polarization of in-situ hydrogen protons present in gaseous phases such as free-gas intervals and solution gas fluids, bulk fluid phases such as oil and aquifer intervals, and immovable fractions of kerogen and bitumen. Analysis of NMR relaxation spectra is performed to record how fluid composition, maturity, and viscosity change NMR experimental results. We present T1-T2 maps as thermal maturity of a water-saturated, sub-mature Woodford shale is increased at temperatures from 125 to 400 degrees Celsius. Experiments with applied fluid pressure in paraffinic mineral oil and DI water with varying fluid pH have been performed to mimic reservoir conditions in analysis of the relaxation of bulk fluid phases. We have recorded NMR spectra, T1-T2 maps, and fluid diffusion coefficients using a low-field (2 MHz) MagritekTM NMR. CO2 was injected at a pressure of 900 psi in an in house developed NMR pressure vessel made of torlon plastic. Observable 2D NMR shifts in immature kerogen formations as thermal maturity is increased show generation of lighter oils with increased maturity. CO2 injection leads to a decrease in bulk fluid relaxation time that is attributed to viscosity modification with gas presence. pH variation with increased CO2 presence were shown to not effect NMR spectra. From this, fluid properties have been shown to greatly affect NMR readings and must be taken into account for more accurate NMR reservoir characterization.

Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

NASA Technical Reports Server (NTRS)

Glass, James F.

1993-01-01

The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

NASA Astrophysics Data System (ADS)

Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

A compact membrane-driven diamond anvil cell and cryostat system for nuclear resonant scattering at high pressure and low temperature

DOE PAGES

Zhao, J. Y.; Bi, W.; Sinogeikin, S.; ...

2017-12-13

In order to study the vibrational and thermal dynamic properties of materials using the nuclear resonant inelastic X-ray scattering (NRIXS) and the hyperfine interactions and magnetic properties using the synchrotron Mössbauer spectroscopy (SMS) at simultaneously high pressure (multi-Mbar) and low temperature (T< 10 K), a new miniature panoramic diamond anvil cell (mini-pDAC) as well as a special gas membrane driven mechanism have been developed and implemented at 3ID, Advanced Photon Source. The gas membrane system allows in situ pressure tuning of the mini- pDAC at low temperature. The mini-pDAC fits into a specially designed compact liquid helium flow cryostat systemmore » to achieve low temperature, where liquid helium flows through the holder of the mini-pDAC to cool the sample more efficiently. The sample temperature as low as 9 K has been achieved. Through the membrane, the sample pressure as high as 1.4 Mbar has been generated from this mini-pDAC. The instrument has been routinely used at 3ID for NRIXS and SMS studies. In this paper, technical details of the mini-pDAC, membrane engaging mechanism and the cryostat system are described, and some experimental results are discussed.« less

A compact membrane-driven diamond anvil cell and cryostat system for nuclear resonant scattering at high pressure and low temperature

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

Zhao, J. Y.; Bi, W.; Sinogeikin, S.

In order to study the vibrational and thermal dynamic properties of materials using the nuclear resonant inelastic X-ray scattering (NRIXS) and the hyperfine interactions and magnetic properties using the synchrotron Mössbauer spectroscopy (SMS) at simultaneously high pressure (multi-Mbar) and low temperature (T< 10 K), a new miniature panoramic diamond anvil cell (mini-pDAC) as well as a special gas membrane driven mechanism have been developed and implemented at 3ID, Advanced Photon Source. The gas membrane system allows in situ pressure tuning of the mini- pDAC at low temperature. The mini-pDAC fits into a specially designed compact liquid helium flow cryostat systemmore » to achieve low temperature, where liquid helium flows through the holder of the mini-pDAC to cool the sample more efficiently. The sample temperature as low as 9 K has been achieved. Through the membrane, the sample pressure as high as 1.4 Mbar has been generated from this mini-pDAC. The instrument has been routinely used at 3ID for NRIXS and SMS studies. In this paper, technical details of the mini-pDAC, membrane engaging mechanism and the cryostat system are described, and some experimental results are discussed.« less

Pressure and temperature fluctuation simulation of J-PARC cryogenic hydrogen system

NASA Astrophysics Data System (ADS)

Tatsumoto, H.; Ohtsu, K.; Aso, T.; Kawakami, Y.

2015-12-01

The J-PARC cryogenic hydrogen system provides supercritical cryogenic hydrogen to the moderators at a pressure of 1.5 MPa and temperature of 18 K and removes 3.8 kW of nuclear heat from the 1 MW proton beam operation. We prepared a heater for thermal compensation and an accumulator, with a bellows structure for volume control, to mitigate the pressure fluctuation caused by switching the proton beam on and off. In this study, a 1-D simulation code named DiSC-SH2 was developed to understand the propagation of pressure and temperature propagations through the hydrogen loop due to on and off switching of the proton beam. We confirmed that the simulated dynamic behaviors in the hydrogen loop for 300-kW and 500-kW proton beam operations agree well with the experimental data under the same conditions.

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Schnitzler, Bruce G.; Borowoski, Stanley

2015-01-01

Four Revised Point of Departure NTR Engines were Designed and Analyzed using MCNP and NESS. All Four Engines Have Thermodynamically Closed Cycles at Nominal Chamber Pressures. 111 kilonewton (25 kip-force) Cermet Design Required Dedicated Heater Elements to Close the Cycle. Cermet Based Designs had Slightly Higher TW Ratios, but Required Substantially More U-235. NERVA Derived Criticality Limited Engine Could Operate at Lower Power and Thrust Levels Compared to the Criticality Limited Cermet Design.

Microhydration of caesium compounds: Cs, CsOH, CsI and Cs₂I₂ complexes with one to three H₂O molecules of nuclear safety interest.

PubMed

Sudolská, Mária; Cantrel, Laurent; Cernušák, Ivan

2014-04-01

Structure and thermodynamic properties (standard enthalpies of formation and Gibbs free energies) of hydrated caesium species of nuclear safety interest, Cs, CsOH, CsI and its dimer Cs₂I₂, with one up to three water molecules, are calculated to assess their possible existence in severe accident occurring to a pressurized water reactor. The calculations were performed using the coupled cluster theory including single, double and non-iterative triple substitutions (CCSD(T)) in conjunction with the basis sets (ANO-RCC) developed for scalar relativistic calculations. The second-order spin-free Douglas-Kroll-Hess Hamiltonian was used to account for the scalar relativistic effects. Thermodynamic properties obtained by these correlated ab initio calculations (entropies and thermal capacities at constant pressure as a function of temperature) are used in nuclear accident simulations using ASTEC/SOPHAEROS software. Interaction energies, standard enthalpies and Gibbs free energies of successive water molecules addition determine the ordering of the complexes. CsOH forms the most hydrated stable complexes followed by CsI, Cs₂I₂, and Cs. CsOH still exists in steam atmosphere even at quite high temperature, up to around 1100 K.

Enrichment of 57Fe isotope in neutron flux of nuclear reactors observed by Mössbauer spectroscopy.

PubMed

Sawicki, Jerzy A

2018-02-01

The abundance of 57 Fe isotope in nuclear reactor core materials can be considerably enriched by neutron-capture 56 Fe(n,γ) reactions. This is demonstrated using the sections of Zr-2.5 wt.%Nb pressure tubes removed from two CANDU* reactors. The tubes contained 0.11 and 0.04wt% Fe and were irradiated for about 10 effective full power years (EFPY) up to ~10 26 n/m 2 fast neutron (E > 1MeV) fluencies. The Mössbauer spectra of 57 Fe in irradiated samples indicated up to 10 times larger areas than unirradiated off-cuts from the same pressure tubes. The observed effect is in accord with the values calculated for known thermal neutron-capture cross-sections and resonance capture integrals, neutron flux profiles and spectra, and times of irradiation. The build-up of 57 Fe facilitated recording Mössbauer absorption spectra of alloys with minor amount of Fe down to ~ 400ppm, despite intense background radiation emitted by samples. These findings can open new possibilities in post-irradiation studies of alloys used in nuclear reactors and in other objects subjected to large neutron fluencies. Copyright © 2017 Elsevier Ltd. All rights reserved.

Performance Expectations of Closed-Brayton-Cycle Heat Exchangers in 100-kWe Nuclear Space Power Systems

NASA Technical Reports Server (NTRS)

Barrett, Michael J.

2003-01-01

Performance expectations of closed-Brayton-cycle heat exchangers to be used in 100-kWe nuclear space power systems were forecast. Proposed cycle state points for a system supporting a mission to three of Jupiter s moons required effectiveness values for the heat-source exchanger, recuperator and rejection exchanger (gas cooler) of 0.98,0.95 and 0.97, respectively. Performance parameters such as number of thermal units (Nm), equivalent thermal conductance (UA), and entropy generation numbers (Ns) varied from 11 to 19,23 to 39 kWK, and 0.019 to 0.023 for some standard heat exchanger configurations. Pressure-loss contributions to entropy generation were significant; the largest frictional contribution was 114% of the heat-transfer irreversibility. Using conventional recuperator designs, the 0.95 effectiveness proved difficult to achieve without exceeding other performance targets; a metallic, plate-fin counterflow solution called for 15% more mass and 33% higher pressure-loss than the target values. Two types of gas-coolers showed promise. Single-pass counterflow and multipass cross-counterflow arrangements both met the 0.97 effectiveness requirement. Potential reliability-related advantages of the cross-countefflow design were noted. Cycle modifications, enhanced heat transfer techniques and incorporation of advanced materials were suggested options to reduce system development risk. Carbon-carbon sheeting or foam proved an attractive option to improve overall performance.

Performance Expectations of Closed-Brayton-Cycle Heat Exchangers in 100-kWe Nuclear Space Power Systems

NASA Technical Reports Server (NTRS)

Barrett, Michael J.

2003-01-01

Performance expectations of closed-Brayton-cycle heat exchangers to be used in 100-k We nuclear space power systems were forecast. Proposed cycle state points for a system supporting a mission to three of Jupiter's moons required effectiveness values for the heat-source exchanger, recuperator and rejection exchanger (gas cooler) of 0.98, 0.95, and 0.97, respectively. Performance parameters such as number of thermal units (Ntu), equivalent thermal conductance (UA), and entropy generation numbers (Ns) varied from 11 to 19, 23 to 39 kW/K, and 0.019 to 0.023 for some standard heat exchanger configurations. Pressure-loss contributions to entropy generation were significant; the largest frictional contribution was 114% of the heat transfer irreversibility. Using conventional recuperator designs, the 0.95 effectiveness proved difficult to achieve without exceeding other performance targets; a metallic, plate-fin counterflow solution called for 15% more mass and 33% higher pressure-loss than the target values. Two types of gas-coolers showed promise. Single-pass counterflow and multipass cross-counterflow arrangements both met the 0.97 effectiveness requirement. Potential reliability-related advantages of the cross-counterflow design were noted. Cycle modifications, enhanced heat transfer techniques and incorporation of advanced materials were suggested options to reduce system development risk. Carbon-carbon sheeting or foam proved an attractive option to improve overall performance.

Spin Crossover and the Magnetic P- T Phase Diagram of Hematite at High Hydrostatic Pressures and Cryogenic Temperatures

NASA Astrophysics Data System (ADS)

Gavriliuk, A. G.; Struzhkin, V. V.; Mironovich, A. A.; Lyubutin, I. S.; Troyan, I. A.; Chow, P.; Xiao, Y.

2018-02-01

The magnetic properties of the α-Fe2O3 hematite at a high hydrostatic pressure have been studied by synchrotron Mössbauer spectroscopy (nuclear forward scattering (NFS)) on iron nuclei. Time-domain NFS spectra of hematite have been measured in a diamond anvil cell in the pressure range of 0-72 GPa and the temperature range of 36-300 K in order to study the magnetic properties at a phase transition near a critical pressure of 50 GPa. In addition, Raman spectra at room temperature have been studied in the pressure range of 0-77 GPa. Neon has been used as a pressure-transmitting medium. The appearance of an intermediate electronic state has been revealed at a pressure of 48 GPa. This state is probably related to the spin crossover in Fe3+ ions at their transition from the high-spin state (HS, S = 5/2) to a low-spin one (LS, S = 1/2). It has been found that the transient pressure range of the HS-LS crossover is extended from 48 to 55 GPa and is almost independent of the temperature. This surprising result differs fundamentally from other cases of the spin crossover in Fe3+ ions observed in other crystals based on iron oxides. The transition region of spin crossover appears because of thermal fluctuations between HS and LS states in the critical pressure range and is significantly narrowed at cooling because of the suppression of thermal excitations. The magnetic P- T phase diagram of α-Fe2O3 at high pressures and low temperatures in the spin crossover region has been constructed according to the results of measurements.

Cryogenic Fluid Management Technology Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, B. D.; Caffrey, J.; Hedayat, A.; Stephens, J.; Polsgrove, R.

2015-01-01

Cryogenic fluid management technology is critical to the success of future nuclear thermal propulsion powered vehicles and long duration missions. This paper discusses current capabilities in key technologies and their development path. The thermal environment, complicated from the radiation escaping a reactor of a nuclear thermal propulsion system, is examined and analysis presented. The technology development path required for maintaining cryogenic propellants in this environment is reviewed. This paper is intended to encourage and bring attention to the cryogenic fluid management technologies needed to enable nuclear thermal propulsion powered deep space missions.

Thermal, High Pressure, and Electric Field Processing Effects on Plant Cell Membrane Integrity and Relevance to Fruit and Vegetable Quality

PubMed Central

Gonzalez, Maria E; Barrett, Diane M

2010-01-01

Advanced food processing methods that accomplish inactivation of microorganisms but minimize adverse thermal exposure are of great interest to the food industry. High pressure (HP) and pulsed electric field (PEF) processing are commercially applied to produce high quality fruit and vegetable products in the United States, Europe, and Japan. Both microbial and plant cell membranes are significantly altered following exposure to heat, HP, or PEF. Our research group sought to quantify the degree of damage to plant cell membranes that occurs as a result of exposure to heat, HP, or PEF, using the same analytical methods. In order to evaluate whether new advanced processing methods are superior to traditional thermal processing methods, it is necessary to compare them. In this review, we describe the existing state of knowledge related to effects of heat, HP, and PEF on both microbial and plant cells. The importance and relevance of compartmentalization in plant cells as it relates to fruit and vegetable quality is described and various methods for quantification of plant cell membrane integrity are discussed. These include electrolyte leakage, cell viability, and proton nuclear magnetic resonance (1H-NMR). PMID:20492210

Thermal, high pressure, and electric field processing effects on plant cell membrane integrity and relevance to fruit and vegetable quality.

PubMed

Gonzalez, Maria E; Barrett, Diane M

2010-09-01

Advanced food processing methods that accomplish inactivation of microorganisms but minimize adverse thermal exposure are of great interest to the food industry. High pressure (HP) and pulsed electric field (PEF) processing are commercially applied to produce high quality fruit and vegetable products in the United States, Europe, and Japan. Both microbial and plant cell membranes are significantly altered following exposure to heat, HP, or PEF. Our research group sought to quantify the degree of damage to plant cell membranes that occurs as a result of exposure to heat, HP, or PEF, using the same analytical methods. In order to evaluate whether new advanced processing methods are superior to traditional thermal processing methods, it is necessary to compare them. In this review, we describe the existing state of knowledge related to effects of heat, HP, and PEF on both microbial and plant cells. The importance and relevance of compartmentalization in plant cells as it relates to fruit and vegetable quality is described and various methods for quantification of plant cell membrane integrity are discussed. These include electrolyte leakage, cell viability, and proton nuclear magnetic resonance (¹H-NMR).

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

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

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

1997-02-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices ofmore » Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.« less

Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

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

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internalmore » convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.« less

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

A Programmatic and Engineering Approach to the Development of a Nuclear Thermal Rocket for Space Exploration

NASA Technical Reports Server (NTRS)

Bordelon, Wayne J., Jr.; Ballard, Rick O.; Gerrish, Harold P., Jr.

2006-01-01

With the announcement of the Vision for Space Exploration on January 14, 2004, there has been a renewed interest in nuclear thermal propulsion. Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions; however, the cost to develop a nuclear thermal rocket engine system is uncertain. Key to determining the engine development cost will be the engine requirements, the technology used in the development and the development approach. The engine requirements and technology selection have not been defined and are awaiting definition of the Mars architecture and vehicle definitions. The paper discusses an engine development approach in light of top-level strategic questions and considerations for nuclear thermal propulsion and provides a suggested approach based on work conducted at the NASA Marshall Space Flight Center to support planning and requirements for the Prometheus Power and Propulsion Office. This work is intended to help support the development of a comprehensive strategy for nuclear thermal propulsion, to help reduce the uncertainty in the development cost estimate, and to help assess the potential value of and need for nuclear thermal propulsion for a human Mars mission.

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in 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. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

First Conclusions of the WPEC/Subgroup-22 Nuclear Data for Improved LEU-LWR Reactivity Predictions

NASA Astrophysics Data System (ADS)

Courcelle, Arnaud

2005-05-01

This paper is a summary of a collective work in the framework of the Working Party in International Nuclear Data Evaluation and Co-operation (WPEC) to investigate the reasons for systematic reactivity underprediction of thermal LEU-LWR (Low-Enriched Uranium, Light-Water Reactor). This keff underprediction (≈ -500 pcm) is observed with the most recent nuclear data libraries (ENDF/B-VI.8, JENDL3.3 and JEFF3.0) This report reviews the evaluation work performed at several laboratories [Oak Ridge National Laboratory (ORNL), Los Alamos National Laboratory (LANL), Commissariat a l'énergie atomique de Bruyeres-Le-Chatel (CEA-BRC), International Atomic Energy Agency (IAEA)] as well as the integral tests (mainly at LANL, Knoll Atomic Power Laboratory (KAPL), Bettis Atomic Power Laboratory (BAPL), Nuclear Research and Consultancy Group NRG-Petten, CEA and IAEA) of the successive versions of the new evaluated files. The present status of the work can be summarized as follows: • Improved evaluations of 238U inelastic data proposed by LANL and CEA-BRC were tested against integral benchmarks and partially improve the reactivity prediction. • The thermal capture cross-section of 238U has been revised, and a new evaluation of 238U resonance parameters, up to 20 keV, is in progress at ORNL. Integral tests have ensured that the modifications of 238U capture cross-section in the thermal and resolved range were still compatible with 238U integral measurements (238U capture rate ratios measured in critical facilities and 239Pu build-up prediction in a depleted pressurized water reactor (PWR) assembly). It is demonstrated that the combination of the new inelastic data (LANL or BRC) with the preliminary ORNL resonance parameter set gives a good correction of the reactivity under-estimation. The provisional conclusions of this collective work are expected to contribute toward the improvement of the future versions of nuclear data libraries.

Nondestructive Technique To Assess Embrittlement In Steels

NASA Technical Reports Server (NTRS)

Allison, Sidney G.; Yost, William T.; Cantrell, John H.

1990-01-01

Recent research at NASA Langley Research Center led to identification of nondestructive technique for detection of temper embrittlement in HY80 steel. Measures magnetoacoustic emission associated with reversible motion of domain walls at low magnetic fields. Of interest to engineers responsible for reliability and safety of various dynamically loaded and/or thermally cycled steel parts. Applications include testing of landing gears, naval vessels, and parts subjected to heat, such as those found in steam-pipe fittings, boilers, turbine rotors, and nuclear pressure vessels.

Current status of GALS setup in JINR

NASA Astrophysics Data System (ADS)

Zemlyanoy, S.; Avvakumov, K.; Fedosseev, V.; Bark, R.; Blazczak, Z.; Janas, Z.

2017-11-01

This is a brief report on the current status of the new GAs cell based Laser ionization Setup (GALS) at the Flerov Laboratory for Nuclear Reactions (FLNR) of the Joint Institute for Nuclear Research (JINR) in Dubna. GALS will exploit available beams from the U-400M cyclotron in low energy multi-nucleon transfer reactions to study exotic neutron-rich nuclei located in the "north-east" region of nuclear map. Products from 4.5 to 9 MeV/nucleon heavy-ion collisions, such as 136Xe on 208Pb, are thermalized and neutralized in a high pressure gas cell and subsequently selectively laser re-ionized. In order to choose the best scheme of ion extraction the results of computer simulations of two different systems are presented. The first off- and online experiment will be performed on osmium atoms that is regarded as a most convenient element for producing isotopes with neutron number in the vicinity of the magic N = 126.

Gibbs free energy of reactions involving SiC, Si3N4, H2, and H2O as a function of temperature and pressure

NASA Technical Reports Server (NTRS)

Isham, M. A.

1992-01-01

Silicon carbide and silicon nitride are considered for application as structural materials and coating in advanced propulsion systems including nuclear thermal. Three-dimensional Gibbs free energy were constructed for reactions involving these materials in H2 and H2/H2O. Free energy plots are functions of temperature and pressure. Calculations used the definition of Gibbs free energy where the spontaneity of reactions is calculated as a function of temperature and pressure. Silicon carbide decomposes to Si and CH4 in pure H2 and forms a SiO2 scale in a wet atmosphere. Silicon nitride remains stable under all conditions. There was no apparent difference in reaction thermodynamics between ideal and Van der Waals treatment of gaseous species.

Compaction and High-Pressure Response of Granular Tantalum Oxide

NASA Astrophysics Data System (ADS)

Vogler, Tracy; Root, Seth; Knudson, Marcus; Thornhill, Tom; Reinhart, William

2015-06-01

The dynamic behavior of nearly fully-dense and porous tantalum oxide (Ta2O5) is studied. Two particle morphologies are used to obtain two distinct initial tap densities, which correspond to approximately 40% and 15% of crystalline density. The response is characterized from low pressures, which result in incomplete compaction, to very high pressures where the thermal component of the EOS dominates. Issues related to a possible phase transformation along the Hugoniot and to establishing reasonable error bars on the experimental data will be discussed. The suitability of continuum and mesoscale models to capture the experimental results will be examined. Sandia National Laboratories is a multi-program laboratory operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Company, for the U.S. Department of Energy's National Nuclear Security Administration under Contract DE-AC04-94AL85000.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Buckling of thermally fluctuating spherical shells: Parameter renormalization and thermally activated barrier crossing

NASA Astrophysics Data System (ADS)

Baumgarten, Lorenz; Kierfeld, Jan

2018-05-01

We study the influence of thermal fluctuations on the buckling behavior of thin elastic capsules with spherical rest shape. Above a critical uniform pressure, an elastic capsule becomes mechanically unstable and spontaneously buckles into a shape with an axisymmetric dimple. Thermal fluctuations affect the buckling instability by two mechanisms. On the one hand, thermal fluctuations can renormalize the capsule's elastic properties and its pressure because of anharmonic couplings between normal displacement modes of different wavelengths. This effectively lowers its critical buckling pressure [Košmrlj and Nelson, Phys. Rev. X 7, 011002 (2017), 10.1103/PhysRevX.7.011002]. On the other hand, buckled shapes are energetically favorable already at pressures below the classical buckling pressure. At these pressures, however, buckling requires to overcome an energy barrier, which only vanishes at the critical buckling pressure. In the presence of thermal fluctuations, the capsule can spontaneously overcome an energy barrier of the order of the thermal energy by thermal activation already at pressures below the critical buckling pressure. We revisit parameter renormalization by thermal fluctuations and formulate a buckling criterion based on scale-dependent renormalized parameters to obtain a temperature-dependent critical buckling pressure. Then we quantify the pressure-dependent energy barrier for buckling below the critical buckling pressure using numerical energy minimization and analytical arguments. This allows us to obtain the temperature-dependent critical pressure for buckling by thermal activation over this energy barrier. Remarkably, both parameter renormalization and thermal activation lead to the same parameter dependence of the critical buckling pressure on temperature, capsule radius and thickness, and Young's modulus. Finally, we study the combined effect of parameter renormalization and thermal activation by using renormalized parameters for the energy barrier in thermal activation to obtain our final result for the temperature-dependent critical pressure, which is significantly below the results if only parameter renormalization or only thermal activation is considered.

Multi-Fluid Geo-Energy Systems for Bulk and Thermal Energy Storage and Dispatchable Renewable and Low-Carbon Electricity

NASA Astrophysics Data System (ADS)

Buscheck, T. A.; Randolph, J.; Saar, M. O.; Hao, Y.; Sun, Y.; Bielicki, J. M.

2014-12-01

Integrating renewable energy sources into electricity grids requires advances in bulk and thermal energy storage technologies, which are currently expensive and have limited capacity. We present an approach that uses the huge fluid and thermal storage capacity of the subsurface to harvest, store, and dispatch energy from subsurface (geothermal) and surface (solar, nuclear, fossil) thermal resources. CO2 captured from fossil-energy systems and N2 separated from air are injected into permeable formations to store pressure, generate artesian flow of brine, and provide additional working fluids. These enable efficient fluid recirculation, heat extraction, and power conversion, while adding operational flexibility. Our approach can also store and dispatch thermal energy, which can be used to levelize concentrating solar power and mitigate variability of wind and solar power. This may allow low-carbon, base-load power to operate at full capacity, with the stored excess energy being available to addresss diurnal and seasonal mismatches between supply and demand. Concentric rings of horizontal injection and production wells are used to create a hydraulic divide to store pressure, CO2, N2, and thermal energy. Such storage can take excess power from the grid and excess thermal energy, and dispatch that energy when it is demanded. The system is pressurized and/or heated when power supply exceeds demand and depressurized when demand exceeds supply. Supercritical CO2 and N2 function as cushion gases to provide enormous pressure-storage capacity. Injecting CO2 and N2 displaces large quantities of brine, reducing the use of fresh water. Geologic CO2 storage is a crucial option for reducing CO2 emissions, but valuable uses for CO2 are needed to justify capture costs. The initial "charging" of our system requires permanently isolating large volumes of CO2 from the atmosphere and thus creates a market for its disposal. Our approach is designed for locations where a permeable geologic formation is overlain by an impermeable formation that constrains migration of buoyant CO2 and/or N2, and heated brine. Such geologic conditions exist over nearly half of the contiguous United States. This work was performed under the auspices of the U.S. DOE by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Underground physics and the barometric pumping effect observed for thermal neutron flux underground

NASA Astrophysics Data System (ADS)

Stenkin, Yu. V.; Alekseenko, V. V.; Gromushkin, D. M.; Sulakov, V. P.; Shchegolev, O. B.

2017-05-01

It is known that neutron background is a major problem for low-background experiments carrying out underground, such as dark matter search, double-beta decay searches and other experiments known as Underground Physics. We present here some results obtained with the en-detector of 0.75 m2, which is running for more than 4 years underground at a depth of 25 m water equivalent in Skobeltsyn Institute of Nuclear Physics, Moscow State University. Some spontaneous increases in thermal neutron flux up to a factor of 3 were observed in delayed anti-correlation with barometric pressure. The phenomenon can be explained by the radon barometric pumping effect resulting in similar effect in neutron flux being produced in (α, n)-reactions by alpha-decays of radon and its daughters in surrounding rock. This is the first demonstration of the barometric pumping effect observed in thermal neutron flux underground.

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

Experimental investigations, modeling, and analyses of high-temperature devices for space applications: Part 1. Final report, June 1996--December 1998

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

Tournier, J.; El-Genk, M.S.; Huang, L.

1999-01-01

The Institute of Space and Nuclear Power Studies at the University of New Mexico has developed a computer simulation of cylindrical geometry alkali metal thermal-to-electric converter cells using a standard Fortran 77 computer code. The objective and use of this code was to compare the experimental measurements with computer simulations, upgrade the model as appropriate, and conduct investigations of various methods to improve the design and performance of the devices for improved efficiency, durability, and longer operational lifetime. The Institute of Space and Nuclear Power Studies participated in vacuum testing of PX series alkali metal thermal-to-electric converter cells and developedmore » the alkali metal thermal-to-electric converter Performance Evaluation and Analysis Model. This computer model consisted of a sodium pressure loss model, a cell electrochemical and electric model, and a radiation/conduction heat transfer model. The code closely predicted the operation and performance of a wide variety of PX series cells which led to suggestions for improvements to both lifetime and performance. The code provides valuable insight into the operation of the cell, predicts parameters of components within the cell, and is a useful tool for predicting both the transient and steady state performance of systems of cells.« less

Experimental investigations, modeling, and analyses of high-temperature devices for space applications: Part 2. Final report, June 1996--December 1998

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

Tournier, J.; El-Genk, M.S.; Huang, L.

1999-01-01

The Institute of Space and Nuclear Power Studies at the University of New Mexico has developed a computer simulation of cylindrical geometry alkali metal thermal-to-electric converter cells using a standard Fortran 77 computer code. The objective and use of this code was to compare the experimental measurements with computer simulations, upgrade the model as appropriate, and conduct investigations of various methods to improve the design and performance of the devices for improved efficiency, durability, and longer operational lifetime. The Institute of Space and Nuclear Power Studies participated in vacuum testing of PX series alkali metal thermal-to-electric converter cells and developedmore » the alkali metal thermal-to-electric converter Performance Evaluation and Analysis Model. This computer model consisted of a sodium pressure loss model, a cell electrochemical and electric model, and a radiation/conduction heat transfer model. The code closely predicted the operation and performance of a wide variety of PX series cells which led to suggestions for improvements to both lifetime and performance. The code provides valuable insight into the operation of the cell, predicts parameters of components within the cell, and is a useful tool for predicting both the transient and steady state performance of systems of cells.« less

Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

NASA Astrophysics Data System (ADS)

McCoy, Kevin; Mays, Claude

2008-04-01

The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO 2 fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor kinf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO 2 fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO 2, UO 2 with 4.0 vol.% BeO, and UO 2 with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO 2 fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the kinf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO 2.

Bimodal Nuclear Thermal Rocket Sizing and Trade Matrix for Lunar, Near Earth Asteroid and Mars Missions

NASA Astrophysics Data System (ADS)

McCurdy, David R.; Krivanek, Thomas M.; Roche, Joseph M.; Zinolabedini, Reza

2006-01-01

The concept of a human rated transport vehicle for various near earth missions is evaluated using a liquid hydrogen fueled Bimodal Nuclear Thermal Propulsion (BNTP) approach. In an effort to determine the preliminary sizing and optimal propulsion system configuration, as well as the key operating design points, an initial investigation into the main system level parameters was conducted. This assessment considered not only the performance variables but also the more subjective reliability, operability, and maintainability attributes. The SIZER preliminary sizing tool was used to facilitate rapid modeling of the trade studies, which included tank materials, propulsive versus an aero-capture trajectory, use of artificial gravity, reactor chamber operating pressure and temperature, fuel element scaling, engine thrust rating, engine thrust augmentation by adding oxygen to the flow in the nozzle for supersonic combustion, and the baseline turbopump configuration to address mission redundancy and safety requirements. A high level system perspective was maintained to avoid focusing solely on individual component optimization at the expense of system level performance, operability, and development cost.

Thermodynamic properties and equation of state of liquid lead and lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Sobolev, V. P.; Schuurmans, P.; Benamati, G.

2008-06-01

Since the 1950s, liquid lead (Pb) and lead-bismuth eutectic (Pb-Bi) have been studied in the USA, Canada and in the former-USSR as potential coolants for nuclear installations due to their very attractive thermophysical and neutronic properties. However, experimental data on the thermal properties of these coolants in the temperature range of interest are still incomplete and often contradictory. This makes it very difficult to perform design calculations and to analyse the normal and abnormal behaviour of nuclear installations where these coolants are expected to be used. Recently, a compilation of heavy liquid metal (HLM) properties along with recommendations for its use was prepared by the OECD/NEA Working Party on Fuel Cycle (WPFC) Expert Group on Lead-Bismuth Eutectic Technology. A brief review of this compilation and some new data are presented in this article. A set of correlations for the temperature dependence of the main thermodynamic properties of Pb and Pb-Bi(e) at normal pressure, and a set of simplified thermal and caloric equations of state for the liquid phase are proposed.

Nuclear thermal propulsion program overview

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1991-01-01

Nuclear thermal propulsion program is described. The following subject areas are covered: lunar and Mars missions; national space policy; international cooperation in space exploration; propulsion technology; nuclear rocket program; and budgeting.

Heavy-section steel technology and irradiation programs-retrospective and prospective views

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

Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M

In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). Dr. Spencer H. Bush of Battelle Northwest Laboratory, the man being honored by this symposium, representing the ACRS, was one of the Staff Advisors for the program and helped to guide its technical direction. In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and this year the HSST/HSSImore » Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrate 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels. This paper presents a summary of results from those programs with a view to future activities. The HSST Program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and non-destructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge (FAVOR) code currently being used for PTS evaluations. In the absence of radiation damage to the RPV, fracture of the vessel is improbable. However, exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The HSSI Program has conducted a series of experiments to assess the effects of neutron irradiation on RPV material behavior, especially fracture toughness. These studies have included RPV plates and welds, varying chemical compositions, and fracture toughness specimens up to 4 in. thickness. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. Results from the HSST and HSSI Program are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety.« less

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

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

Nanstad, Randy K.; Odette, George R.; Almirall, Nathan

2017-05-01

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughnessmore » loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.« less

Shock equation of state of 6LiH to 1.1 TPa

NASA Astrophysics Data System (ADS)

Lazicki, A.; London, R. A.; Coppari, F.; Erskine, D.; Whitley, H. D.; Caspersen, K. J.; Fratanduono, D. E.; Morales, M. A.; Celliers, P. M.; Eggert, J. H.; Millot, M.; Swift, D. C.; Collins, G. W.; Kucheyev, S. O.; Castor, J. I.; Nilsen, J.

2017-10-01

Using laser-generated shock waves, we have measured pressure, density, and temperature of LiH on the principal Hugoniot between 260 and 1100 GPa (2.6-11 Mbar) and on a second-shock Hugoniot up to 1400 GPa to near fivefold compression, extending the maximum pressure reached in non-nuclear experiments by a factor of two. We observe the onset of metal-like reflectivity consistent with temperature-induced ionization of the Li 2s electron, and no sign of additional changes in ionization up to the maximum pressure. Our measurements are in good agreement with gas gun, Z-machine, and underground test data and are accurately described by quantum molecular dynamics simulations. The results confirm the validity of equation of state models built on an average-atom description of the electron-thermal contribution to the free energy and a density-dependent Grüneisen parameter to describe shock response of LiH over this pressure range.

The Liquid Annular Reactor System (LARS) propulsion

NASA Technical Reports Server (NTRS)

Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

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

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Effects of the Fabrication Process and Thermal Cycling on the Oxidation of Zirconium-Niobium Pressure Tubes

NASA Astrophysics Data System (ADS)

Nam, Cheol

2009-12-01

Pressure tubes made of Zr-2.5%Nb alloy are used to contain fuels and coolant in CANDU nuclear power reactors The pressure tube oxidizes during reactor operation and hydrogen ingress through the oxide grown on the tube limits its lifetime. Little attention was paid to the intermediate tube manufacturing processes in enhancing the oxidation resistance. In addition, the oxide grown on the tube experiences various thermal cycles depending on the reactor shutdown and startup cycles. To address these two aspects and to better understand the oxidation process of the Zr-2.5Nb tube, research was conducted in two parts: (i) effects of tube fabrication on oxidation behavior, and (ii) thermal cycling behaviors of oxides grown on a pressure tube. In the first part, the optimum manufacturing process was pursued to improve the corrosion resistance of Zr-2.5Nb tubes. Experimental micro-tubes were fabricated with various manufacturing routes in the stages of billet preparation, hot extrusion and cold drawing. These were oxidized in air at 400°C and 500°C, and in an autoclave at 360°C lithiated water. Microstructure and texture of the tubes and oxides were characterized with X-ray diffraction, scanning electron microscope and optical microscope. Special emphasis was given to examinations of the metal/oxide interface structures. A correlation between the manufacturing process and oxidation resistance was investigated in terms of tube microstructure and the metal/oxide interface structure. As a result, it was consistently observed that uniform interface structures were formed on the tubes which had a fine distribution of secondary phases. These microstructures were found to be beneficial in enhancing the oxidation resistance as opposed to the tubes that had coarse and continuous beta-Zr phases. Based on these observations, a schematic model of the oxidation process was proposed with respect to the oxidation resistance under oxidizing temperatures of 360°C, 400°C and 500°C. In the second part, the oxides grown on a standard Zr-2.5Nb pressure tube were analyzed by X-ray diffraction peak broadening and line shift. Crystallite size, t-ZrO2 fraction and residual stress of the zirconium oxides were investigated upon several thermal cycles at DeltaT range of 500°C--750°C. The oxide residual stresses measured by the sin2psi method were always compressive around 2 GPa. Different stress-states were noticed with the oxides grown on different sections of pressure tube. The compressive stress was released when the oxide was thermally cycled at the highest DeltaT of 750°C. Discussion was given to the effects of anisotropic nature of thermal expansion coefficients and crystallographic texture on the stress-state of Zr oxides.

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

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

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Thermal separation of soil particles from thermal conductivity measurement under various air pressures.

PubMed

Lu, Sen; Ren, Tusheng; Lu, Yili; Meng, Ping; Zhang, Jinsong

2017-01-05

The thermal conductivity of dry soils is related closely to air pressure and the contact areas between solid particles. In this study, the thermal conductivity of two-phase soil systems was determined under reduced and increased air pressures. The thermal separation of soil particles, i.e., the characteristic dimension of the pore space (d), was then estimated based on the relationship between soil thermal conductivity and air pressure. Results showed that under both reduced and increased air pressures, d estimations were significantly larger than the geometrical mean separation of solid particles (D), which suggested that conductive heat transfer through solid particles dominated heat transfer in dry soils. The increased air pressure approach gave d values lower than that of the reduced air pressure method. With increasing air pressure, more collisions between gas molecules and solid surface occurred in micro-pores and intra-aggregate pores due to the reduction of mean free path of air molecules. Compared to the reduced air pressure approach, the increased air pressure approach expressed more micro-pore structure attributes in heat transfer. We concluded that measuring thermal conductivity under increased air pressure procedures gave better-quality d values, and improved soil micro-pore structure estimation.

High pressure hydrogen stabilised by quantum nuclear motion

NASA Astrophysics Data System (ADS)

Needs, Richard; Monserrat, Bartomeu; Pickard, Chris

Hydrogen under extreme pressures is of fundamental interest, as it might exhibit exotic physical phenomena, and of practical interest, as it is a major component of many astrophysical objects. Structure searches have been successful at identifying promising candidates for the known phases of high pressure hydrogen. However, these searches have so far been restricted to the location of minima of the potential energy landscape. In this talk, we will describe a new structure searching method, ``saddle-point ab initio random structure searching'' (sp-AIRSS), that allows us to identify structures associated with saddle points of the potential energy landscape. Using sp-AIRSS, we find two new high-pressure hydrogen structures that exhibit a harmonic dynamical instability, but quantum and thermal anharmonic motion render them dynamically stable. These structures are formed by mixed layers of strongly and softly bound hydrogen molecules, and become thermodynamically competitive at the highest pressures reached in experiment. The experimental implications of these new structures will also be discussed. BM is supported by Robinson College, Cambridge, and the Cambridge Philosophical Society. RJN and CJP are supported by the Engineering and Physical Sciences Research Council (EPSRC) of the UK.

Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

NASA Astrophysics Data System (ADS)

Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

2004-02-01

A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

EV space suit gloves (passive)

NASA Technical Reports Server (NTRS)

Fletcher, E. G.; Dodson, J. D.; Elkins, W.; Tickner, E. G.

1975-01-01

A pair of pressure and thermal insulating overgloves to be used with an Extravehicular (EV) suit assembly was designed, developed, fabricated, and tested. The design features extensive use of Nomex felt materials in lieu of the multiple layer insulation formerly used with the Apollo thermal glove. The glove theoretically satisfies all of the thermal requirements. The presence of the thermal glove does not degrade pressure glove tactility by more than the acceptable 10% value. On the other hand, the thermal glove generally degrades pressure glove mobility by more than the acceptable 10% value, primarily in the area of the fingers. Life cycling tests were completed with minimal problems. The thermal glove/pressure glove ensemble was also tested for comfort; the test subjects found no problems with the thermal glove although they did report difficulties with pressure points on the pressure glove which were independent of the thermal glove.

Mean-field study of hot β -stable protoneutron star matter: Impact of the symmetry energy and nucleon effective mass

NASA Astrophysics Data System (ADS)

Tan, Ngo Hai; Loan, Doan Thi; Khoa, Dao T.; Margueron, Jerome

2016-03-01

A consistent Hartree-Fock study of the equation of state (EOS) of asymmetric nuclear matter at finite temperature has been performed using realistic choices of the effective, density-dependent nucleon-nucleon (NN ) interaction, which were successfully used in different nuclear structure and reaction studies. Given the importance of the nuclear symmetry energy in the neutron star formation, EOSs associated with different behaviors of the symmetry energy were used to study hot asymmetric nuclear matter. The slope of the symmetry energy and nucleon effective mass with increasing baryon density was found to affect the thermal properties of nuclear matter significantly. Different density-dependent NN interactions were further used to study the EOS of hot protoneutron star (PNS) matter of the n p e μ ν composition in β equilibrium. The hydrostatic configurations of PNS in terms of the maximal gravitational mass Mmax and radius, central density, pressure, and temperature at the total entropy per baryon S /A =1 ,2 , and 4 have been determined in both the neutrino-free and neutrino-trapped scenarios. The obtained results show consistently a strong impact of the symmetry energy and nucleon effective mass on thermal properties and composition of hot PNS matter. Mmax values obtained for the (neutrino-free) β -stable PNS at S /A =4 were used to assess time tBH of the collapse of a 40 M⊙ protoneutron progenitor to a black hole, based on a correlation between tBH and Mmax found from the hydrodynamic simulation by Hempel et al. [Astrophys. J. 748, 70 (2012), 10.1088/0004-637X/748/1/70].

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

2013-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

Progress on DCLL Blanket Concept

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

Wong, Clement; Abdou, M.; Katoh, Yutai

2013-09-01

Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (~700°C) frommore » the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. When performing the function as the Interface Coordinator for the DCLL blanket concept, we had been developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We had estimated the necessary ancillary equipment that will be needed at the ITER site and a detailed safety impact report has been prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper will be a summary report on the progress of the DCLL TBM design and R&Ds for the DCLL blanket concept.« less

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to 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. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Variations of Thermal Pressure for Solids along the Principal Hugoniot

NASA Astrophysics Data System (ADS)

Gong, Zizheng; Yu, Hui; Deng, Liwei; Zhang, Li; Yang, Jinke

2006-07-01

The behavior of thermal pressure PTH for all kinds of solid materials was investigated using the lattice dynamics theory up to 500GPa. The results show that for most metals, ionic crystal and minerals, the thermal pressure is approximately independent on volume, whereas the thermal pressure of a few solids has strong dependence on volume. The volume dependence of thermal pressure has no relation with the chemical bonding type and crystal structure of materials, but is correlated with the Debye temperature ΘD and the second Grüneisen parameter q. The ratio of the thermal pressure to the total pressure (PTH /PTotal) along the Hugoniot keeps constant over a wide compression range, not only for non-porous materials but also for porous materials within certain porosity, which could explain the existence of material constant parameter β along solid Hugoniot.

Permeability and elastic properties of cracked glass under pressure

NASA Astrophysics Data System (ADS)

Ougier-Simonin, A.; GuéGuen, Y.; Fortin, J.; Schubnel, A.; Bouyer, F.

2011-07-01

Fluid flow in rocks is allowed through networks of cracks and fractures at all scales. In fact, cracks are of high importance in various applications ranging from rock elastic and transport properties to nuclear waste disposal. The present work aims at investigating thermomechanical cracking effects on elastic wave velocities, mechanical strength, and permeability of cracked glass under pressure. We performed the experiments on a triaxial cell at room temperature which allows for independent controls of the confining pressure, the axial stress, and pore pressure. We produced cracks in original borosilicate glass samples with a reproducible method (thermal treatment with a thermal shock of 300°C). The evolution of the elastic and transport properties have been monitored using elastic wave velocity sensors, strain gage, and flow measurements. The results obtained evidence for (1) a crack family with identified average aspect ratio and crack aperture, (2) a very small permeability which decreases as a power (exponential) function of pressure, and depends on (3) the crack aperture cube. We also show that permeability behavior of a cracked elastic brittle solid is reversible and independent of the fluid nature. Two independent methods (permeability and elastic wave velocity measurements) give these consistent results. This study provides data on the mechanical and transport properties of an almost ideal elastic brittle solid in which a crack population has been introduced. Comparisons with similar data on rocks allow for drawing interesting conclusions. Over the timescale of our experiments, our results do not provide any data on stress corrosion, which should be considered in further study.

The thermal conductivity of mixed fuel U xPu 1-xO 2: molecular dynamics simulations

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

Liu, Xiang-Yang; Cooper, Michael William Donald; Stanek, Christopher Richard

2015-10-16

Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO 2-PuO 2 system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO 2 fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. Formore » this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO 2. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of U xPu 1-xO 2, as a function of PuO 2 concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.« less

Blast biology: a study of the primary and tertiary effects of blast in open underground protective shelters. Project 33. 1 of Operation Plumbbob

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

Ricmond, D.R.; Taborelli, R.V.; Bowen, I.G.

1959-02-01

Dogs, pigs, rabbits, guinea pigs, and mice were exposed to nuclear detonations in two open underground partitioned shelters. The shelters were of similar construction, and each was exposed to separate detonations. Each inner chamber filled through its own orifice; thus four separate pressure environments were obtained. An aerodynamic mound was placed over the escape hatch of each structure to determine its effect on the pressure-curve shape inside the chamber. In one test a sieve plate bolted across the top of the mound was evaluated. Wind protective baffles of solid plate and of heavy wire screen were installed in the sheltersmore » to compare primary and tertiary blast effects on dogs. The shelters also contained static and dynamic pressure gages, radiation detectors, telemetering devices, and, in one test, air-temperature measuring instruments, dust-collecting trays, and eight pigs for the biological assessment of thermal effects. One dog was severely injured from tertiary blast effects associated with a maximal dynamic pressure (Q) of 10.5 psi, and one was undamaged with a maximal Q of 2 psi. Primary blast effects resulting from peak overpressures of 30.3, 25.5, 9.5, and 4.1 psi were minimal. The mortality was 19% of the mice exposed to a peak pressure of 30.3 psi and 5 and 3% of the guinea pigs and mice exposed to a peak pressure of 25.5 psi. Many of the rabbits, guinea pigs, and mice sustained slight lung hemorrhages at maximum pressues of 25.5 and 30.3 psi. Eardrum perforation data for all species, except mice, were recorded. Following shot 2, thermal effects were noted. Animals of the groups saved for observation have died from ionizing-radiation effects.« less

BLAST BIOLOGY--A STUDY OF THE PRIMARY AND TERTIARY EFFECTS OF BLAST IN OPEN UNDERGROUND PROTECTIVE SHELTERS

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

Ricmond, D.R.; Taborelli, R.V.; Bowen, I.G.

1959-02-01

Dogs, pigs, rabbits, guinea pigs, and mice were exposed to nuclear detonatiors in two open underground pantitioned shelters. The shelters were of similar constructions and each was exposed to separate detonations. Each inner chamber filled through its own orifice; thus four separate pressure enviromments were obtained. An aerodynamic mound was placed over the escape hatch of each structure to determine its effect on the pressurecurve shape inside the chamber. In one test a sieve plate bolted across the top of the mound was evaluated. Wind protective baffles of solid plate and of heavy wire screen were installed in the sheltersmore » to compare primary and tertiary blast effects on dogs. The shelters also contained static and dynamic pressure gages, radiation detectors, telemetering devices, and, in one test, air-temperature measuring instruments, dustcollecting trays, and eight pigs for the biological assessment of thermal effects. One dog was severely injured from tertiary blast effects associated with a maximal dynamic pressure (Q) of 10.5 psi, and one was undamaged with a maximal Q of 2 psi. Primary blast effects resulting from peak overpressures of 30.3, 25.5, 9.5. and 4.1 psi were minimal. The mortality was 19 per cent of the mice exposed to a peak pressure of 30.3 psi and 5 and 3 per cent of the guinea pigs and mice exposed to a peak pressure of 25.5 psi. Many of the rabbits, guinea pigs, and mice sustained slight lung hemorrhages at maximum pressures of 25.5 and 30.3 psi. Eardrum perforation data for all species, except mice, were recorded. Following shot 2, thermal effects were noted. Animals of the groups saved for observation have died from ionizing-radiation effects. (auth)« less

Thermal properties of nuclear matter in a variational framework with relativistic corrections

NASA Astrophysics Data System (ADS)

Zaryouni, S.; Hassani, M.; Moshfegh, H. R.

2014-01-01

The properties of hot symmetric nuclear matter for a wide range of densities and temperatures are investigated by employing the AV14 potential within the lowest order constrained variational (LOCV) method with the inclusion of a phenomenological three-body force as well as relativistic corrections. The relativistic corrections of many-body kinetic energies as well as the boot interaction corrections are presented for a wide range of densities and temperatures. The free energy, pressure, incompressibility, and other thermodynamic quantities of symmetric nuclear matter are obtained and discussed. The critical temperature is found, and the liquid-gas phase transition is analyzed both with and without the inclusion of three-body forces and relativistic corrections in the LOCV approach. It is shown that the critical temperature is strongly affected by the three-body forces but does not depend on the relativistic corrections. Finally, the results obtained in the present study are compared with other many-body calculations and experimental predictions.

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

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

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

All-optical technique for measuring thermal properties of materials at static high pressure

NASA Astrophysics Data System (ADS)

Pangilinan, G. I.; Ladouceur, H. D.; Russell, T. P.

2000-10-01

The development and implementation of an all-optical technique for measuring thermal transport properties of materials at high pressure in a gem anvil cell are reported. Thermal transport properties are determined by propagating a thermal wave in a material subjected to high pressures, and measuring the temperature as a function of time using an optical sensor embedded downstream in the material. Optical beams are used to deposit energy and to measure the sensor temperature and replace the resistive heat source and the thermocouples of previous methods. This overcomes the problems introduced with pressure-induced resistance changes and the spatial limitations inherent in previous high-pressure experimentation. Consistent with the heat conduction equation, the material's specific heat, thermal conductivity, and thermal diffusivity (κ) determine the sensor's temperature rise and its temporal profile. The all-optical technique described focuses on room-temperature thermal properties but can easily be applied to a wide temperature range (77-600 K). Measurements of thermal transport properties at pressure up to 2.0 GPa are reported, although extension to much higher pressures are feasible. The thermal properties of NaCl, a commonly used material for high-pressure experiments are measured and shown to be consistent with those obtained using the traditional methods.

Sport fishing at a thermal discharge into Lake Michigan

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

Spigarelli, S.A.; Thommes, M.M.

1976-07-01

Sport fishing censuses were conducted during 1972 and 1973 at the Point Beach Nuclear Plant on Lake Michigan (Two Rivers, Wisconsin). The objectives of this study were to describe the fishery at a typical shoreline thermal discharge into the upper Great Lakes and to make comparisons with reference fisheries in unheated areas. Extensive sport fishing at this power plant resulted in a relatively large catch of trout (4 species) and sporadic catches of salmon and non-salmonid species. Species composition of the catch and catch-per-unit-effort varied daily and seasonally and generally reflected trends in reference fisheries. A comparison between years showedmore » increased fishing effort, total catch, and proportion of trout in 1973, while success (catch-per-unit-effort) decreased. Despite this heavy fishing pressure, catch-per-unit-effort was generally higher at Point Beach than in reference shoreline fisheries. The economic value of thermal discharge fisheries on Lake Michigan is estimated using available value and expenditure data.« less

Impact Response of Thermally Sprayed Metal Deposits

NASA Astrophysics Data System (ADS)

Wise, J. L.; Hall, A. C.; Moore, N. W.; Pautz, S. D.; Franke, B. C.; Scherzinger, W. M.; Brown, D. W.

2017-06-01

Gas-gun experiments have probed the impact response of tantalum specimens that were additively manufactured using a controlled thermal spray deposition process. Velocity interferometer (VISAR) diagnostics provided time-resolved measurements of sample response under one-dimensional (i . e . , uniaxial strain) shock compression to peak stresses ranging between 1 and 4 GPa. The acquired wave-profile data have been analyzed to determine the Hugoniot Elastic Limit (HEL), Hugoniot equation of state, and high-pressure yield strength of the thermally deposited samples for comparison to published baseline results for conventionally wrought tantalum. The effects of composition, porosity, and microstructure (e . g . , grain/splat size and morphology) are assessed to explain differences in the dynamic mechanical behavior of spray-deposited versus conventional material. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

NASA Astrophysics Data System (ADS)

Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Nishiyama, Y.; Onizawa, K.

2011-08-01

The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% δ-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M 23C 6 type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

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

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Pressure-induced reversal between thermal contraction and expansion in ferroelectric PbTiO3.

PubMed

Zhu, Jinlong; Zhang, Jianzhong; Xu, Hongwu; Vogel, Sven C; Jin, Changqing; Frantti, Johannes; Zhao, Yusheng

2014-01-15

Materials with zero/near zero thermal expansion coefficients are technologically important for applications in thermal management and engineering. To date, this class of materials can only be produced by chemical routes, either by changing chemical compositions or by composting materials with positive and negative thermal expansion. Here, we report for the first time a physical route to achieve near zero thermal expansion through application of pressure. In the stability field of tetragonal PbTiO3 we observed pressure-induced reversals between thermal contraction and expansion between ambient pressure and 0.9 GPa. This hybrid behavior leads to a mathematically infinite number of crossover points in the pressure-volume-temperature space and near-zero thermal expansion coefficients comparable to or even smaller than those attained by chemical routes. The observed pressures for this unusual phenomenon are within a small range of 0.1-0.9 GPa, potentially feasible for designing stress-engineered materials, such as thin films and nano-crystals, for thermal management applications.

Coupled THMC models for bentonite in clay repository for nuclear waste

NASA Astrophysics Data System (ADS)

Zheng, L.; Rutqvist, J.; Birkholzer, J. T.; Li, Y.; Anguiano, H. H.

2015-12-01

Illitization, the transformation of smectite to illite, could compromise some beneficiary features of an engineered barrier system (EBS) that is composed primarily of bentonite and clay host rock. It is a major determining factor to establish the maximum design temperature of the repositories because it is believed that illitization could be greatly enhanced at temperatures higher than 100 oC and thus significantly lower the sorption and swelling capacity of bentonite and clay rock. However, existing experimental and modeling studies on the occurrence of illitization and related performance impacts are not conclusive, in part because the relevant couplings between the thermal, hydrological, chemical, and mechanical (THMC) processes have not been fully represented in the models. Here we present fully coupled THMC simulations of a generic nuclear waste repository in a clay formation with bentonite-backfilled EBS. Two scenarios were simulated for comparison: a case in which the temperature in the bentonite near the waste canister can reach about 200 oC and a case in which the temperature in the bentonite near the waste canister peaks at about 100 oC. The model simulations demonstrate that illitization is in general more significant at higher temperatures. We also compared the chemical changes and the resulting swelling stress change for two types of bentonite: Kunigel-VI and FEBEX bentonite. Higher temperatures also lead to much higher stress in the near field, caused by thermal pressurization and vapor pressure buildup in the EBS bentonite and clay host rock. Chemical changes lead to a reduction in swelling stress, which is more pronounced for Kunigel-VI bentonite than for FEBEX bentonite.

Nuclear Propulsion Technical Interchange Meeting, volume 2

NASA Technical Reports Server (NTRS)

1993-01-01

The purpose of the meeting was to review the work performed in fiscal year 1992 in the areas of nuclear thermal and nuclear electric propulsion technology development. These proceedings are an accumulation of the presentations provided at the meeting along with annotations provided by authors. The proceedings cover system concepts, technology development, and system modeling for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP). The test facilities required for the development of the nuclear propulsion systems are also discussed.

Study of Defect Levels in InAs/InAsSb Type-II Superlattice Using Pressure-Dependent Photoluminescence

DTIC Science & Technology

2015-07-07

pressure is confirmed from power dependent PL measurements. We also examined the thermal activation energies at ambient pressure and close to the...with pressure is confirmed from power dependent PL measurements. We also examined the thermal activation energies at ambient pressure and close to...We also examined the thermal activation energies at ambient pressure and close to the crossover pressure. These results support and are consistent

(3) He Spin Filter for Neutrons.

PubMed

Batz, M; Baeßler, S; Heil, W; Otten, E W; Rudersdorf, D; Schmiedeskamp, J; Sobolev, Y; Wolf, M

2005-01-01

The strongly spin-dependent absorption of neutrons in nuclear spin-polarized (3)He opens up the possibility of polarizing neutrons from reactors and spallation sources over the full kinematical range of cold, thermal and hot neutrons. This paper gives a report on the neutron spin filter (NSF) development program at Mainz. The polarization technique is based on direct optical pumping of metastable (3)He atoms combined with a polarization preserving mechanical compression of the gas up to a pressure of several bar, necessary to run a NSF. The concept of a remote type of operation using detachable NSF cells is presented which requires long nuclear spin relaxation times of order 100 hours. A short survey of their use under experimental conditions, e.g. large solid-angle polarization analysis, is given. In neutron particle physics NSFs are used in precision measurements to test fundamental symmetry concepts.

A performance comparison of nuclear electric and nuclear thermal propulsion for Mars cargo missions across the 15-17 year synodic cycle

NASA Technical Reports Server (NTRS)

Sponaugle, Steven J.; Davis, Steven F.; Everett, Shonn F.

1992-01-01

This paper examines the effects of the Earth-Mars synodic cycle on Mars cargo missions. Cargo vehicles that use nuclear thermal propulsion are compared with those that use nuclear electric propulsion. It will be shown that for low energy class cargo missions, nuclear electric systems exhibit far less variation in peak performance over the synodic cycle than comparable nuclear thermal systems. Performance is measured by the amount of usable mass delivered to Mars, as well as the initial mass requirements in nuclear safe orbit. Nuclear electric propulsion systems also have significantly longer injection window opportunities for a given 26 month synodic period, resulting in much greater mission design flexibility. Injection window opportunities over a 20 year period from 2010 to 2030 are examined. This covers a complete synodic cycle and shows its effects on performance for Mars cargo missions.

Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

NASA Technical Reports Server (NTRS)

Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2017-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

DOEpatents

Bowman, C.D.

1992-11-03

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

DOEpatents

Bowman, Charles D.

1992-01-01

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Experiment of monitoring thermal discharge drained from nuclear plant through airborne infrared remote sensing

NASA Astrophysics Data System (ADS)

Wang, Difeng; Pan, Delu; Li, Ning

2009-07-01

The State Development and Planning Commission has approved nuclear power projects with the total capacity of 23,000 MW. The plants will be built in Zhejiang, Jiangsu, Guangdong, Shandong, Liaoning and Fujian Province before 2020. However, along with the nuclear power policy of accelerated development in our country, the quantity of nuclear plants and machine sets increases quickly. As a result the environment influence of thermal discharge will be a problem that can't be slid over. So evaluation of the environment influence and engineering simulation must be performed before station design and construction. Further more real-time monitoring of water temperature need to be arranged after fulfillment, reflecting variety of water temperature in time and provided to related managing department. Which will help to ensure the operation of nuclear plant would not result in excess environment breakage. At the end of 2007, an airborne thermal discharge monitoring experiment has been carried out by making use of MAMS, a marine multi-spectral scanner equipped on the China Marine Surveillance Force airplane. And experimental subject was sea area near Qin Shan nuclear plant. This paper introduces the related specification and function of MAMS instrument, and decrypts design and process of the airborne remote sensing experiment. Experiment showed that applying MAMS to monitoring thermal discharge is viable. The remote sensing on a base of thermal infrared monitoring technique told us that thermal discharge of Qin Shan nuclear plant was controlled in a small scope, never breaching national water quality standard.

75 FR 10410 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-08

... Toughness Requirements for Protection Against Pressurized Thermal Shock Events; Correcting Amendment AGENCY... Commission (NRC) is revising its regulations to add a table that was inadvertently omitted in a correction... toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water...

Hybrid Geo-Energy Systems for Energy Storage and Dispatchable Renewable and Low-Carbon Electricity

NASA Astrophysics Data System (ADS)

Buscheck, Thomas; Bielicki, Jeffrey; Ogland-Hand, Jonathan; Hao, Yue; Sun, Yunwei; Randolph, Jimmy; Saar, Martin

2015-04-01

Three primary challenges for energy systems are to (1) reduce the amount of carbon dioxide (CO2) being emitted to the atmosphere, (2) increase the penetration of renewable energy technologies, and (3) reduce the water intensity of energy production. Integrating variable renewable energy sources (wind, sunlight) into electric grids requires advances in energy storage approaches, which are currently expensive, and tend to have limited capacity and/or geographic deployment potential. Our approach uses CO2, that would otherwise be emitted to the atmosphere, to generate electricity from geothermal resources, to store excess energy from variable (wind, solar photovoltaic) and thermal (nuclear, fossil, concentrated solar power) sources, and to thus enable increased penetration of renewable energy technologies. We take advantage of the enormous fluid and thermal storage capacity of the subsurface to harvest, store, and dispatch energy. Our approach uses permeable geologic formations that are vertically bounded by impermeable layers to constrain pressure and the migration of buoyant CO2 and heated brine. Supercritical CO2 captured from fossil power plants is injected into these formations as a cushion gas to store pressure (bulk energy), provide an heat efficient extraction fluid for efficient power conversion in Brayton Cycle turbines, and generate artesian flow of brine -- which can be used to cool power plants and/or pre-heated (thermal storage) prior to re-injection. Concentric rings of injection and production wells create a hydraulic divide to store pressure, CO2, and thermal energy. The system is pressurized and/or heated when power supply exceeds demand and depressurized when demand exceeds supply. Time-shifting the parasitic loads from pressurizing and injecting brine and CO2 provides bulk energy storage over days to months, whereas time-shifting thermal-energy supply provides dispatchable power and addresses seasonal mismatches between supply and demand. These conditions enable efficient fluid recirculation, heat extraction, power conversion, and add operational flexibility to dispatch electricity. Overall, the system can (a) levelize concentrating solar power, (b) mitigate variability of wind and solar power, (c) reduce water and carbon intensity of energy systems, (d) avoid wasting or curtailing high-capital cost, low-carbon energy resources and (e) allow low-carbon, base-load power to operate at full capacity. This work was performed under the auspices of the U.S. DOE by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344, and has been funded by the U.S. National Science Foundation Sustainable Energy Pathways Program (1230691) and the U.S. Department of Energy Geothermal Technologies Office (DE-FOA-0000336).

Pressurizer with a mechanically attached surge nozzle thermal sleeve

DOEpatents

Wepfer, Robert M

2014-03-25

A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle.

Electron and Nuclear Pressures in Electron-Nucleus Mixtures

NASA Astrophysics Data System (ADS)

Chihara, J.; Yamagiwa, M.

2007-12-01

For a solid metal with frozen nuclei, the density-functional theory provides a unique definition of the electron pressure in an electron-nucleus mixture, and the total pressure of this mixture is represented as the sum of the electron and nuclear pressures. This fact leads to definitions of the electron and nuclear pressures on the basis of the virial theorem in terms of the wall potentials confining the electrons and nuclei. These definitions take a general form applicable without use of the adiabatic approximation. In this situation, we show that Janak's definition of the electron pressure in terms of the nuclear virial term is inappropriate; a similar statement holds for the definition of the stress tensor in this mixture. It is also demonstrated that both the electron and nuclear pressures become zero individually for a metal in vacuum, in contrast to the conventional understanding, according to which zero pressure is realized as a result of a cancellation of the elect ron and nuclear pressures. On the basis of these facts, a simple equation of state for liquid metals is derived, and it is examined numerically for the case of liquid alkaline metals by use of the quantum hypernetted chain equation and the Ashcroft model potential.

Temperature dependence of thermal pressure for NaCl

NASA Astrophysics Data System (ADS)

Singh, Chandra K.; Pande, Brijesh K.; Pandey, Anjani K.

2018-05-01

Engineering applications of the materials can be explored upto the desired limit of accuracy with the better knowledge of its mechanical and thermal properties such as ductility, brittleness and Thermal Pressure. For the resistance to fracture (K) and plastic deformation (G) the ratio K/G is treated as an indication of ductile or brittle character of solids. In the present work we have tested the condition of ductility and brittleness with the calculated values of K/G for the NaCl. It is concluded that the nature of NaCl can be predicted upto high temperature simply with the knowledge of its elastic stiffness constant only. Thermoelastic properties of materials at high temperature is directly related to thermal pressure and volume expansion of the materials. An expression for the temperature dependence of thermal pressure is formulated using basic thermodynamic identities. It is observed that thermal pressure ΔPth calculated for NaCl by using Kushwah formulation is in good agreement with the experimental values also the thermal pressure increases with the increase in temperature.

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

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

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, themore » NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.« less

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

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

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to 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. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, 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 relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

The Thermal Pressure in Low Metallicity Galaxies

NASA Astrophysics Data System (ADS)

Wolfire, Mark; McKee, Christopher; Ostriker, Eve C.; Bolatto, Alberto; Jenkins, Edward

2015-08-01

The thermal pressure in the diffuse interstellar medium (ISM) is a relatively small fraction of the total ISM pressure yet it is extremely important for the evolution of the ISM phases. A multi-phase medium can exist between a range of thermal pressures Pmin < Pth < Pmax. The phase separation is driven by thermal instability and produces a cold (T ˜ 100 K) neutral atomic gas and a warm (T ˜ 8000 K) neutral atomic gas separated by thermally unstable gas. At thermal pressures greater than Pmax only the cold phase can exist and at thermal pressures less than Pmin only the warm phase can exist. The ISM is also highly turbulent and turbulence can both initiate the thermal phase transition and be produced in a rapid phase transition. Hydrodynamic modeling also points to a strong two-phase distribution (.e.g., Kim et al. 2011; Audit & Hennebelle 2010) with a median thermal pressure in the cold gas very near the expected two-phase pressure. Global, theoretical models including star-formation feedback have been developed for the molecular fraction in galactic disks using, at their core, the paradigm that thermal pressure determines the phase transitions to warm, cold, or multiphase medium (e.g., Krumholz et al. 2009; Ostriker et al. 2010).Here we present a phase diagram for a low metallicity galaxy using the Small Magellanic Clouds as an example. We find that although the heating rates and metallicities can differ by factors of 5 to 10 from the Milky Way, the resulting two-phase pressure and physical conditions of the phases are not very different from Galactic. We also confirm that a widely used fitting function for Pmin presented in Wolfire et al. 2003 provides an accurate prediction for the new results. We demonstrate how the variation in input parameters determine the final pressures and physical conditions.

Shear Heating-Induced Thermal Pressurization During the Nucleation of Earthquakes

NASA Astrophysics Data System (ADS)

Schmitt, S. V.; Segall, P.

2008-12-01

Shear heating-induced thermal pressurization has long been posited as a weakening mechanism during earthquakes. It is often assumed that thermal pressurization does not become important until earthquakes become moderate to large in magnitude. Schmitt et al. [AGU, 2007] confirmed the estimate of Segall and Rice [JGR, 2006] that thermal pressurization becomes dominant during the quasi-static nucleation phase by conducting 2D numerical simulations that account for full thermomechanical coupling, with rate and state dependent friction. In that work, thermal pressurization becomes the dominant weakening mechanism at slip rates of 10-5 to 10-3 m/s, depending on the fault zone hydraulic diffusivity. Interestingly, the thermal pressurization process leads to a contraction of the nucleation zone, rather than the growing crack (aging law) or unidirectional slip pulse (slip law) associated with drained rate- and state-dependent frictional nucleation. The results of Schmitt et al. [AGU, 2007] had a shortcoming in that the principal slip surface was treated as a zero-width feature, while in reality it should be a finite-width shear zone. We address that shortcoming with a new set of numerical simulations. We assume a finite-width fault governed by rate and state friction with the radiation damping approximation to simulate inertial effects. Both thermal and hydraulic diffusion are computed via finite differences on separate, coupled grids that adaptively remesh to minimize computational expense while maintaining accuracy. New results suggest that the thermal pressurization effect is modestly reduced by including the finite thickness of the shear zone. Despite the reduction in the effect, the new results still indicate that (1) thermal pressurization is important before seismic slip and (2) thermal pressurization restricts growth of the nucleation zone.

Materials Coating Techniques

DTIC Science & Technology

1980-03-01

applications from decorative to utilitarian over significant segments of the engineering, chemical, nuclear , microelectronics, and related Industries. PVD...Thermal-control coating. Boron 2430 Cermet component, nuclear shielding and controlrod material; Carbide wear- and temperature-resistant. Calcium...Zirconium Oxide (Hafnia-Pree � Thermal-barrier coatings for nuclear applications. Lime Stabi!Aed) Zirconium 2563 Resistant to high-temperature

Research Technology

NASA Image and Video Library

1960-01-01

Originally investigated in the 1960's by Marshall Space Flight Center plarners as part of the Nuclear Energy for Rocket Vehicle Applications (NERVA) program, nuclear-thermal rocket propulsion has been more recently considered in spacecraft designs for interplanetary human exploration. This artist's concept illustrates a nuclear-thermal rocket with an aerobrake disk as it orbits Mars.

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

Lappa, Marcello, E-mail: marcello.lappa@strath.ac.uk

The relevance of non-equilibrium phenomena, nonlinear behavior, gravitational effects and fluid compressibility in a wide range of problems related to high-temperature gas-dynamics, especially in thermal, mechanical and nuclear engineering, calls for a concerted approach using the tools of the kinetic theory of gases, statistical physics, quantum mechanics, thermodynamics and mathematical modeling in synergy with advanced numerical strategies for the solution of the Navier–Stokes equations. The reason behind such a need is that in many instances of relevance in this field one witnesses a departure from canonical models and the resulting inadequacy of standard CFD approaches, especially those traditionally used tomore » deal with thermal (buoyancy) convection problems. Starting from microscopic considerations and typical concepts of molecular dynamics, passing through the Boltzmann equation and its known solutions, we show how it is possible to remove past assumptions and elaborate an algorithm capable of targeting the broadest range of applications. Moving beyond the Boussinesq approximation, the Sutherland law and the principle of energy equipartition, the resulting method allows most of the fluid properties (density, viscosity, thermal conductivity, heat capacity and diffusivity, etc.) to be derived in a rational and natural way while keeping empirical contamination to the minimum. Special attention is deserved as well to the well-known pressure issue. With the application of the socalled multiple pressure variables concept and a projection-like numerical approach, difficulties with such a term in the momentum equation are circumvented by allowing the hydrodynamic pressure to decouple from its thermodynamic counterpart. The final result is a flexible and modular framework that on the one hand is able to account for all the molecule (translational, rotational and vibrational) degrees of freedom and their effective excitation, and on the other hand can guarantee adequate interplay between molecular and macroscopic-level entities and processes. Performances are demonstrated by computing some incompressible and compressible benchmark test cases for thermal (gravitational) convection, which are then extended to the high-temperature regime taking advantage of the newly developed features.« less

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

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

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

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

Cooling apparatus with a resilient heat conducting member

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

Chainer, Timothy J.; Parida, Pritish R.; Schultz, Mark D.

2016-06-14

A cooling structure including a thermally conducting central element having a channel formed therein, the channel being configured for flow of cooling fluid there through, a first pressure plate, and a first thermally conductive resilient member disposed between the thermally conducting central element and the first pressure plate, wherein the first pressure plate, the first thermally conductive resilient member, and the thermally conducting central element form a first heat transfer path.

Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

NASA Astrophysics Data System (ADS)

Fekete, Balazs; Trampus, Peter

2015-09-01

The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

Effect of impregnation pressure and time on the porosity, structure and properties of polyacrylonitrile-fiber based carbon composites

NASA Astrophysics Data System (ADS)

Venugopalan, Ramani; Roy, Mainak; Thomas, Susy; Patra, A. K.; Sathiyamoorthy, D.; Tyagi, A. K.

2013-02-01

Carbon-carbon composites may find applications in critical parts of advanced nuclear reactors. A series of carbon-carbon composites were prepared using polyacrylonitrile (PAN) based carbon fibers. The materials were densified by impregnating two-dimensional (2D) preforms with liquid phenol formaldehyde resin at different pressures and for different periods of time and then carbonizing those by slowly heating at 1000 °C. Effects of the processing parameters on the structure of the composites were extensively studied. The study showed conclusively that open porosity decreased with increasing impregnation pressure, whereas impregnation time had lesser effect. Matrix-resin bonding also improved at higher pressure. d002 spacing decreased and ordering along c-axis increased with concomitant increase in sp2-carbon fraction at higher impregnation pressures. The fiber reinforced composites exhibited short range ordering of carbon atoms and satisfied structural conditions (d002 values) of amorphous carbon according to the turbostratic model for non-graphitic carbon materials. The composites had pellet-density of ˜85% of the theoretical value, low thermal expansion and negligible neutron-poisoning. They maintained structural integrity and retained disordered nature even on heat-treatment at ca. 1800 °C.

Nuclear Propulsion Technical Interchange Meeting, volume 1

NASA Technical Reports Server (NTRS)

1993-01-01

The Nuclear Propulsion Technical Interchange Meeting (NP-TIM-92) was sponsored and hosted by the Nuclear Propulsion Office at the NASA Lewis Research Center. The purpose of the meeting was to review the work performed in fiscal year 1992 in the areas of nuclear thermal and nuclear electric propulsion technology development. These proceedings are a compilation of the presentations given at the meeting (many of the papers are presented in outline or viewgraph form). Volume 1 covers the introductory presentations and the system concepts and technology developments related to nuclear thermal propulsion.

Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

NASA Astrophysics Data System (ADS)

Li, Bo-Shiuan

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission gas release. An optimum design is sought considering both thermal and mechanical models of this ceramic cladding with UO2 and advanced high density fuels.

Cryogenic Fluid Management Technology Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Caffrey, Jarvis; Hedayat, Ali; Stephens, Jonathan; Polsgrove, Robert

2015-01-01

The purpose of this paper is to investigate, facilitate a discussion and determine a path forward for technology development of cryogenic fluid management technology that is necessary for long duration deep space missions utilizing nuclear thermal propulsion systems. There are a number of challenges in managing cryogenic liquids that must be addressed before long durations missions into deep space, such as a trip to Mars can be successful. The leakage rate of hydrogen from pressure vessels, seals, lines and valves is a critical factor that must be controlled and minimized. For long duration missions, hydrogen leakage amounts to large increases in hydrogen and therefore vehicle mass. The size of a deep space vehicle, such as a mars transfer vehicle, must be kept small to control cost and the logistics of a multi launch, assembled in orbit vehicle. The boil off control of the cryogenic fluid is an additional obstacle to long duration missions. The boil off caused by heat absorption results in the growth of the propellant needs of the vehicle and therefore vehicle mass. This is a significant problem for a vehicle using nuclear (fission) propulsion systems. Radiation from the engines deposits large quantities of heat into the cryogenic fluid, greatly increasing boil off beyond that caused by environmental heat leakage. Addressing and resolving these challenges is critical to successful long duration space exploration. This paper discusses the state of the technology needed to address these challenges and discuss the path forward needed in technology development.

Thermal properties of the Cobourg Limestone

NASA Astrophysics Data System (ADS)

Pitts, Michelle

The underground storage of used nuclear fuel in Deep Geologic Repositories (DGRs) has been a subject of research in Canada for decades. One important technical aspect of repository design is the accommodation of the mechanical impacts of thermal inputs (heating) from the fuel as it goes through the remainder of its life cycle. Placement room spacing, a major factor in project cost, will be determined by the ability of the host rock to dissipate heat. The thermal conductivity and linear thermal expansion will determine the evolution of the temperature and thermally-induced stress fields. Thermal processes must be well understood to design a successful DGR. This thesis examines the thermal properties of rocks, how they are influenced by factors such as temperature, pressure, mineralogy, porosity, and saturation; and common methods for calculating and/or measuring these properties. A brief overview of thermal and thermally-coupled processes in the context of DGRs demonstrates the degree to which they would impact design, construction, and operation of these critical structures. Several case histories of major in situ heating experiments are reviewed to determine how the lessons learned could be applied to a Canadian Underground Demonstration Facility (UDF). A mineralogy investigation using X-Ray Diffraction (XRD) and Scanning Electron Microscopy (SEM) examines samples of the Cobourg Limestone from the Bowmanville and Bruce sites, and demonstrates geographical variability within the Cobourg Formation. The thermal properties of samples from the Bowmanville site are determined. A divided bar apparatus was constructed and used to measure thermal conductivity. The temperature measurement component of the divided bar apparatus was used to measure linear thermal expansion. Finally, the past investigations into the thermal impact of a DGR are reviewed, and the implications of the laboratory testing results on similar analyses are discussed.

Evaluation of High-Performance Space Nuclear Electric Generators for Electric Propulsion Application

NASA Technical Reports Server (NTRS)

Woodcock, Gordon; Kross, Dennis A. (Technical Monitor)

2002-01-01

Electric propulsion applications are enhanced by high power-to-mass ratios for their electric power sources. At multi-megawatt levels, we can expect thrust production systems to be less than 5 kg/kWe. Application of nuclear electric propulsion to human Mars missions becomes an attractive alternative to nuclear thermal propulsion if the propulsion system is less than about 10 kg/kWe. Recent references have projected megawatt-plus nuclear electric sources at specific mass values from less than 1 kg/kWe to about 5 kg/kWe. Various assumptions are made regarding power generation cycle (turbogenerator; MHD (magnetohydrodynamics)) and reactor heat source design. The present paper compares heat source and power generation options on the basis of a parametric model that emphasizes heat transfer design and realizable hardware concept. Pressure drop (important!) is included in the power cycle analysis, and MHD and turbogenerator cycles are compared. Results indicate that power source specific mass less than 5 kg/kWe is attainable, even if peak temperatures achievable are limited to 1500 K. Projections of specific mass less than 1 kg/kWe are unrealistic, even at the highest peak temperatures considered.

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

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

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Test Plan for the Boiling Water Reactor Dry Cask Simulator

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

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing themore » internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.« less

The Thermal Conductivity of Granular Materials as a Function of Grain Size Distribution and Gas Pressure

NASA Astrophysics Data System (ADS)

Hütter, Erika S.; Kömle, Norbert I.

2007-08-01

Many planetary bodies - in particular those with no or thin atmospheres - are covered by so-called regolith layers which usually constitute the uppermost metres of their surfaces. Examples are the Moon, the icy satellites of the outer solar system, asteroids and comets. The thermal conductivity of these surface layers controls to a high extent the energy balance of the body as a whole. Under low pressure conditions the effective thermal conductivity of granular materials is known to be very low, because the mutual contact area contact between individual particles is small. Therefore regolith surface layers are acting as thermal insulators. Up to now only a few thermal conductivity measurements in an extraterrestrial environment have been carried out, namely on the Moon in the frame of the Apollo Moon Lander missions. For the future several missions involving landers on asteroids, comets, and the Moon are planned by various space agencies. Thus the development of reliable instruments for the measurement of the thermal properties of regolith is of high interest. For this purpose thermal conductivity measurements with various regolith analogue materials under low pressure conditions need to be done. In order to contribute to this goal, we have performed a series of experiments using glass beads with various size distributions as analogue materials. To sort out the influence of the environmental gas pressure on the effective thermal conductivity each sample was embedded into a nitrogen atmosphere and the pressure was systematically varied from 10-4mbar (high vacuum range) up to 1 bar. The grain sizes used for the glass spheres were in the range from 0.1 mm to 4.3 mm. Additionally a mixture of different grain sizes was analysed. We report on the results of thermal conductivity measurements obtained for the different size fractions as a function of gas pressure. Our results indicate a strong influence of both the gas pressure and the grain size on the value of the thermal conductivity of the glass beads samples. For all cases measured a decrease of the pressure led to a corresponding decrease of the thermal conductivity. In the high vacuum conditions it was found to be approximately 30 times smaller than under normal atmospheric pressure. The strongest decay occurs in the pressure range from 102 down to 10-1mbar. At lower pressures no significant dependence of the thermal conductivity on the gas pressure was observed. The relation between the used grain sizes and the thermal conductivity was found to be linear.

Non-thermal Plasma Activates Human Keratinocytes by Stimulation of Antioxidant and Phase II Pathways

PubMed Central

Schmidt, Anke; Dietrich, Stephan; Steuer, Anna; Weltmann, Klaus-Dieter; von Woedtke, Thomas; Masur, Kai; Wende, Kristian

2015-01-01

Non-thermal atmospheric pressure plasma provides a novel therapeutic opportunity to control redox-based processes, e.g. wound healing, cancer, and inflammatory diseases. By spatial and time-resolved delivery of reactive oxygen and nitrogen species, it allows stimulation or inhibition of cellular processes in biological systems. Our data show that both gene and protein expression is highly affected by non-thermal plasma. Nuclear factor erythroid-related factor 2 (NRF2) and phase II enzyme pathway components were found to act as key controllers orchestrating the cellular response in keratinocytes. Additionally, glutathione metabolism, which is a marker for NRF2-related signaling events, was affected. Among the most robustly increased genes and proteins, heme oxygenase 1, NADPH-quinone oxidoreductase 1, and growth factors were found. The roles of NRF2 targets, investigated by siRNA silencing, revealed that NRF2 acts as an important switch for sensing oxidative stress events. Moreover, the influence of non-thermal plasma on the NRF2 pathway prepares cells against exogenic noxae and increases their resilience against oxidative species. Via paracrine mechanisms, distant cells benefit from cell-cell communication. The finding that non-thermal plasma triggers hormesis-like processes in keratinocytes facilitates the understanding of plasma-tissue interaction and its clinical application. PMID:25589789

New Laboratory Observations of Thermal Pressurization Weakening

NASA Astrophysics Data System (ADS)

Badt, N.; Tullis, T. E.; Hirth, G.

2017-12-01

Dynamic frictional weakening due to pore fluid thermal pressurization has been studied under elevated confining pressure in the laboratory, using a rotary-shear apparatus having a sample with independent pore pressure and confining pressure systems. Thermal pressurization is directly controlled by the permeability of the rocks, not only for the initiation of high-speed frictional weakening but also for a subsequent sequence of high-speed sliding events. First, the permeability is evaluated at different effective pressures using a method where the pore pressure drop and the flow-through rate are compared using Darcy's Law as well as a pore fluid oscillation method, the latter method also permitting measurement of the storage capacity. Then, the samples undergo a series of high-speed frictional sliding segments at a velocity of 2.5 mm/s, under an applied confining pressure and normal stress of 45 MPa and 50 MPa, respectively, and an initial pore pressure of 25 MPa. Finally the rock permeability and storage capacity are measured again to assess the evolution of the rock's pore fluid properties. For samples with a permeability of 10-20 m2 thermal pressurization promotes a 40% decrease in strength. However, after a sequence of three high-speed sliding events, the magnitude of weakening diminishes progressively from 40% to 15%. The weakening events coincide with dilation of the sliding interface. Moreover, the decrease in the weakening degree with progressive fast-slip events suggest that the hydraulic diffusivity may increase locally near the sliding interface during thermal pressurization-enhanced slip. This could result from stress- or thermally-induced damage to the host rock, which would perhaps increase both permeability and storage capacity, and so possibly decrease the susceptibility of dynamic weakening due to thermal pressurization in subsequent high-speed sliding events.

Behavior of a tapered hub flange with a bolted flat cover in transient temperature field

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

Sawa, T.; Nakagomi, Y.; Hirose, T.

1996-02-01

When bolted flange connections with gaskets are used in mechanical structures such as pipe connections, bolted covers of casks, and pressure vessels in nuclear and chemical plants and cylinder heads in internal combustion engines, they are usually subjected to transient thermal conditions. An experimental and analytical study was made on a bolted connection subjected to thermal loading. The connection consists of an aluminum alloy tapered hub flange and a flat cover, including a gasket fastened by steel bolts and nuts. Temperature distribution in the connection was measured with thermocouples, and the axial bolt force, the maximum bolt stress, and themore » hub stress were measured by strain gages under a thermal condition that the inner surface of the flanges was heated and the outer surfaces of the flanges and the cover were held at room temperature. Finite difference analysis was made to obtain the temperature distributions in the connection due to a transient thermal condition. This paper demonstrates the method for obtaining an increment in axial bolt force and the maximum bolt stress. In all cases, the analytical results were fairly consistent with the experimental results.« less

Nuclear factor-kappa B decoy suppresses nerve injury and improves mechanical allodynia and thermal hyperalgesia in a rat lumbar disc herniation model.

PubMed

Suzuki, Munetaka; Inoue, Gen; Gemba, Takefumi; Watanabe, Tomoko; Ito, Toshinori; Koshi, Takana; Yamauchi, Kazuyo; Yamashita, Masaomi; Orita, Sumihisa; Eguchi, Yawara; Ochiai, Nobuyasu; Kishida, Shunji; Takaso, Masashi; Aoki, Yasuchika; Takahashi, Kazuhisa; Ohtori, Seiji

2009-07-01

Nuclear factor-kappa B (NF-kappaB) is a gene transcriptional regulator of inflammatory cytokines. We investigated the transduction efficiency of NF-kappaB decoy to dorsal root ganglion (DRG), as well as the decrease in nerve injury, mechanical allodynia, and thermal hyperalgesia in a rat lumbar disc herniation model. Forty rats were used in this study. NF-kappaB decoy-fluorescein isothiocyanate (FITC) was injected intrathecally at the L5 level in five rats, and its transduction efficiency into DRG measured. In another 30 rats, mechanical pressure was placed on the DRG at the L5 level and nucleus pulposus harvested from the rat coccygeal disc was transplanted on the DRG. Rats were classified into three groups of ten animals each: a herniation + decoy group, a herniation + oligo group, and a herniation only group. For behavioral testing, mechanical allodynia and thermal hyperalgesia were evaluated. In 15 of the herniation rats, their left L5 DRGs were resected, and the expression of activating transcription factor 3 (ATF-3) and calcitonin gene-related peptide (CGRP) was evaluated immunohistochemically compared to five controls. The total transduction efficiency of NF-kappaB decoy-FITC in DRG neurons was 10.8% in vivo. The expression of CGRP and ATF-3 was significantly lower in the herniation + decoy group than in the other herniation groups. Mechanical allodynia and thermal hyperalgesia were significantly suppressed in the herniation + decoy group. NF-kappaB decoy was transduced into DRGs in vivo. NF-kappaB decoy may be useful as a target for clarifying the mechanism of sciatica caused by lumbar disc herniation.

The thermal and mechanical deformation study of up-stream pumping mechanical seal

NASA Astrophysics Data System (ADS)

Chen, H. L.; Xu, C.; Zuo, M. Z.; Wu, Q. B.

2015-01-01

Taking the viscosity-temperature relationship of the fluid film into consideration, a 3-D numerical model was established by ANSYS software which can simulate the heat transfer between the upstream pumping mechanical seal stationary and rotational rings and the fluid film between them as well as simulate the thermal deformation, structure deformation and the coupling deformation of them. According to the calculation result, thermal deformation causes the seal face expansion and the maximum thermal deformation appears at the inside of the seal ring. Pressure results in a mechanical deformation, the maximum deformation occurs at the top of the spiral groove and the overall trend is inward the mating face, opposite to the thermal deformation. The coupling deformation indicate that the thermal deformation can be partly counteracted by pressure deformation. Using this model, the relationship between deformation and shaft speed and the sealing liquid pressure was studied. It's found that the shaft speed will both enhance the thermal and structure deformation and the fluid pressure will enhance the structure deformation but has little to do with the thermal deformation. By changing the sealing material, it's found that material with low thermal expansion coefficient and low elastic modulus will suffer less thermal-pressure deformation.

Pronounced low-frequency vibrational thermal transport in C60 fullerite realized through pressure-dependent molecular dynamics simulations

NASA Astrophysics Data System (ADS)

Giri, Ashutosh; Hopkins, Patrick E.

2017-12-01

Fullerene condensed-matter solids can possess thermal conductivities below their minimum glassy limit while theorized to be stiffer than diamond when crystallized under pressure. These seemingly disparate extremes in thermal and mechanical properties raise questions into the pressure dependence on the thermal conductivity of C60 fullerite crystals, and how the spectral contributions to vibrational thermal conductivity changes under applied pressure. To answer these questions, we investigate the effect of strain on the thermal conductivity of C60 fullerite crystals via pressure-dependent molecular dynamics simulations under the Green-Kubo formalism. We show that the thermal conductivity increases rapidly with compressive strain, which demonstrates a power-law relationship similar to their stress-strain relationship for the C60 crystals. Calculations of the density of states for the crystals under compressive strains reveal that the librational modes characteristic in the unstrained case are diminished due to densification of the molecular crystal. Over a large compression range (0-20 GPa), the Leibfried-Schlömann equation is shown to adequately describe the pressure dependence of thermal conductivity, suggesting that low-frequency intermolecular vibrations dictate heat flow in the C60 crystals. A spectral decomposition of the thermal conductivity supports this hypothesis.

Experience gained in France on heat recovery from nuclear plants for agriculture and pisciculture

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

Balligand, P.; Le Gouellec, P.; Dumont, M.

1978-04-01

Since 1972, the Commissariat a l'Energie Atomique, Electricite de France, and the French Ministry of Agriculture have jointly examined the possibility of using thermal wastes from nuclear power plants for the benefit of agricultural production. A new process to heat greenhouses with water at 303 K using a double-wall plastic mulching laid directly on the soil has been successfully used for a few years on several hectares. When necessary, heat pumps are utilized. Very good results have been obtained for tomatoes, cucumbers, flowers, and strawberries, etc. Outdoor soil heating with buried pipes has been tested in Cadarache near an experimentalmore » pressurized water reactor for market garden crops and forestry. Gains in precocity and yield have been excellent, especially for asparagus, strawberries, and potatoes. Growing of eels has been four times faster in warm water over one year.« less

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.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

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

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

3He Spin Filter for Neutrons

PubMed Central

Batz, M.; Baeßler, S.; Heil, W.; Otten, E. W.; Rudersdorf, D.; Schmiedeskamp, J.; Sobolev, Y.; Wolf, M.

2005-01-01

The strongly spin-dependent absorption of neutrons in nuclear spin-polarized 3He opens up the possibility of polarizing neutrons from reactors and spallation sources over the full kinematical range of cold, thermal and hot neutrons. This paper gives a report on the neutron spin filter (NSF) development program at Mainz. The polarization technique is based on direct optical pumping of metastable 3He atoms combined with a polarization preserving mechanical compression of the gas up to a pressure of several bar, necessary to run a NSF. The concept of a remote type of operation using detachable NSF cells is presented which requires long nuclear spin relaxation times of order 100 hours. A short survey of their use under experimental conditions, e.g. large solid-angle polarization analysis, is given. In neutron particle physics NSFs are used in precision measurements to test fundamental symmetry concepts. PMID:27308139

Active cooling-based surface confinement system for thermal soil treatment

DOEpatents

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

1997-10-28

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

Active cooling-based surface confinement system for thermal soil treatment

DOEpatents

Aines, Roger D.; Newmark, Robin L.

1997-01-01

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

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

2003-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

Nuclear thermal propulsion technology: Results of an interagency panel in FY 1991

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcdaniel, Patrick; Howe, Steven; Helms, Ira; Stanley, Marland

1993-01-01

NASA LeRC was selected to lead nuclear propulsion technology development for NASA. Also participating in the project are NASA MSFC and JPL. The U.S. Department of Energy will develop nuclear technology and will conduct nuclear component, subsystem, and system testing at appropriate DOE test facilities. NASA program management is the responsibility of NASA/RP. The project includes both nuclear electric propulsion (NEP) and nuclear thermal propulsion (NTP) technology development. This report summarizes the efforts of an interagency panel that evaluated NTP technology in 1991. Other panels were also at work in 1991 on other aspects of nuclear propulsion, and the six panels worked closely together. The charters for the other panels and some of their results are also discussed. Important collaborative efforts with other panels are highlighted. The interagency (NASA/DOE/DOD) NTP Technology Panel worked in 1991 to evaluate nuclear thermal propulsion concepts on a consistent basis. Additionally, the panel worked to continue technology development project planning for a joint project in nuclear propulsion for the Space Exploration Initiative (SEI). Five meetings of the panel were held in 1991 to continue the planning for technology development of nuclear thermal propulsion systems. The state-of-the-art of the NTP technologies was reviewed in some detail. The major technologies identified were as follows: fuels, coatings, and other reactor technologies; materials; instrumentation, controls, health monitoring and management, and associated technologies; nozzles; and feed system technology, including turbopump assemblies.

Space Nuclear Thermal Propulsion (SNTP) Air Force facility

NASA Technical Reports Server (NTRS)

Beck, David F.

1993-01-01

The Space Nuclear Thermal Propulsion (SNTP) Program is an initiative within the US Air Force to acquire and validate advanced technologies that could be used to sustain superior capabilities in the area or space nuclear propulsion. The SNTP Program has a specific objective of demonstrating the feasibility of the particle bed reactor (PBR) concept. The term PIPET refers to a project within the SNTP Program responsible for the design, development, construction, and operation of a test reactor facility, including all support systems, that is intended to resolve program technology issues and test goals. A nuclear test facility has been designed that meets SNTP Facility requirements. The design approach taken to meet SNTP requirements has resulted in a nuclear test facility that should encompass a wide range of nuclear thermal propulsion (NTP) test requirements that may be generated within other programs. The SNTP PIPET project is actively working with DOE and NASA to assess this possibility.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

2012-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

Pressure And Thermal Modeling Of Rocket Launches

NASA Technical Reports Server (NTRS)

Smith, Sheldon D.; Myruski, Brian L.; Farmer, Richard C.; Freeman, Jon A.

1995-01-01

Report presents mathematical model for use in designing rocket-launching stand. Predicts pressure and thermal environment, as well as thermal responses of structures to impinging rocket-exhaust plumes. Enables relatively inexperienced analyst to determine time-varying distributions and absolute levels of pressure and heat loads on structures.

Propulsion Research at the Propulsion Research Center of the NASA Marshall Space Flight Center

NASA Technical Reports Server (NTRS)

Blevins, John; Rodgers, Stephen

2003-01-01

The Propulsion Research Center of the NASA Marshall Space Flight Center is engaged in research activities aimed at providing the bases for fundamental advancement of a range of space propulsion technologies. There are four broad research themes. Advanced chemical propulsion studies focus on the detailed chemistry and transport processes for high-pressure combustion, and on the understanding and control of combustion stability. New high-energy propellant research ranges from theoretical prediction of new propellant properties through experimental characterization propellant performance, material interactions, aging properties, and ignition behavior. Another research area involves advanced nuclear electric propulsion with new robust and lightweight materials and with designs for advanced fuels. Nuclear electric propulsion systems are characterized using simulated nuclear systems, where the non-nuclear power source has the form and power input of a nuclear reactor. This permits detailed testing of nuclear propulsion systems in a non-nuclear environment. In-space propulsion research is focused primarily on high power plasma thruster work. New methods for achieving higher thrust in these devices are being studied theoretically and experimentally. Solar thermal propulsion research is also underway for in-space applications. The fourth of these research areas is advanced energetics. Specific research here includes the containment of ion clouds for extended periods. This is aimed at proving the concept of antimatter trapping and storage for use ultimately in propulsion applications. Another activity in this involves research into lightweight magnetic technology for space propulsion applications.

Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

PubMed

Huber, Patrick; Jaffke, Patrick

2016-03-25

We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

Embedded fiber Bragg grating pressure measurement during thermal ignition of a high explosive

DOE PAGES

Rodriguez, George; Smilowitz, Laura Beth; Henson, Bryan Fayne

2016-10-17

A high-speed fiber Bragg grating based pressure-only measurement is reported for the high explosive PBXN-9 under thermal initiation conditions. During exothermic thermal runaway, an explosion rise time of 500 μs reaching a peak pressure of 660 MPa is measured. Lastly, the approach offers a direct measure pressure diagnostic useful for quantifying reaction violence for high explosive chemistry.

Embedded fiber Bragg grating pressure measurement during thermal ignition of a high explosive

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

Rodriguez, George; Smilowitz, Laura Beth; Henson, Bryan Fayne

A high-speed fiber Bragg grating based pressure-only measurement is reported for the high explosive PBXN-9 under thermal initiation conditions. During exothermic thermal runaway, an explosion rise time of 500 μs reaching a peak pressure of 660 MPa is measured. Lastly, the approach offers a direct measure pressure diagnostic useful for quantifying reaction violence for high explosive chemistry.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

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

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Radiation Shielding for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Caffrey, Jarvis A.

2016-01-01

Design and analysis of radiation shielding for nuclear thermal propulsion has continued at Marshall Space Flight Center. A set of optimization tools are in development, and strategies for shielding optimization will be discussed. Considerations for the concurrent design of internal and external shielding are likely required for a mass optimal shield design. The task of reducing radiation dose to crew from a nuclear engine is considered to be less challenging than the task of thermal mitigation for cryogenic propellant, especially considering the likely implementation of additional crew shielding for protection from solar particles and cosmic rays. Further consideration is thus made for the thermal effects of radiation absorption in cryogenic propellant. Materials challenges and possible methods of manufacturing are also discussed.

Sequestration and disposal of dissolved Cs+ using zeolite 13X

NASA Astrophysics Data System (ADS)

Park, M.; Park, J.; Jeong, H. Y.

2017-12-01

Low-to-intermediate level liquid radioactive wastes (LILLW) typically contain high levels of radioactive 137Cs. Due to the great radiational and thermal stability as well as the high selectivity, zeolite has been commonly utilized to sequester radioactive isotopes from nuclear wastewater effluents. In this study, an Al-rich synthetic zeolite 13X was evaluated for the sorption capacity of Cs+ as a function of pH (4.0-10.5), ionic strength (0.05 and 0.2 M), and initial Cs+ concentration (1×10-6-5×10-3 M). For safe disposal, Cs+-exchanged 13X was both thermally and hydrothermally treated under different temperature and pressure. Subsequently, the resultant materials were examined for the phase transition by X-ray diffraction (XRD) and the local coordination chemistry by X-ray absorption spectroscopy (XAS). Our experimental results will detail the Cs+ sorption behavior by 13X under varying solution compositions. Also, the structural changes of Cs+-exchanged 13X upon thermal and hydrothermal treatment will be delineated to assess the stability of Cs+ in the treated materials.

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Development of the Variable Atmosphere Testing Facility for Blow-Down Analysis of the Mars Hopper Prototype

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

Nathan D. Jerred; Robert C. O'Brien; Steven D. Howe

Recent developments at the Center for Space Nuclear Research (CSNR) on a Martian exploration probe have lead to the assembly of a multi-functional variable atmosphere testing facility (VATF). The VATF has been assembled to perform transient blow-down analysis of a radioisotope thermal rocket (RTR) concept that has been proposed for the Mars Hopper; a long-lived, long-ranged mobile platform for the Martian surface. This study discusses the current state of the VATF as well as recent blow-down testing performed on a laboratory-scale prototype of the Mars Hopper. The VATF allows for the simulation of Mars ambient conditions within the pressure vesselmore » as well as to safely perform blow-down tests through the prototype using CO2 gas; the proposed propellant for the Mars Hopper. Empirical data gathered will lead to a better understanding of CO2 behavior and will provide validation of simulation models. Additionally, the potential of the VATF to test varying propulsion system designs has been recognized. In addition to being able to simulate varying atmospheres and blow-down gases for the RTR, it can be fitted to perform high temperature hydrogen testing of fuel elements for nuclear thermal propulsion.« less

MELCOR computer code manuals

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Thermal Recovery of Plastic Deformation in Dissimilar Metal Weld

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

Qiao, Dongxiao; Yu, Xinghua; Zhang, Wei

Stainless steel has been widely used in challenging environments typical to nuclear power plant structures, due its excellent corrosion resistance. Nickel filler metals containing high chromium concentration, including Alloy 82/182, are used for joining stainless steel to carbon steel components to achieve similar high resistance to stress corrosion cracking. However, the joint usually experience weld metal stress corrosion cracking (SCC), which affects the safety and structural integrity of light water nuclear reactor systems. A primary driving force for SCC is the high tensile residual stress in these welds. Due to large dimension of pressure vessel and limitations in the field,more » non-destructive residual stress measurement is difficult. As a result, finite element modeling has been the de facto method to evaluate the weld residual stresses. Recent studies on this subject from researchers worldwide report different residual stress value in the weldments [5]. The discrepancy is due to the fact that most of investigations ignore or underestimate the thermal recovery in the heat-affect zone or reheated region in the weld. In this paper, the effect of heat treatment on thermal recovery and microhardness is investigated for materials used in dissimilar metal joint. It is found that high equivalent plastic strains are predominately accumulated in the buttering layer, the root pass, and the heat affected zone, which experience multiple welding thermal cycles. The final cap passes, experiencing only one or two welding thermal cycles, exhibit less plastic strain accumulation. Moreover, the experimental residual plastic strains are compared with those predicted using an existing weld thermo-mechanical model with two different strain hardening rules. The importance of considering the dynamic strain hardening recovery due to high temperature exposure in welding is discussed for the accurate simulation of weld residual stresses and plastic strains. In conclsuion, the experimental result reveals that the typical post-buttering heat treatment for residual stress relief may not be adequate to completely eliminate the residual plastic strains in the buttering layer.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael

2009-01-01

A detailed description of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) is presented. The contents include: 1) Design Requirements; 2) NTREES Layout; 3) Data Acquisition and Control System Schematics; 4) NTREES System Schematic; and 5) NTREES Setup.

Relationship between pressure and reaction violence in thermal explosions

NASA Astrophysics Data System (ADS)

Smilowitz, L.; Henson, B. F.; Rodriguez, G.; Remelius, D.; Baca, E.; Oschwald, D.; Suvorova, N.

2017-01-01

Reaction violence of a thermal explosion is determined by the energy release rate of the explosive and the coupling of that energy to the case and surroundings. For the HMX and TATB based secondary high explosives studied, we have observed that temperature controls the time to explosion and pressure controls the final energy release rate subsequent to ignition. Pressure measurements in the thermal explosion regime have been notoriously difficult to make due to the extreme rise in temperature which is also occurring during a thermal explosion. We have utilized several different pressure measurement techniques for several different secondary high explosives. These techniques include commercially available piezoelectric and piezoresistive sensors which we have utilized in the low pressure (sub 30 MPa) range of PBX 9502 thermal explosions, and fiber Bragg grating sensors for the higher pressure range (up to GPa) for PBX9501 experiments. In this talk, we will compare the measurement techniques and discuss the pressures measured for the different formulations studied. Simultaneous x-ray radiography measurements of burn velocity will also be shown and correlations between pressure, burn velocity, and reaction violence will be discussed.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Thermal-neutron capture for A=26-35

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

Chunmei, Z.; Firestone, R.B.

2001-06-01

The prompt gamma-ray data of thermal- neutron captures fornuclear mass number A=26-35 had been evaluated and published in "ATOMICDATA AND NUCLEAR DATA TABLES, 26, 511 (1981)". Since that time themanyexperimental data of the thermal-neutron captures have been measuredand published. The update of the evaluated prompt gamma-ray data is verynecessary for use in PGAA of high-resolution analytical prompt gamma-rayspectroscopy. Besides, the evaluation is also very needed in theEvaluated Nuclear Structure Data File, ENSDF, because there are no promptgamma-ray data in ENSDF. The levels, prompt gamma-rays and decay schemesof thermal-neutron captures for nuclides (26Mg, 27Al, 28Si, 29Si, 30Si,31P, 32S, 33S, 34S, andmore » 35Cl) with nuclear mass number A=26-35 have beenevaluated on the basis of all experimental data. The normalizationfactors, from which absolute prompt gamma-ray intensity can be obtained,and necessary comments are given in the text. The ENSDF format has beenadopted in this evaluation. The physical check (intensity balance andenergy balance) of evaluated thermal-neutron capture data has been done.The evaluated data have been put into Evaluated Nuclear Structure DataFile, ENSDF. This evaluation may be considered as an update of the promptgamma-ray from thermal-neutron capture data tables as published in"ATOMIC DATA AND NUCLEAR DATA TABLES, 26, 511 (1981)".« less

Thermal-neutron capture for A=36-44

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

Chunmei, Z.; Firestone, R.B.

2003-01-01

The prompt gamma-ray data of thermal- neutron captures fornuclear mass number A=26-35 had been evaluated and published in "ATOMICDATA AND NUCLEAR DATA TABLES, 26, 511 (1981)". Since that time the manyexperimental data of the thermal-neutron captures have been measured andpublished. The update of the evaluated prompt gamma-ray data is verynecessary for use in PGAA of high-resolution analytical prompt gamma-rayspectroscopy. Besides, the evaluation is also very needed in theEvaluated Nuclear Structure Data File, ENSDF, because there are no promptgamma-ray data in ENSDF. The levels, prompt gamma-rays and decay schemesof thermal-neutron captures fornuclides (26Mg, 27Al, 28Si, 29Si, 30Si,31P, 32S, 33S, 34S, andmore » 35Cl) with nuclear mass number A=26-35 have beenevaluated on the basis of all experimental data. The normalizationfactors, from which absolute prompt gamma-ray intensity can be obtained,and necessary comments are given in the text. The ENSDF format has beenadopted in this evaluation. The physical check (intensity balance andenergy balance) of evaluated thermal-neutron capture data has been done.The evaluated data have been put into Evaluated Nuclear Structure DataFile, ENSDF. This evaluation may be considered as an update of the promptgamma-ray from thermal-neutron capture data tables as published in"ATOMIC DATA AND NUCLEAR DATA TABLES, 26, 511 (1981)".« less

Variable pressure thermal insulating jacket

DOEpatents

Nelson, Paul A.; Malecha, Richard F.; Chilenskas, Albert A.

1994-01-01

A device for controlled insulation of a thermal device. The device includes a thermal jacket with a closed volume able to be evacuated to form an insulating jacket around the thermal source. A getter material is in communcation with the closed volume of the thermal jacket. The getter material can absorb and desorb a control gas to control gas pressure in the volume of the thermal jacket to control thermal conductivity in the thermal jacket.

Gas Foil Bearing Technology Advancements for Closed Brayton Cycle Turbines

NASA Technical Reports Server (NTRS)

Howard, Samuel A.; Bruckner, Robert J.; DellaCorte, Christopher; Radil, Kevin C.

2007-01-01

Closed Brayton Cycle (CBC) turbine systems are under consideration for future space electric power generation. CBC turbines convert thermal energy from a nuclear reactor, or other heat source, to electrical power using a closed-loop cycle. The operating fluid in the closed-loop is commonly a high pressure inert gas mixture that cannot tolerate contamination. One source of potential contamination in a system such as this is the lubricant used in the turbomachine bearings. Gas Foil Bearings (GFB) represent a bearing technology that eliminates the possibility of contamination by using the working fluid as the lubricant. Thus, foil bearings are well suited to application in space power CBC turbine systems. NASA Glenn Research Center is actively researching GFB technology for use in these CBC power turbines. A power loss model has been developed, and the effects of a very high ambient pressure, start-up torque, and misalignment, have been observed and are reported here.

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

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

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

Methods of measuring water levels in deep wells

USGS Publications Warehouse

Garber, M.S.; Koopman, F. C.

1968-01-01

Accurate measurement of water levels deeper than 1,000 feet in wells requires specialized equipment. Corrections for stretch and thermal expansion of measuring tapes must be considered, and other measuring devices must be calibrated periodically. Bore-hole deviation corrections also must be made. Devices for recording fluctuation of fluid level usually require mechanical modification for use at these depths. A multichannel recording device utilizing pressure transducers has been constructed. This device was originally designed to record aquifer response to nearby underground nuclear explosions but can also be used for recording data from multi-well pumping tests. Bottom-hole recording devices designed for oil-field use have been utilized in a limited manner. These devices were generally found to lack the precision required, in ground-water investigations at the Nevada Test Site but may be applicable in other areas. A newly developed bottom-hole recording pressure gauge of improved accuracy has been used with satisfactory results.

Gas Phase Pressure Effects on the Apparent Thermal Conductivity of JSC-1A Lunar Regolith Simulant

NASA Technical Reports Server (NTRS)

Yuan, Zeng-Guang; Kleinhenz, Julie E.

2011-01-01

Gas phase pressure effects on the apparent thermal conductivity of a JSC-1A/air mixture have been experimentally investigated under steady state thermal conditions from 10 kPa to 100 kPa. The result showed that apparent thermal conductivity of the JSC-1A/air mixture decreased when pressure was lowered to 80 kPa. At 10 kPa, the conductivity decreased to 0.145 W/m/degree C, which is significantly lower than 0.196 W/m/degree C at 100 kPa. This finding is consistent with the results of previous researchers. The reduction of the apparent thermal conductivity at low pressures is ascribed to the Knudsen effect. Since the characteristic length of the void space in bulk JSC-1A varies over a wide range, both the Knudsen regime and continuum regime can coexist in the pore space. The volume ratio of the two regimes varies with pressure. Thus, as gas pressure decreases, the gas volume controlled by Knudsen regime increases. Under Knudsen regime the resistance to the heat flow is higher than that in the continuum regime, resulting in the observed pressure dependency of the apparent thermal conductivity.

Pressurized heat treatment of glass-ceramic to control thermal expansion

DOEpatents

Kramer, Daniel P.

1985-01-01

A method of producing a glass-ceramic having a specified thermal expansion value is disclosed. The method includes the step of pressurizing the parent glass material to a predetermined pressure during heat treatment so that the glass-ceramic produced has a specified thermal expansion value. Preferably, the glass-ceramic material is isostatically pressed. A method for forming a strong glass-ceramic to metal seal is also disclosed in which the glass-ceramic is fabricated to have a thermal expansion value equal to that of the metal. The determination of the thermal expansion value of a parent glass material placed in a high-temperature environment is also used to determine the pressure in the environment.

Thermal conductivity and thermal diffusivity of methane hydrate formed from compacted granular ice

NASA Astrophysics Data System (ADS)

Zhao, Jie; Sun, Shicai; Liu, Changling; Meng, Qingguo

2018-05-01

Thermal conductivity and thermal diffusivity of pure methane hydrate samples, formed from compacted granular ice (0-75 μm), and were measured simultaneously by the transient plane source (TPS) technique. The temperature dependence was measured between 263.15 and 283.05 K, and the gas-phase pressure dependence was measured between 2 and 10 MPa. It is revealed that the thermal conductivity of pure methane hydrate exhibits a positive trend with temperature and increases from 0.4877 to 0.5467 W·m-1·K-1. The thermal diffusivity of methane hydrate has inverse dependence on temperature and the values in the temperature range from 0.2940 to 0.3754 mm2·s-1, which is more than twice that of water. The experimental results show that the effects of gas-phase pressure on the thermal conductivity and thermal diffusivity are very small. Thermal conductivity of methane hydrate is found to have weakly positive gas-phase pressure dependence, whereas the thermal diffusivity has slightly negative trend with gas-phase pressure.

Effects of high hydrostatic pressure and thermal processing on bioactive compounds, antioxidant activity, and volatile profile of mulberry juice.

PubMed

Wang, Fan; Du, Bao-Lei; Cui, Zheng-Wei; Xu, Li-Ping; Li, Chun-Yang

2017-03-01

The aim of this study was to investigate the effects of high hydrostatic pressure and thermal processing on microbiological quality, bioactive compounds, antioxidant activity, and volatile profile of mulberry juice. High hydrostatic pressure processing at 500 MPa for 10 min reduced the total viable count from 4.38 log cfu/ml to nondetectable level and completely inactivated yeasts and molds in raw mulberry juice, ensuring the microbiological safety as thermal processing at 85 ℃ for 15 min. High hydrostatic pressure processing maintained significantly (p < 0.05) higher contents of total phenolic, total flavonoid and resveratrol, and antioxidant activity of mulberry juice than thermal processing. The main volatile compounds of mulberry juice were aldehydes, alcohols, and ketones. High hydrostatic pressure processing enhanced the volatile compound concentrations of mulberry juice while thermal processing reduced them in comparison with the control. These results suggested that high hydrostatic pressure processing could be an alternative to conventional thermal processing for production of high-quality mulberry juice.

Pressure dependence of thermal conductivity and specific heat in CeRh2Si2 measured by an extended thermal relaxation method

NASA Astrophysics Data System (ADS)

Nishigori, Shijo; Seida, Osamu

2018-05-01

We have developed a new technique for measuring thermal conductivity and specific heat under pressure by improving a thermal relaxation method. In this technique, a cylindrical sample with a small disc heater is embedded in the pressure-transmitting medium, then temperature variations of the sample and heater were directly measured by thermocouples during a heating and cooling process. Thermal conductivity and specific heat are estimated by comparing the experimental data with temperature variations simulated by a finite element method. The obtained thermal conductivity and specific heat of the test sample CeRh2Si2 exhibit a small enhancement and a clear peak arising from antiferromagnetic transition, respectively. The observation of these typical behaviors for magnetic compounds indicate that the technique is valid for the study on thermal properties under pressure.

Vaporisation of candidate nuclear fuels and targets for transmutation of minor actinides

NASA Astrophysics Data System (ADS)

Gotcu-Freis, P.; Hiernaut, J.-P.; Colle, J.-Y.; Nästrén, C.; Carretero, A. Fernandez; Konings, R. J. M.

2011-04-01

The thermal stability and high temperature behaviour of candidate fuels and targets for transmutation of minor actinides has been investigated. Zirconia-based solid solution, MgO-based CERCER and molybdenum-based CERMET fuels containing Am and/or Pu in various concentrations were heated up to 2700 K in a Knudsen cell coupled with a quadrupole mass spectrometer, to measure their vapour pressure and vapour composition. The results reveal that the vaporisation of the actinides from the samples is not only determined by the thermodynamics of the system but is also related to the dynamic evolution of multi-component mixtures with complex composition or microstructure.

Variable reluctance displacement transducer temperature compensated to 650$sup 0$F

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

Piper, T.C.; Colson, J.B.

650 deg F. Piper, T. C.; Colson, J. B. (Aerojet Nuclear Co., Idaho Falls, Idaho (USA)). Nov 1973. Contract AT(10-1)-1375. ' 16p. Dep. NTlS 00. A variable reluctnce transducer (VRT) with plus or minus 0.100- inch stroke to operate to 550 deg F is described. Specific design effort was taken to optimize the desiga to have minimum change in the VRT's null and sensitivity with rapid thermal and pressure transients. Null and sensitivity stabilities of plus or minus 0.5% and 1.8% of full scale, respectively, were obtained. Appentice; 'ineate the various effects considered in optimizing the design. (auth)

BLAST BIOLOGY. Technical Progress Report

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

White, C.S.; Richmond, D.R.

1959-09-18

Experimental data regarding the biologic consequences of exposure to several environmental variations associated with actual and simulated explosive detonations were reviewed. Blast biology is discussed relative to primary, secondary, tentiary, and miscellaneous blast effects as those attributable, respectively, to variations in environmental pressure, trauma from blast-produced missiles (both penetrating and nonpenetrating), the consequences of physical displacement of biological targets by blast-produced winds, and hazards due to ground shock, dust, and thermal phenomena not caused by thermal radiation per se. Primary blast effects were considered, noting physical-biophysical factors contributing to the observed pathophysiology. A simple hydrostatic model was utilized diagrammatically inmore » pointing out possible etiologic mechanisms. The gross biologic response to single. "fast"-rising overpressures were described as was the tolerance of mice, rats, guinea pigs. and rabbits to "long"-duration pressure pulses rising "rapidly" in single and double steps. Data regarding biological response to "slowly" rising over-pressures of "long" duration are discussed. Attention was called to the similarities under certain circumstances between thoracic trauma from nonpenetrating missiles and that noted from air blast. The association between air emboli, increase in lung weight (hemorrhage and edema), and mortality was discussed. Data relevant to the clinical symptoms and therapy of blast injury are presented. The relation of blast hazards to nuclear explosions was assessed and one approach to predicting the maximal potential casualties from blast phenomena is presented making use of arbitrary and tentative criteria. (auth)« less

Lattice vibrational contribution to equation of state for tetrahedral compounds

NASA Astrophysics Data System (ADS)

Kagaya, H.-Matsuo; Kotoku, H.; Soma, T.

1989-02-01

The lattice vibrational contributions to the Helmholtz free energy and the thermal pressure of tetrahedral compounds such as GaP, InP, ZnS, ZnSe, ZnTe and CdTe are investigated from the electronic theory of solids in the dynamical treatment based on our presented binding force. The temperature dependence of Helmholtz free energy and thermal pressure from lattice vibrational term are quantitatively obtained, and vibrational contributions to free energy are small compared with the static crystal energy. The influence of the thermal pressure is important to the equation of state in high temperatures, and the reformulation of the volume scale for the pressure-volume relation is given by considering the thermal pressure.

Cryogenic Fluid Management Technology and Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian D.; Caffrey, Jarvis; Hedayat, Ali; Stephens, Jonathan; Polsgrove, Robert

2016-01-01

Cryogenic fluid management (CFM) is critical to the success of future nuclear thermal propulsion powered vehicles. While this is an issue for any propulsion system utilizing cryogenic propellants, this is made more challenging by the radiation flux produced by the reactor in a nuclear thermal rocket (NTR). Managing the cryogenic fuel to prevent propellant loss to boil off and leakage is needed to limit the required quantity of propellant to a reasonable level. Analysis shows deposition of energy into liquid hydrogen fuel tanks in the vicinity of the nuclear thermal engine. This is on top of ambient environment sources of heat. Investments in cryogenic/thermal management systems (some of which are ongoing at various organizations) are needed in parallel to nuclear thermal engine development in order to one day see the successful operation of an entire stage. High durability, low thermal conductivity insulation is one developmental need. Light weight cryocoolers capable of removing heat from large fluid volumes at temperatures as low as approx. 20 K are needed to remove heat leak from the propellant of an NTR. Valve leakage is an additional CFM issue of great importance. Leakage rates of state of the art, launch vehicle size valves (which is approximately the size valves needed for a Mars transfer vehicle) are quite high and would result in large quantities of lost propellant over a long duration mission. Additionally, the liquid acquisition system inside the propellant tank must deliver properly conditioned propellant to the feed line for successful engine operation and avoid intake of warm or gaseous propellant. Analysis of the thermal environment and the CFM technology development are discussed in the accompanying presentation.

The Importance of Non-Thermal Pressures in the Heliosheath: Towards New Methods of Analysis

NASA Astrophysics Data System (ADS)

Roelof, E. C.; Gruntman, M.; Krimigis, S. M.; Mitchell, D. G.; McComas, D. J.; Funsten, H. O.

2009-12-01

The in-situ plasma measurements when Voyager 2 crossed the termination shock into the heliosheath revealed that only ~20% of the downstream pressure resided in the thermal ion population at energies << 1 keV. The LECP ion measurements > 30 keV at both VGRs 1 and 2 implied a partial pressure ΔP~0.02pPa that accounted for another ~15% of the total pressure. Adding in the missing 70% of the non-thermal pressure, the total non-thermal pressure at the VGRs must be P~0.12pPa. Consensus estimates of the local interstellar magnetic field (ISMF) are near B~0.25nT which gives a hydrostatic magnetic pressure B2/2μ0~0.25pPa. Cassini/INCA all-sky images of 5-44keV ENAs from the heliosheath [Krimigis et al., this session] show that neither VGR1 nor VGR2 is in the direction of maximum ENA emission. Consequently, it is possible that the pressure of non-thermal protons in the heliosheath is comparable to the hydrostatic pressure of the interstellar magnetic field (ISMF) that confines the heliosheath. An immediate corollary is that we will not understand the physics of the heliosheath until we find ways of quantitatively describing the dynamics of pressures produced by non-thermal ion populations. Present MHD theories and simulations simply do not capture these essential dynamical processes. We point out that the magnetospheric communities studying the dynamics of non-thermal ion injections (with plasma beta>1) at Earth and Saturn revealed by ENA imaging have been making significant progress in a quite similar problem. We offer some possible approaches for the quantitative analysis of the heliosheath, based on the magnetospheric experience.

Variable pressure thermal insulating jacket

DOEpatents

Nelson, P.A.; Malecha, R.F.; Chilenskas, A.A.

1994-09-20

A device for controlled insulation of a thermal device is disclosed. The device includes a thermal jacket with a closed volume able to be evacuated to form an insulating jacket around the thermal source. A getter material is in communication with the closed volume of the thermal jacket. The getter material can absorb and desorb a control gas to control gas pressure in the volume of the thermal jacket to control thermal conductivity in the thermal jacket. 10 figs.

The Economic Potential of Two Nuclear-Renewable Hybrid Energy Systems

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

Ruth, Mark; Cutler, Dylan; Flores-Espino, Francisco

Tightly coupled nuclear-renewable hybrid energy systems (N-R HESs) are an option that can generate zero-carbon, dispatchable electricity and provide zero-carbon energy for industrial processes at a lower cost than alternatives. N-R HESs are defined as systems that are managed by a single entity and link a nuclear reactor that generates heat, a thermal power cycle for heat to electricity conversion, at least one renewable energy source, and an industrial process that uses thermal and/or electrical energy. This report provides results of an analysis of two N-R HES scenarios. The first is a Texas-synthetic gasoline scenario that includes four subsystems: amore » nuclear reactor, thermal power cycle, wind power plant, and synthetic gasoline production technology. The second is an Arizona-desalination scenario with its four subsystems a nuclear reactor, thermal power cycle, solar photovoltaics, and a desalination plant. The analysis focuses on the economics of the N-R HESs and how they compare to other options, including configurations without all the subsystems in each N-R HES and alternatives where the energy is provided by natural gas.« less

ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

NASA Technical Reports Server (NTRS)

Walton, J. T.

1994-01-01

ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Thermal safety characterization on PETN, PBX-9407, LX-10-2, LX-17-1 and detonator in the LLNL's P-ODTX system

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

Hsu, P. C.; Strout, S.; Reynolds, J. G.

Incidents caused by fire and other thermal events can heat energetic materials that may lead to thermal explosion and result in structural damage and casualty. Thus, it is important to understand the response of energetic materials to thermal insults. The One-Dimensional-Time to Explosion (ODTX) system at the Lawrence Livermore National Laboratory (LLNL) has been used for decades to characterize thermal safety of energetic materials. In this study, an integration of a pressure monitoring element has been added into the ODTX system (P-ODTX) to perform thermal explosion (cook-off) experiments (thermal runaway) on PETN powder, PBX-9407, LX-10-2, LX-17-1, and detonator samples (cupmore » tests). The P-ODTX testing generates useful data (thermal explosion temperature, thermal explosion time, and gas pressures) to assist with the thermal safety assessment of relevant energetic materials and components. This report summarizes the results of P-ODTX experiments that were performed from May 2015 to July 2017. Recent upgrades to the data acquisition system allows for rapid pressure monitoring in microsecond intervals during thermal explosion. These pressure data are also included in the report.« less

ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

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

Wang, Jy-An John; Yan, Yong; Wang, Hong

A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydridesmore » in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation tests of hydrided Zircaloy-4 cladding, which served as a guideline to prepare in-cell hydride reorientation samples with high burnup HBR fuel segments. This report also provides the Phase II CIRFT test data for the hydride reorientation irradiated samples. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The CIRFT results appear to indicate that hydride reoriented treatment (HRT) have a negative effect on fatigue life, in addition to hydride reorientation effect. For HR4 specimen that had no pressurization procedure applied, the thermal annealing treatment alone showed a negative impact on the fatigue life compared to the HBR rod.« less

Effect of thermal pressurization on dynamic rupture propagation under depth-dependent stress

NASA Astrophysics Data System (ADS)

Urata, Y.; Kuge, K.; Kase, Y.

2009-12-01

Fluid and pore pressure evolution can affect dynamic propagation of earthquake ruptures owing to thermal pressurization (e.g., Mase and Smith, 1985). We investigate dynamic rupture propagation with thermal pressurization on a fault subjected to depth-dependent stress, on the basis of 3-D numerical simulations for spontaneous dynamic ruptures. We put a vertical strike-slip rectangular fault in a semi-infinite, homogenous, and elastic medium. The length and width of the fault are 8 and 3 km, respectively. We assume a depth-dependent stress estimated by Yamashita et al. (2004). The numerical algorithm is based on the finite-difference method by Kase and Kuge (2001). A rupture is initiated by increasing shear stress in a small patch at the bottom of the fault, and then proceeds spontaneously, governed by a slip-weakening law with the Coulomb failure criteria. Coefficients of friction and Dc are homogeneous on the fault. On a fault with thermal pressurization, we allow effective normal stress to vary with pore pressure change due to frictional heating by the formulation of Bizzarri and Cocco (2006). When thermal pressurization does not work, tractions drop in the same way everywhere and rupture velocity is subshear except near the free surface. Due to thermal pressurization, dynamic friction on the fault decreases and is heterogeneous not only vertically but horizontally, slip increases, and rupture velocity along the strike direction becomes supershear. As a result, plural peaks of final slip appear, as observed in the case of undrained dip-slip fault by Urata et al. (2008). We found in this study that the early stage of rupture growth under the depth-dependent stress is affected by the location of an initial crack. When a rupture is initiated at the center of the fault without thermal pressurization, the rupture cannot propagate and terminates. Thermal pressurization can help such a powerless rupture to keep propagating.

Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

NASA Astrophysics Data System (ADS)

Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

2014-06-01

In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

Studying the processes relating to oxidation of organic substances contained in the coolant of thermal and nuclear power stations

NASA Astrophysics Data System (ADS)

Khodyrev, B. N.; Krichevtsov, A. L.; Sokolyuk, A. A.

2010-07-01

A radical-chain mechanism governing thermal-oxidation destruction of organic substances contained in the coolant of thermal and nuclear power stations is considered. Hypotheses on the chemical nature of antioxidation properties of amines are presented. Theoretical conjectures about the fundamental processes through which protective amine films are formed on the surface of metals are suggested.

DOE`s annealing prototype demonstration projects

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

Warren, J.; Nakos, J.; Rochau, G.

1997-02-01

One of the challenges U.S. utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels (RPVs). As a nuclear plant operates, its RPV is exposed to neutrons. For certain plants, this neutron exposure can cause embrittlement of some of the RPV welds which can shorten the useful life of the RPV. This RPV embrittlement issue has the potential to affect the continued operation of a number of operating U.S. pressurized water reactor (PWR) plants. However, RPV material properties affected by long-term irradiation are recoverable throughmore » a thermal annealing treatment of the RPV. Although a dozen Russian-designed RPVs and several U.S. military vessels have been successfully annealed, U.S. utilities have stated that a successful annealing demonstration of a U.S. RPV is a prerequisite for annealing a licensed U.S. nuclear power plant. In May 1995, the Department of Energy`s Sandia National Laboratories awarded two cost-shared contracts to evaluate the feasibility of annealing U.S. licensed plants by conducting an anneal of an installed RPV using two different heating technologies. The contracts were awarded to the American Society of Mechanical Engineers (ASME) Center for Research and Technology Development (CRTD) and MPR Associates (MPR). The ASME team completed its annealing prototype demonstration in July 1996, using an indirect gas furnace at the uncompleted Public Service of Indiana`s Marble Hill nuclear power plant. The MPR team`s annealing prototype demonstration was scheduled to be completed in early 1997, using a direct heat electrical furnace at the uncompleted Consumers Power Company`s nuclear power plant at Midland, Michigan. This paper describes the Department`s annealing prototype demonstration goals and objectives; the tasks, deliverables, and results to date for each annealing prototype demonstration; and the remaining annealing technology challenges.« less

Simulation and test of the thermal behavior of pressure switch

NASA Astrophysics Data System (ADS)

Liu, Yifang; Chen, Daner; Zhang, Yao; Dai, Tingting

2018-04-01

Little, lightweight, low-power microelectromechanical system (MEMS) pressure switches offer a good development prospect for small, ultra-long, simple atmosphere environments. In order to realize MEMS pressure switch, it is necessary to solve one of the key technologies such as thermal robust optimization. The finite element simulation software is used to analyze the thermal behavior of the pressure switch and the deformation law of the pressure switch film under different temperature. The thermal stress releasing schemes are studied by changing the structure of fixed form and changing the thickness of the substrate, respectively. Finally, the design of the glass substrate thickness of 2.5 mm is used to ensure that the maximum equivalent stress is reduced to a quarter of the original value, only 154 MPa when the structure is in extreme temperature (80∘C). The test results show that after the pressure switch is thermally optimized, the upper and lower electrodes can be reliably contacted to accommodate different operating temperature environments.

Ascorbic acid is the only bioactive that is better preserved by high hydrostatic pressure than by thermal treatment of a vegetable beverage.

PubMed

Barba, Francisco J; Esteve, Maria J; Frigola, Ana

2010-09-22

Variations in levels of antioxidant compounds (ascorbic acid, total phenolics, and total carotenoids), total antioxidant capacity, and color changes in a vegetable (tomato, green pepper, green celery, onion, carrot, lemon, and olive oil) beverage treated by high hydrostatic pressure (HHP) were evaluated in this work. The effects of HHP treatment, four different pressures (100, 200, 300, and 400 MPa) and four treatment times for each pressure (from 120 to 540 s) were compared with those of thermal treatment (90-98 °C for 15 and 21 s). High pressure treatment retained significantly more ascorbic acid in the vegetable beverage than thermal treatment. However, no significant changes in total phenolics were observed between HHP treated and thermally processed vegetable beverage and unprocessed beverage. Color changes (a*, b*, L, chroma, h°, and ΔE) were less for pressurized beverage than thermally treated samples compared with unprocessed beverage.

Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

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

Aakre, Shaun R.; Jentz, Ian W.; Anderson, Mark H.

The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminarymore » observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO 2 systems, which commonly employ printed-circuit heat exchangers.« less

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

Remote monitoring of the Gravelly Run thermal plume at Hopewell and the thermal plume at the Surry Nuclear Power Plant on the James River

NASA Technical Reports Server (NTRS)

Talay, T. A.; Sykes, K. W.; Kuo, C. Y.

1979-01-01

On May 17, 1977, a remote sensing experiment was conducted on the James River, Virginia, whereby thermal spectrometer and near-infrared photography data of thermal discharges at Hopewell and the Surry nuclear power plant were obtained by an aircraft for one tidal cycle. These data were used in subsequent investigations into the near field discharge trajectories. For the Gravelly Run thermal plume at Hopewell, several empirical expressions for the plume centerline were evaluated by comparisons of the computed trajectories and those observed in the remote sensing images.

Experimental studies of thermal and chemical interactions between molten aluminum and water

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

Farahani, A.A.; Corradini, M.L.

The possibility of rapid physical and chemical aluminum/water interactions during a core melt accident in a noncommercial reactor (e.g., HFIR, ATR) has resulted in extensive research to determine the mechanism by which these interactions occur and propagate on an explosive time scale. These events have been reported in nuclear testing facilities, i.e., during SPERT 1D experiment, and also in aluminum casting industries. Although rapid chemical reactions between molten aluminum and water have been subject of many studies, very few reliable measurements of the extent of the chemical reactions have thus far been made. We have modified an existing 1-D shockmore » tube facility to perform experiments in order to determine the extent of the explosive thermal/chemical interactions between molton aluminum and water by measuring important physical quantities such as the maximum dynamic pressure and the amount of the generated hydrogen. Experimental results show that transient pressures greater than 69 MPa with a rise time of less than 125 {mu}sec can occur as the result of the chemical reaction of 4.2 grams of molton aluminum (approximately 15% of the total mass of the fuel of 28 grams) at 980 C with room temperature water.« less

Thermal effects in supernova matter

NASA Astrophysics Data System (ADS)

Constantinou, Constantinos

A crucial ingredient in simulations of core collapse supernova (SN) explosions is the equation of state (EOS) of nucleonic matter for densities extending from 10-7 fm-3 to 1 ffm-3, temperatures up to 50 MeV, and proton-to-baryon fraction in the range 0 to 1/2. SN explosions release 99% of the progenitor star's gravitational potential energy in the form of neutrinos and, additionally, they are responsible for populating the universe with elements heavier than 56Fe. Therefore, the importance of understanding this phenomenon cannot be overstated as it could shed light onto the underlying nuclear and neutrino physics. A realistic EOS of SN matter must incorporate the nucleon-nucleon interaction in a many-body environment. We treat this problem with a non-relativistic potential model as well as relativistic mean-field theoretical one. In the former approach, we employ the Skyrme-like Hamiltonian density constructed by Akmal, Pandharipande, and Ravenhall which takes into account the long scattering lengths of nucleons that determine the low density characteristics. In the latter, we use a Walecka-like Lagrangian density supplemented by non-linear interactions involving scalar, vector, and isovector meson exchanges, calibrated so that known properties of nuclear matter are reproduced. We focus on the bulk homogeneous phase and calculate its thermodynamic properties as functions of baryon density, temperature, and proton-to-baryon ratio. The exact numerical results are then compared to those in the degenerate and non-degenerate limits for which analytical formulae have been derived. We find that the two models bahave similarly for densities up to nuclear saturation but exhibit differences at higher densities most notably in the isospin susceptibilities, the chemical potentials, and the pressure. The importance of the correct momentum dependence in the single particle potential that fits optical potentials of nucleon-nucleus scattering was highlighted in the context of intermediate energy heavy-ion collisions. To explore the effect momentum-dependent interactions have on the thermal properties of dense matter we study a schematic model constructed by Welke et al. in which the appropriate momentum dependence that fits optical potential data is built through finite-range exchange forces of the Yukawa type. We look into the finite-temperature properties of this model in the context of infinite, isospin-symmetric nucleonic matter. The exact numerical results are compared to analytical ones in the quantum regime where we rely on Landau's Fermi-Liquid Theory, and in the classical regime where the state variables are obtained through a steepest descent calculation. Detailed comparisons with similarly calibrated Skyrme models are also performed. We find that the high-density behavior of the thermal pressure is once again a differentiating feature. We attribute this to the temperature dependence of the energy spectrum of the finite-range and the meson-exchange models which leads to a higher specific heat and thus a lower pressure.

Prediction of Ablation Rates from Solid Surfaces Exposed to High Temperature Gas Flow

NASA Technical Reports Server (NTRS)

Akyuzlu, Kazim M.; Coote, David

2013-01-01

A mathematical model and a solution algorithm is developed to study the physics of high temperature heat transfer and material ablation and identify the problems associated with the flow of hydrogen gas at very high temperatures and velocities through pipes and various components of Nuclear Thermal Rocket (NTR) motors. Ablation and melting can be experienced when the inner solid surface of the cooling channels and the diverging-converging nozzle of a Nuclear Thermal Rocket (NTR) motor is exposed to hydrogen gas flow at temperatures around 2500 degrees Kelvin and pressures around 3.4 MPa. In the experiments conducted on typical NTR motors developed in 1960s, degradation of the cooling channel material (cracking in the nuclear fuel element cladding) and in some instances melting of the core was observed. This paper presents the results of a preliminary study based on two types of physics based mathematical models that were developed to simulate the thermal-hydrodynamic conditions that lead to ablation of the solid surface of a stainless steel pipe exposed to high temperature hydrogen gas near sonic velocities. One of the proposed models is one-dimensional and assumes the gas flow to be unsteady, compressible and viscous. An in-house computer code was developed to solve the conservations equations of this model using a second-order accurate finite-difference technique. The second model assumes the flow to be three-dimensional, unsteady, compressible and viscous. A commercial CFD code (Fluent) was used to solve the later model equations. Both models assume the thermodynamic and transport properties of the hydrogen gas to be temperature dependent. In the solution algorithm developed for this study, the unsteady temperature of the pipe is determined from the heat equation for the solid. The solid-gas interface temperature is determined from an energy balance at the interface which includes heat transfer from or to the interface by conduction, convection, radiation, and ablation. Two different ablation models are proposed to determine the heat loss from the solid surface due to the ablation of the solid material. Both of them are physics based. Various numerical simulations were carried out using both models to predict the temperature distribution in the solid and in the gas flow, and then predict the ablation rates at a typical NTR motor hydrogen gas temperature and pressure. Solid mass loss rate per foot of a pipe was also calculated from these predictions. The results are presented for fully developed turbulent flow conditions in a sample SS pipe with a 6 inch diameter.

Physical Limitations of Nuclear Propulsion for Earth to Orbit

NASA Technical Reports Server (NTRS)

Blevins, John A.; Patton, Bruce; Rhys, Noah O.; Schafer, Charles F. (Technical Monitor)

2001-01-01

An assessment of current nuclear propulsion technology for application in Earth to Orbit (ETO) missions has been performed. It can be shown that current nuclear thermal rocket motors are not sufficient to provide single stage performance as has been stated by previous studies. Further, when taking a systems level approach, it can be shown that NTRs do not compete well with chemical engines where thrust to weight ratios of greater than I are necessary, except possibly for the hybrid chemical/nuclear LANTR (LOX Augmented Nuclear Thermal Rocket) engine. Also, the ETO mission requires high power reactors and consequently large shielding weights compared to NTR space missions where shadow shielding can be used. In the assessment, a quick look at the conceptual ASPEN vehicle proposed in 1962 in provided. Optimistic NTR designs are considered in the assessment as well as discussion on other conceptual nuclear propulsion systems that have been proposed for ETO. Also, a quick look at the turbulent, convective heat transfer relationships that restrict the exchange of nuclear energy to thermal energy in the working fluid and consequently drive the reactor mass is included.

Pressure-assisted thermal sterilization of soup

NASA Astrophysics Data System (ADS)

Shibeshi, Kidane; Farid, Mohammed M.

2010-12-01

The overall efficiency of an existing scale-up pressure-assisted thermal sterilization (PATS) unit was investigated with regards to inactivation of Geobacillus stearothermophilus spores suspended in pumpkin soup. The PATS unit is a double pipe heat exchanger in which the soup is pumped into its inner high pressure tube and constrained by two high pressure valves, while steam is continuously passed through the annular region to heat the content. The technology is based on pressure generation by thermal expansion of the liquid in an enclosure. In this work, the addition of an air line to push the treated liquid food out of the existing PATS unit has improved the overall quality of the treated samples, as evidenced by achieving higher log reduction of the spores. Compared with thermal processing, the application of PATS shows the potential for lowering the thermal treatment temperature, offering improved food quality.

Radio-Frequency Driven Dielectric Heaters for Non-Nuclear Testing in Nuclear Core Development

NASA Technical Reports Server (NTRS)

Sims, William Herbert, III (Inventor); Godfroy, Thomas J. (Inventor); Bitteker, Leo (Inventor)

2006-01-01

Apparatus and methods are provided through which a radiofrequency dielectric heater has a cylindrical form factor, a variable thermal energy deposition through variations in geometry and composition of a dielectric, and/or has a thermally isolated power input.

Non-thermal plasma activates human keratinocytes by stimulation of antioxidant and phase II pathways.

PubMed

Schmidt, Anke; Dietrich, Stephan; Steuer, Anna; Weltmann, Klaus-Dieter; von Woedtke, Thomas; Masur, Kai; Wende, Kristian

2015-03-13

Non-thermal atmospheric pressure plasma provides a novel therapeutic opportunity to control redox-based processes, e.g. wound healing, cancer, and inflammatory diseases. By spatial and time-resolved delivery of reactive oxygen and nitrogen species, it allows stimulation or inhibition of cellular processes in biological systems. Our data show that both gene and protein expression is highly affected by non-thermal plasma. Nuclear factor erythroid-related factor 2 (NRF2) and phase II enzyme pathway components were found to act as key controllers orchestrating the cellular response in keratinocytes. Additionally, glutathione metabolism, which is a marker for NRF2-related signaling events, was affected. Among the most robustly increased genes and proteins, heme oxygenase 1, NADPH-quinone oxidoreductase 1, and growth factors were found. The roles of NRF2 targets, investigated by siRNA silencing, revealed that NRF2 acts as an important switch for sensing oxidative stress events. Moreover, the influence of non-thermal plasma on the NRF2 pathway prepares cells against exogenic noxae and increases their resilience against oxidative species. Via paracrine mechanisms, distant cells benefit from cell-cell communication. The finding that non-thermal plasma triggers hormesis-like processes in keratinocytes facilitates the understanding of plasma-tissue interaction and its clinical application. © 2015 by The American Society for Biochemistry and Molecular Biology, Inc.

Numerical Modeling of Thermal-Hydrology in the Near Field of a Generic High-Level Waste Repository

NASA Astrophysics Data System (ADS)

Matteo, E. N.; Hadgu, T.; Park, H.

2016-12-01

Disposal in a deep geologic repository is one of the preferred option for long term isolation of high-level nuclear waste. Coupled thermal-hydrologic processes induced by decay heat from the radioactive waste may impact fluid flow and the associated migration of radionuclides. This study looked at the effects of those processes in simulations of thermal-hydrology for the emplacement of U. S. Department of Energy managed high-level waste and spent nuclear fuel. Most of the high-level waste sources have lower thermal output which would reduce the impact of thermal propagation. In order to quantify the thermal limits this study concentrated on the higher thermal output sources and on spent nuclear fuel. The study assumed a generic nuclear waste repository at 500 m depth. For the modeling a representative domain was selected representing a portion of the repository layout in order to conduct a detailed thermal analysis. A highly refined unstructured mesh was utilized with refinements near heat sources and at intersections of different materials. Simulations looked at different values for properties of components of the engineered barrier system (i.e. buffer, disturbed rock zone and the host rock). The simulations also looked at the effects of different durations of surface aging of the waste to reduce thermal perturbations. The PFLOTRAN code (Hammond et al., 2014) was used for the simulations. Modeling results for the different options are reported and include temperature and fluid flow profiles in the near field at different simulation times. References:G. E. Hammond, P.C. Lichtner and R.T. Mills, "Evaluating the Performance of Parallel Subsurface Simulators: An Illustrative Example with PFLOTRAN", Water Resources Research, 50, doi:10.1002/2012WR013483 (2014). Sandia National Laboratories is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND2016-7510 A

Thermal radiative and thermodynamic properties of solid and liquid uranium and plutonium carbides in the visible-near-infrared range

NASA Astrophysics Data System (ADS)

Fisenko, Anatoliy I.; Lemberg, Vladimir F.

2016-09-01

The knowledge of thermal radiative and thermodynamic properties of uranium and plutonium carbides under extreme conditions is essential for designing a new metallic fuel materials for next generation of a nuclear reactor. The present work is devoted to the study of the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides at their melting/freezing temperatures. The Stefan-Boltzmann law, total energy density, number density of photons, Helmholtz free energy density, internal energy density, enthalpy density, entropy density, heat capacity at constant volume, pressure, and normal total emissivity are calculated using experimental data for the frequency dependence of the normal spectral emissivity of liquid and solid uranium and plutonium carbides in the visible-near infrared range. It is shown that the thermal radiative and thermodynamic functions of uranium carbide have a slight difference during liquid-to-solid transition. Unlike UC, such a difference between these functions have not been established for plutonium carbide. The calculated values for the normal total emissivity of uranium and plutonium carbides at their melting temperatures is in good agreement with experimental data. The obtained results allow to calculate the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides for any size of samples. Based on the model of Hagen-Rubens and the Wiedemann-Franz law, a new method to determine the thermal conductivity of metals and carbides at the melting points is proposed.

Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

NASA Astrophysics Data System (ADS)

Woo, Kyoungsuk

Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation, the flow becomes stable below a certain heat flux regardless of the inlet subcooling at the core and system pressure. At higher heat flux, unstable phenomena were indentified within a certain range of inlet subcooling. The unstable region diminishes as the system pressure increases. In natural circulation BWRs, the significant gravitational pressure drop over the tall chimney section induces a Type-I instability. The Type-I instability becomes especially important during low power and pressure conditions during reactor start-up. Under these circumstances the effect of pressure variations on the saturation enthalpy becomes significant. An experimental study shows that the flashing phenomenon in the adiabatic chimney section is dominant during the start-up of a natural circulation BWR. Since flashing occurs outside the core, nuclear feedback effects on the stability are small. Furthermore, the thermal-hydraulic oscillation period is much longer than power fluctuation period caused by void reactivity feedback. In the natural circulation system increasing the inlet restriction reduces the natural circulation flow rate, shifting the unstable region to higher inlet subcooling.

Effects of nuclear forces on ion thermalization in high-temperature plasmas

NASA Technical Reports Server (NTRS)

Gould, R. J.

1982-01-01

A number of investigations have been concerned with the kinetic theory and processes associated with a relativistic electron gas. Gould (1981) has considered a condition in which upon the ultimate thermalization the temperature can be such that the electron gas is highly relativistic while the gas of protons and other ions is nonrelativistic. With the nuclear component nonrelativistic but having energies in the MeV range and above, it is necessary to consider the effects of nuclear forces in the scattering of the ions in their thermalization. The effects of nuclear forces in the thermalization of ions in plasmas have been computed, principally in connection with problems of controlle; fusion. The present investigation is concerned with an attempt to express results in analytic form to as great a degree as possible. The p-p problem, which is the fundamental problem in astrophysical plasma, is studied. Attention is given to a low-energy formulation, the s-wave phase shift, the effective stopping number, Fokker-Planck operators, and the interaction with the electron gas.

Nuclear Thermal Rocket - Arc Jet Integrated System Model

NASA Technical Reports Server (NTRS)

Taylor, Brian D.; Emrich, William

2016-01-01

In the post-shuttle era, space exploration is moving into a new regime. Commercial space flight is in development and is planned to take on much of the low earth orbit space flight missions. With the development of a heavy lift launch vehicle, the Space Launch, System, NASA has become focused on deep space exploration. Exploration into deep space has traditionally been done with robotic probes. More ambitious missions such as manned missions to asteroids and Mars will require significant technology development. Propulsion system performance is tied to the achievability of these missions and the requirements of other developing technologies that will be required. Nuclear thermal propulsion offers a significant improvement over chemical propulsion while still achieving high levels of thrust. Opportunities exist; however, to build upon what would be considered a standard nuclear thermal engine to attain improved performance, thus further enabling deep space missions. This paper discuss the modeling of a nuclear thermal system integrated with an arc jet to further augment performance. The performance predictions and systems impacts are discussed.

Processing of baby food using pressure-assisted thermal sterilization (PATS) and comparison with thermal treatment

NASA Astrophysics Data System (ADS)

Wang, Yubin; Ismail, Marliya; Farid, Mohammed

2017-10-01

Currently baby food is sterilized using retort processing that gives an extended shelf life. However, this type of heat processing leads to reduction of organoleptic and nutrition value. Alternatively, the combination of pressure and heat could be used to achieve sterilization at reduced temperatures. This study investigates the potential of pressure-assisted thermal sterilization (PATS) technology for baby food sterilization. Here, baby food (apple puree), inoculated with Bacillus subtilis spores was treated using PATS at different operating temperatures, pressures and times and was compared with thermal only treatment. The results revealed that the decimal reduction time of B. subtilis in PATS treatment was lower than that of thermal only treatment. At a similar spore inactivation, the retention of ascorbic acid of PATS-treated sample was higher than that of thermally treated sample. The results indicated that PATS could be a potential technology for baby food processing while minimizing quality deterioration.

Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs

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

James Werner

The development of a Nuclear Thermal Propulsion (NTP) system rests heavily upon being able to fabricate and demonstrate the performance of a high temperature nuclear fuel as well as demonstrating an integrated system prior to launch. A number of studies have been performed in the past which identified the facilities needed and the capabilities available to meet the needs and requirements identified at that time. Since that time, many facilities and capabilities within the Department of Energy have been removed or decommissioned. This paper provides a brief overview of the anticipated facility needs and identifies some promising concepts to bemore » considered which could support the development of a nuclear thermal propulsion system. Detailed trade studies will need to be performed to support the decision making process.« less

Open cycle gas core nuclear rockets

NASA Technical Reports Server (NTRS)

Ragsdale, Robert

1991-01-01

The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

A Model of Thermal Conductivity for Planetary Soils: 1. Theory for Unconsolidated Soils

NASA Technical Reports Server (NTRS)

Piqueux, S.; Christensen, P. R.

2009-01-01

We present a model of heat conduction for mono-sized spherical particulate media under stagnant gases based on the kinetic theory of gases, numerical modeling of Fourier s law of heat conduction, theoretical constraints on the gas thermal conductivity at various Knudsen regimes, and laboratory measurements. Incorporating the effect of the temperature allows for the derivation of the pore-filling gas conductivity and bulk thermal conductivity of samples using additional parameters (pressure, gas composition, grain size, and porosity). The radiative and solid-to-solid conductivities are also accounted for. Our thermal model reproduces the well-established bulk thermal conductivity dependency of a sample with the grain size and pressure and also confirms laboratory measurements finding that higher porosities generally lead to lower conductivities. It predicts the existence of the plateau conductivity at high pressure, where the bulk conductivity does not depend on the grain size. The good agreement between the model predictions and published laboratory measurements under a variety of pressures, temperatures, gas compositions, and grain sizes provides additional confidence in our results. On Venus, Earth, and Titan, the pressure and temperature combinations are too high to observe a soil thermal conductivity dependency on the grain size, but each planet has a unique thermal inertia due to their different surface temperatures. On Mars, the temperature and pressure combination is ideal to observe the soil thermal conductivity dependency on the average grain size. Thermal conductivity models that do not take the temperature and the pore-filling gas composition into account may yield significant errors.

Nuclear thermal rocket nozzle testing and evaluation program

NASA Technical Reports Server (NTRS)

Davidian, Kenneth O.; Kacynski, Kenneth J.

1993-01-01

Performance characteristics of the Nuclear Thermal Rocket can be enhanced through the use of unconventional nozzles as part of the propulsion system. The Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis is outlined and the advantages of a plug nozzle are described. A facility description, experimental designs and schematics are given. Results of pretest performance analyses show that high nozzle performance can be attained despite substantial nozzle length reduction through the use of plug nozzles as compared to a convergent-divergent nozzle. Pretest measurement uncertainty analyses indicate that specific impulse values are expected to be within + or - 1.17 pct.

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

NASA Astrophysics Data System (ADS)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Experimental study of thermal conductivity at high pressures: Implications for the deep Earth’s interior

DOE PAGES

Goncharov, Alexander F.; Lobanov, Sergey S.; Tan, Xiaojing; ...

2015-02-24

Lattice thermal conductivity of ferropericlase and radiative thermal conductivity of iron bearing magnesium silicate perovskite (bridgmanite) – the major mineral of Earth’s lower mantle– has been measured at room temperature up to 30 and 46 GPa, respectively, using time domain thermoreflectance and optical spectroscopy techniques in diamond anvil cells. The results provide new constraints for the pressure dependencies of the thermal conductivities of Fe bearing minerals. The lattice thermal conductivity of ferropericlase (Mg 0.9Fe 0.1)O is 5.7(6) W/(m*K) at ambient conditions, which is almost 10 times smaller than that of pure MgO; however, it increases with pressure much faster (6.1(7)%/GPamore » vs 3.6%/GPa). The radiative conductivity of Mg 0.94Fe 0.06SiO 3 bridgmanite single crystal agrees with previously determined values at ambient pressure; it is almost pressure-independent in the investigated pressure range. Furthermore, our results confirm the reduced radiative conductivity scenario for the Earth’s lower mantle, while the assessment of the heat flow through the core-mantle boundary still requires in situ measurements at the relevant pressure-temperature conditions.« less

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Room temperature thermal conductivity measurements of neat MOF-5 compacts with high pressure hydrogen and helium

DOE PAGES

Semelsberger, Troy Allen; Veenstra, Mike; Dixon, Craig

2016-02-09

Metal-organic frameworks (MOFs) are a highly porous crystalline material with potential in various applications including on-board vehicle hydrogen storage for fuel cell vehicles. The thermal conductivity of MOFs is an important parameter in the design and ultimate performance of an on-board hydrogen storage system. However, in-situ thermal conductivity measurements have not been previously reported. The present study reports room temperature thermal conductivity and thermal diffusivity measurements performed on neat MOF-5 cylindrical compacts (ρ = 0.4 g/mL) as a function of pressure (0.27–90 bar) and gas type (hydrogen and helium). The transient plane source technique was used to measure both themore » non-directional thermal properties (isotropic method) and the directional thermal properties (anisotropic method). High pressure measurements were made using our in-house built low-temperature, high pressure thermal conductivity sample cell. The intrinsic thermal properties of neat MOF-5 measured under vacuum were—Isotropic: k isotropic = 0.1319 W/m K, α isotropic = 0.4165 mm 2/s; Anisotropic: k axial = 0.1477 W/m K, k radial = 0.1218 W/m K, α axial = 0.5096 mm 2/s, and α radial = 0.4232 mm 2/s. The apparent thermal properties of neat MOF-5 increased with increasing hydrogen and helium pressure, with the largest increase occurring in the narrow pressure range of 0–10 bar and then monotonically asymptoting with increasing pressures up to around 90 bar. On average, a greater than two-fold enhancement in the apparent thermal properties was observed with neat MOF-5 in the presence of helium and hydrogen compared to the intrinsic values of neat MOF-5 measured under vacuum. The apparent thermal properties of neat MOF-5 measured with hydrogen were higher than those measured with helium, which were directly related to the gas-specific thermal properties of helium and hydrogen. Neat MOF-5 exhibited a small degree of anisotropy under all conditions measured with thermal conductivities and diffusivities in the axial direction being higher than those in the radial direction. As a result, the low temperature specific heat capacities of neat MOF-5 were also measured and reported for the temperature range of 93–313 K (–180–40 °C).« less

Room temperature thermal conductivity measurements of neat MOF-5 compacts with high pressure hydrogen and helium

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

Semelsberger, Troy Allen; Veenstra, Mike; Dixon, Craig

Metal-organic frameworks (MOFs) are a highly porous crystalline material with potential in various applications including on-board vehicle hydrogen storage for fuel cell vehicles. The thermal conductivity of MOFs is an important parameter in the design and ultimate performance of an on-board hydrogen storage system. However, in-situ thermal conductivity measurements have not been previously reported. The present study reports room temperature thermal conductivity and thermal diffusivity measurements performed on neat MOF-5 cylindrical compacts (ρ = 0.4 g/mL) as a function of pressure (0.27–90 bar) and gas type (hydrogen and helium). The transient plane source technique was used to measure both themore » non-directional thermal properties (isotropic method) and the directional thermal properties (anisotropic method). High pressure measurements were made using our in-house built low-temperature, high pressure thermal conductivity sample cell. The intrinsic thermal properties of neat MOF-5 measured under vacuum were—Isotropic: k isotropic = 0.1319 W/m K, α isotropic = 0.4165 mm 2/s; Anisotropic: k axial = 0.1477 W/m K, k radial = 0.1218 W/m K, α axial = 0.5096 mm 2/s, and α radial = 0.4232 mm 2/s. The apparent thermal properties of neat MOF-5 increased with increasing hydrogen and helium pressure, with the largest increase occurring in the narrow pressure range of 0–10 bar and then monotonically asymptoting with increasing pressures up to around 90 bar. On average, a greater than two-fold enhancement in the apparent thermal properties was observed with neat MOF-5 in the presence of helium and hydrogen compared to the intrinsic values of neat MOF-5 measured under vacuum. The apparent thermal properties of neat MOF-5 measured with hydrogen were higher than those measured with helium, which were directly related to the gas-specific thermal properties of helium and hydrogen. Neat MOF-5 exhibited a small degree of anisotropy under all conditions measured with thermal conductivities and diffusivities in the axial direction being higher than those in the radial direction. As a result, the low temperature specific heat capacities of neat MOF-5 were also measured and reported for the temperature range of 93–313 K (–180–40 °C).« less

Irradiation of nuclear track emulsions with thermal neutrons, heavy ions, and muons

NASA Astrophysics Data System (ADS)

Artemenkov, D. A.; Bradnova, V.; Zaitsev, A. A.; Zarubin, P. I.; Zarubina, I. G.; Kattabekov, R. R.; Mamatkulov, K. Z.; Rusakova, V. V.

2015-07-01

Exposures of test samples of nuclear track emulsion were analyzed. Angular and energy correlations of products originating from the thermal-neutron-induced reaction n th +10 B → 7 Li + (γ)+ α were studied in nuclear track emulsions enriched in boron. Nuclear track emulsions were also irradiated with 86Kr+17 and 124Xe+26 ions of energy about 1.2 MeV per nucleon. Measurements of ranges of heavy ions in nuclear track emulsionsmade it possible to determine their energies on the basis of the SRIM model. The formation of high-multiplicity nuclear stars was observed upon irradiating nuclear track emulsions with ultrarelativistic muons. Kinematical features studied in this exposure of nuclear track emulsions for events of the muon-induced splitting of carbon nuclei to three alpha particles are indicative of the nucleardiffraction interaction mechanism.

The Pressure Dependence of Thermal Expansion of Core-Forming Alloys: A Key Parameter in Determining the Convective Style of Planetary Cores

NASA Astrophysics Data System (ADS)

Williams, Q. C.; Manghnani, M. H.

2017-12-01

The convective style of planetary cores is critically dependent on the thermal properties of iron alloys. In particular, the relation between the adiabatic gradient and the melting curve governs whether planetary cores solidify from their top down (when the adiabat is steeper than the melting curve) or the bottom up (the converse). Molten iron alloys, in general, have large, ambient pressure thermal expansions: values in excess of 1.2 x 10^-4/K are dictated by data derived from levitated and sessile drop techniques. These high values of the thermal expansion imply that the adiabatic gradients within early planetesimals and present day moons that have comparatively low-pressure, iron-rich cores are steep (typically greater than 35 K/GPa at low pressures): values, at low pressures, that are greater than the slope of the melting curve, and hence show that the cores of small solar system objects probably crystallize from the top-down. Here, we deploy a different manifestation of these large values of thermal expansion to determine the pressure dependence of thermal expansion in iron-rich liquids: a difficult parameter to experimentally measure, and critical for determining the size range of cores in which top-down core solidification predominates. In particular, the difference between the adiabatic and isothermal bulk moduli of iron liquids is in the 20-30% range at the melting temperature, and scales as the product of the thermal expansion, the Grüneisen parameter, and the temperature. Hence, ultrasonic (and adiabatic) moduli of iron alloy liquids, when coupled with isothermal sink-float measurements, can yield quantitative constraints on the pressure dependence of thermal expansion. For liquid iron alloys containing 17 wt% Si, we find that the thermal expansion is reduced by 50% over the first 8 GPa of compression. This "squeezing out" of the anomalously high low-pressure thermal expansion of iron-rich alloys at relatively modest conditions likely limits the size range over which top-down crystallizing cores are anticipated within planetary bodies.

Near-zero thermal expansion in magnetically ordered state in dysprosium at high pressures and low temperatures

NASA Astrophysics Data System (ADS)

Hope, Kevin M.; Samudrala, Gopi K.; Vohra, Yogesh K.

2017-01-01

The atomic volume of rare earth metal dysprosium (Dy) has been measured up to high pressures of 35 GPa and low temperatures between 200 and 7 K in a diamond anvil cell using angle dispersive X-ray diffraction at a synchrotron source. The hexagonal close-packed (hcp), alpha-Samarium (α-Sm), and double hexagonal close-packed (dhcp) phases are observed to be stable in Dy under high-pressure and low-temperature conditions achieved in our experiments. Dy is known to undergo magnetic ordering below 176 K at ambient pressure with magnetic ordering Néel temperature (TN) that changes rapidly with increasing pressure. Our experimental measurement shows that Dy has near-zero thermal expansion in the magnetically ordered state and normal thermal expansion in the paramagnetic state for all the three known high pressure phases (hcp, α-Sm, and dhcp) to 35 GPa. This near-zero thermal expansion behavior in Dy is observed below the magnetic ordering temperature TN at all pressures up to 35 GPa.

Propagation of Pressure Waves, Caused by a Thermal Shock, in Liquid Metals Containing Gas Bubbles

NASA Astrophysics Data System (ADS)

Okita, Kohei; Takagi, Shu; Matsumoto, Yoichiro

The propagation of pressure waves caused by a thermal shock in liquid mercury containing micro gas bubbles has been simulated numerically. In the present study, we clarify the influences of the introduced bubble size and void fraction on the absorption of thermal expansion of liquid mercury and attenuation of pressure waves. The mass, momentum and energy conservation equations for both bubbly mixture and gas inside each bubble are solved, in which the bubble dynamics is represented by the Keller equation. The results show that when the initial void fraction is larger than the rate of the thermal expansion of liquid mercury, the pressure rise caused by the thermal expansion decreases with decreasing the bubble radius, because of the increase of the natural frequency of bubbly mixture. On the other hand, as the bubble radius increases, the peak of pressure waves which propagate at the sound speed of mixture decreases gradually due to the dispersion effect of mixture. When the natural frequency of the mixture with large bubbles is lower than that of the thremal shock, the peak pressure at the wall increases because the pressure waves propagate through the mixture at the sound speed of liquid mercury. The comparison of the results with and without heat transfer through the gas liquid interface shows that the pressure waves are attenuated greatly by the thermal damping effect with the decrease of the void fraction which enhances the nonlinearity of bubble oscillation.

Transmutation Fuel Performance Code Thermal Model Verification

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

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Nuclear war between Israel and Iran: lethality beyond the pale

PubMed Central

2013-01-01

Background The proliferation of nuclear technology in the politically volatile Middle East greatly increases the likelihood of a catastrophic nuclear war. It is widely accepted, while not openly declared, that Israel has nuclear weapons, and that Iran has enriched enough nuclear material to build them. The medical consequences of a nuclear exchange between Iran and Israel in the near future are envisioned, with a focus on the distribution of casualties in urban environments. Methods Model estimates of nuclear war casualties employed ESRI's ArcGIS 9.3, blast and prompt radiation were calculated using the Defense Nuclear Agency's WE program, and fallout radiation was calculated using the Defense Threat Reduction Agency's (DTRA's) Hazard Prediction and Assessment Capability (HPAC) V404SP4, as well as custom GIS and database software applications. Further development for thermal burn casualties was based on Brode, as modified by Binninger, to calculate thermal fluence. ESRI ArcGISTM programs were used to calculate affected populations from the Oak Ridge National Laboratory's LandScanTM 2007 Global Population Dataset for areas affected by thermal, blast and radiation data. Results Trauma, thermal burn, and radiation casualties were thus estimated on a geographic basis for three Israeli and eighteen Iranian cities. Nuclear weapon detonations in the densely populated cities of Iran and Israel will result in an unprecedented millions of numbers of dead, with millions of injured suffering without adequate medical care, a broad base of lingering mental health issues, a devastating loss of municipal infrastructure, long-term disruption of economic, educational, and other essential social activity, and a breakdown in law and order. Conclusions This will cause a very limited medical response initially for survivors in Iran and Israel. Strategic use of surviving medical response and collaboration with international relief could be expedited by the predicted casualty distributions and locations. The consequences for health management of thermal burn and radiation patients is the worst, as burn patients require enormous resources to treat, and there will be little to no familiarity with the treatment of radiation victims. Any rational analysis of a nuclear war between Iran and Israel reveals the utterly unacceptable outcomes for either nation. PMID:23663406

Pressurized heat treatment of glass ceramic

DOEpatents

Kramer, D.P.

1984-04-19

A method of producing a glass-ceramic having a specified thermal expansion value is disclosed. The method includes the step of pressurizing the parent glass material to a predetermined pressure during heat treatment so that the glass-ceramic produced has a specified thermal expansion value. Preferably, the glass-ceramic material is isostatically pressed. A method for forming a strong glass-ceramic to metal seal is also disclosed in which the glass-ceramic is fabricated to have a thermal expansion value equal to that of the metal. The determination of the thermal expansion value of a parent glass material placed in a high-temperature environment is also used to determine the pressure in the environment.

Fluctuation of a Piston in Vacuum Induced by Thermal Radiation Pressure

NASA Astrophysics Data System (ADS)

Inui, Norio

2017-10-01

We consider the displacement of a piston dividing a vacuum cavity at a finite temperature T induced by fluctuations in the thermal radiation pressure. The correlation function of the thermal radiation pressure is calculated using the theoretical framework developed by Barton, which was first applied to the fluctuation of the Casimir force at absolute zero. We show that the variance of the radiation pressure at a fixed point is proportional to T8 and evaluate the mean square displacement for a piston with a small cross section in a characteristic correlation timescale ħ/(kBT). At room temperature, the contribution of the thermal radiation to the fluctuation is larger than that of the vacuum fluctuation.

Mini-MITEE: Ultra Small, Ultra Light NTP Engines for Robotic Science and Manned Exploration Missions

NASA Astrophysics Data System (ADS)

Powell, James; Maise, George; Paniagua, John

2006-01-01

A compact, ultra lightweight Nuclear Thermal Propulsion (NTP) engine design is described with the capability to carry out a wide range of unique and important robotic science missions that are not possible using chemical or Nuclear Electric Propulsion (NEP). The MITEE (MInature ReacTor EnginE) reactor uses hydrogeneous moderator, such as solid lithium-7 hydride, and high temperature cermet tungsten/UO2 nuclear fuel. The reactor is configured as a modular pressure tube assembly, with each pressure tube containing an outer annual shell of moderator with an inner annular region of W/UO2 cermet fuel sheets. H2 propellant flows radially inwards through the moderator and fuel regions, exiting at ~3000 K into a central channel that leads to a nozzle at the end of the pressure tube. Power density in the fuel region is 10 to 20 megawatts per liter, depending on design, producing a thrust output on the order of 15,000 Newtons and an Isp of ~1000 seconds. 3D Monte Carlo neutronic analyses are described for MITEE reactors utilizing various fissile fuel options (U-235, U-233, and Am242m) and moderators (7LiH and BeH2). Reactor mass ranges from a maximum of 100 kg for the 7LiH/U-235 option to a minimum of 28 kg for the BeH2/Am-242 m option. Pure thrust only and bi-modal (thrust plus electric power generation) MITEE designs are described. Potential unique robotic science missions enabled by the MITEE engine are described, including landing on Europa and exploring the ice sheet interior with return of samples to Earth, hopping to and exploring multiple sites on Mars, unlimited ramjet flight in the atmospheres of Jupiter, Saturn, Uranus, and Neptune and landing on, and sample return from Pluto.

Mini-MITEE: Ultra Small, Ultra Light NTP Engines for Robotic Science and Manned Exploration Missions

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

Powell, James; Maise, George; Paniagua, John

2006-01-20

A compact, ultra lightweight Nuclear Thermal Propulsion (NTP) engine design is described with the capability to carry out a wide range of unique and important robotic science missions that are not possible using chemical or Nuclear Electric Propulsion (NEP). The MITEE (MInature ReacTor EnginE) reactor uses hydrogeneous moderator, such as solid lithium-7 hydride, and high temperature cermet tungsten/UO2 nuclear fuel. The reactor is configured as a modular pressure tube assembly, with each pressure tube containing an outer annual shell of moderator with an inner annular region of W/UO2 cermet fuel sheets. H2 propellant flows radially inwards through the moderator andmore » fuel regions, exiting at {approx}3000 K into a central channel that leads to a nozzle at the end of the pressure tube. Power density in the fuel region is 10 to 20 megawatts per liter, depending on design, producing a thrust output on the order of 15,000 Newtons and an Isp of {approx}1000 seconds. 3D Monte Carlo neutronic analyses are described for MITEE reactors utilizing various fissile fuel options (U-235, U-233, and Am242m) and moderators (7LiH and BeH2). Reactor mass ranges from a maximum of 100 kg for the 7LiH/U-235 option to a minimum of 28 kg for the BeH2/Am-242 m option. Pure thrust only and bi-modal (thrust plus electric power generation) MITEE designs are described. Potential unique robotic science missions enabled by the MITEE engine are described, including landing on Europa and exploring the ice sheet interior with return of samples to Earth, hopping to and exploring multiple sites on Mars, unlimited ramjet flight in the atmospheres of Jupiter, Saturn, Uranus, and Neptune and landing on, and sample return from Pluto.« less

Startup of Pumping Units in Process Water Supplies with Cooling Towers at Thermal and Nuclear Power Plants

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

Berlin, V. V., E-mail: vberlin@rinet.ru; Murav’ev, O. A., E-mail: muraviov1954@mail.ru; Golubev, A. V., E-mail: electronik@inbox.ru

Aspects of the startup of pumping units in the cooling and process water supply systems for thermal and nuclear power plants with cooling towers, the startup stages, and the limits imposed on the extreme parameters during transients are discussed.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Study of methods to increase cluster/dislocation loop densities in electrodes

NASA Astrophysics Data System (ADS)

Yang, Xiaoling; Miley, George H.

2009-03-01

Recent research has developed a technique for imbedding ultra-high density deuterium ``clusters'' (50 to 100 atoms per cluster) in various metals such as Palladium (Pd), Beryllium (Be) and Lithium (Li). It was found the thermally dehydrogenated PdHx retained the clusters and exhibited up to 12 percent lower resistance compared to the virginal Pd samplesootnotetextA. G. Lipson, et al. Phys. Solid State. 39 (1997) 1891. SQUID measurements showed that in Pd these condensed matter clusters approach metallic conditions, exhibiting superconducting propertiesootnotetextA. Lipson, et al. Phys. Rev. B 72, 212507 (2005ootnotetextA. G. Lipson, et al. Phys. Lett. A 339, (2005) 414-423. If the fabrication methods under study are successful, a large packing fraction of nuclear reactive clusters can be developed in the electrodes by electrolyte or high pressure gas loading. This will provide a much higher low-energy-nuclear- reaction (LENR) rate than achieved with earlier electrodeootnotetextCastano, C.H., et al. Proc. ICCF-9, Beijing, China 19-24 May, 2002..

Two-Phase flow instrumentation for nuclear accidents simulation

NASA Astrophysics Data System (ADS)

Monni, G.; De Salve, M.; Panella, B.

2014-11-01

The paper presents the research work performed at the Energy Department of the Politecnico di Torino, concerning the development of two-phase flow instrumentation and of models, based on the analysis of experimental data, that are able to interpret the measurement signals. The study has been performed with particular reference to the design of power plants, such as nuclear water reactors, where the two-phase flow thermal fluid dynamics must be accurately modeled and predicted. In two-phase flow typically a set of different measurement instruments (Spool Piece - SP) must be installed in order to evaluate the mass flow rate of the phases in a large range of flow conditions (flow patterns, pressures and temperatures); moreover, an interpretative model of the SP need to be developed and experimentally verified. The investigated meters are: Turbine, Venturi, Impedance Probes, Concave sensors, Wire mesh sensor, Electrical Capacitance Probe. Different instrument combinations have been tested, and the performance of each one has been analyzed.

Origin of coronal mass ejection and magnetic cloud: Thermal or magnetic driven?

NASA Technical Reports Server (NTRS)

Zhang, Gong-Liang; Wang, Chi; He, Shuang-Hua

1995-01-01

A fundamental problem in Solar-Terrestrial Physics is the origin of the solar transient plasma output, which includes the coronal mass ejection and its interplanetary manifestation, e.g. the magnetic cloud. The traditional blast wave model resulted from solar thermal pressure impulse has faced with challenge during recent years. In the MHD numerical simulation study of CME, the authors find that the basic feature of the asymmetrical event on 18 August 1980 can be reproduced neither by a thermal pressure nor by a speed increment. Also, the thermal pressure model fails in simulating the interplanetary structure with low thermal pressure and strong magnetic field strength, representative of a typical magnetic cloud. Instead, the numerical simulation results are in favor of the magnetic field expansion as the likely mechanism for both the asymmetrical CME event and magnetic cloud.

Development and preliminary verification of the 3D core neutronic code: COCO

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

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Thermal-neutron capture gamma-rays. Volume 2

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

Tuli, J.K.

1997-05-01

The energy and photon intensity of gamma rays as seen in thermal-neutron capture are presented ordered by Z, A of target nuclei. All gamma-rays with intensity of {ge}2% of the strongest transition are included. The strongest transition is indicated in each case. Where the target nuclide mass number is indicated as nat the natural target was used. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. All data for A > 44 are taken from Evaluated Nuclear Structure Data File (4/97), a computer file of evaluated nuclear structure data maintainedmore » by the National Nuclear Data Center, Brookhaven National Laboratory, on behalf of the Nuclear Structure and Decay and Decay Data network, coordinated by the International Atomic Energy Agency, Vienna. These data are published in Nuclear Data Sheets, Academic Press, San Diego, CA. The data for A {le} 44 is taken from ``Prompt Gamma Rays from Thermal-Neutron Capture,`` M.A. Lone, R.A. Leavitt, D.A. Harrison, Atomic Data and Nuclear Data Tables 26, 511 (1981).« less

Thermal-neutron capture gamma-rays. Volume 1

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

Tuli, J.K.

1997-05-01

The energy and photon intensity of gamma rays as seen in thermal-neutron capture are presented in ascending order of gamma energy. All those gamma-rays with intensity of {ge} 2% of the strongest transition are included. The two strongest transitions seen for the target nuclide are indicated in each case. Where the target nuclide mass number is indicated as nat the natural target was used. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. All data for A > 44 are taken from Evaluated Nuclear Structure Data File (4/97), a computermore » file of evaluated nuclear structure data maintained by the National Nuclear Data Center, Brookhaven National Laboratory, on behalf of the Nuclear Structure and Decay and Decay Data network, coordinated by the International Atomic Energy Agency, Vienna. These data are published in Nuclear Data Sheets, Academic Press, San Diego, CA. The data for A {le} 44 is taken from ``Prompt Gamma Rays from Thermal-Neutron Capture,`` M.A. Lone, R.A. Leavitt, D.A. Harrison, Atomic Data and Nuclear Data Tables 26, 511 (1981).« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Non-Destructive Analysis of Natural Uranium Pellet

NASA Astrophysics Data System (ADS)

Wigley, Samantha; Glennon, Kevin; Kitcher, Evans; Folden, Cody

2017-09-01

As part of ongoing nuclear forensics research, samples of natUO2 have been irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) with the goal of simulating a pressurized heavy water reactor. Non-destructive gamma ray analysis has been performed on the samples to assay various nuclides in order to determine the burnup and time since irradiation. The quantity of 137Cs was used to determine the burnup directly, and a maximum likelihood method has been used to estimate both the burnup and the time since irradiation. This poster will discuss the most recent results of these analyses. National Science Foundation (PHY-1659847), Department of Energy (DE-FG02-93ER40773).

US/CIS integrated NTRE

NASA Astrophysics Data System (ADS)

Bulman, M. J.; Culver, D. W.; McIlwain, M. C.; Rochow, Richard; D'Yakov, E. K.; Smetannikov, V. P.

1993-06-01

The paper describes the Nuclear Thermal Energy (NTRE) engine, developed by taking advantage of mature fuel technology developed in the former Soviet Union, thus shortening the development schedule of this engine for moon and Mars explorations. The near-term NTRE engine has a number of features that provide safety, mission performance, cost, and risk benefits. These include: (1) high-temperature long-life CIS fuel, (2) high-pressure recuperated expander cycle, (3) assured restart, (4) long-life cooled nozzle with thin inner wall, (5) long-life turbopumps, (6) heat radiation and electrical power generation, and (7) component integration synergy. Diagrams of the reactor core, the recuperated bottoming cycle flow schematic, and the recuperated bottoming cycle engine schematic are presented.

Modular heat exchanger

DOEpatents

Culver, Donald W.

1978-01-01

A heat exchanger for use in nuclear reactors includes a heat exchange tube bundle formed from similar modules each having a hexagonal shroud containing a large number of thermally conductive tubes which are connected with inlet and outlet headers at opposite ends of each module, the respective headers being adapted for interconnection with suitable inlet and outlet manifold means. In order to adapt the heat exchanger for operation in a high temperature and high pressure environment and to provide access to all tube ports at opposite ends of the tube bundle, a spherical tube sheet is arranged in sealed relation across the chamber with an elongated duct extending outwardly therefrom to provide manifold means for interconnection with the opposite end of the tube bundle.

Development of crawler type device using new measuring system

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

Maruyama, T.; Sasaki, T.; Yagi, T.

1995-08-01

This paper reports the development and field application of a new device which examine shell to shell weld joints of RPV. In a BWR type nuclear power plant, there is narrow space around the Reactor Pressure Vessel (RPV) because RPV is enclosed by the Reactor Shield Wall (RSW) and thermal insulations. The developed device is characterized by a new position measuring system and magnet wheels for driving. The new position measuring system uses laser beam and ultrasonic wave. The magnet wheels make the device travel freely in the narrow space between RPV and insulation. This device is tested on mock-upsmore » and applied examination of RPVs to verify field applicability.« less

CASKS (Computer Analysis of Storage Casks): A microcomputer based analysis system for storage cask review

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

Chen, T.F.; Mok, G.C.; Carlson, R.W.

1996-12-01

CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules--the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on the impactmore » analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage asks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.« less

Casks (computer analysis of storage casks): A microcomputer based analysis system for storage cask review

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

Chen, T.F.; Mok, G.C.; Carlson, R.W.

1995-08-01

CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules: the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on themore » impact analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage casks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.« less

In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

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

Kymaelaeinen, O.; Tuomisto, H.; Theofanous, T.G.

1997-02-01

The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This papermore » summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.« less

Monitoring corrosion and chemistry phenomena in supercritical aqueous systems

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

Macdonald, D.D.; Pang, J.; Liu, C.

1994-12-31

The in situ monitoring of the chemistry and electrochemistry of aqueous heat transport fluids in thermal (nuclear and fossil) power plants is now considered essential if adequate assessment and close control of corrosion and mass transfer phenomena are to be achieved. Because of the elevated temperatures and pressures involved. new sensor technologies are required that are able to measure key parameters under plant operating conditions for extended periods of time. In this paper, the authors outline a research and development program that is designed to develop practical sensors for use in thermal power plants. The current emphasis is on sensorsmore » for measuring corrosion potential, pH, the concentrations of oxygen and hydrogen, and the electrochemical noise generated by corrosion processes at temperatures ranging from {approximately}250 C to 500 C. The program is currently at the laboratory stage, but testing of prototype sensors in a coal-fired supercritical power plant in Spain will begin shortly.« less

Assessment of candidates for target window material in accelerator-driven molybdenum-99 production

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

Strons, Philip; Bailey, James; Makarashvili, Vakhtang

2016-10-01

NorthStar Medical Technologies is pursuing production of an important medical isotope, Mo-99, through a photo-nuclear reaction of a Mo-100 target using a high-power electron accelerator. The current target utilizes an Inconel 718 window. The purpose of this study was to evaluate other candidate materials for the target window, which separates the high-pressure helium gas inside the target from the vacuum inside the accelerator beamline and is subjected to significant stress. Our initial analysis assessed the properties (density, thermal conductivity, maximum stress, minimum window thickness, maximum temperature, and figure of merit) for a range of materials, from which the three mostmore » promising were chosen: Inconel 718, 250 maraging steel, and standard-grade beryllium. These materials were subjected to further analysis to determine the effects of thermal and mechanical strain versus beam power at varying thicknesses. Both beryllium and the maraging steel were calculated to withstand more than twice as high beam power than Inconel 718.« less

Long-life, space-maintainable nuclear stage regulators and shutoff valves

NASA Technical Reports Server (NTRS)

1972-01-01

The six most promising valve, regulator, and remote coupling concepts, representing the more radical designs from twenty concepts generated, were investigated. Of the three valves, one has no moving parts because shutoff sealing is accomplished by an electromagnetic field which ionized the flowing fluid. Another valve uses liquid metal to obtain sealing. In the third valve, high sealing forces are generated by heating and expanding trapped hydrogen. The pressure regulator is an electronically controlled, electromechanically operated, single state valve. Its complexity is in electronic circuitry, and the design results in less weight, increased reliability and performance flexibility, and multipurpose application. The two remote couplings feature the minimization of weight and mechanical complexity. One concept uses a low melting temperature metal alloy which is injected into the joint cavity; upon solidification, the alloy provides a seal and a structural joint. The second concept is based on the differential thermal expansion of the coupling mating parts. At thermal equilibrium there is a predetermined interference between the parts, and sealing is achieved by interference loading.

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

Nuclear fuel cycle waste stream immobilization with cermets for improved thermal properties and waste consolidation

NASA Astrophysics Data System (ADS)

Ortega, Luis H.; Kaminski, Michael D.; Zeng, Zuotao; Cunnane, James

2013-07-01

In the pursuit of methods to improve nuclear waste form thermal properties and combine potential nuclear fuel cycle wastes, a bronze alloy was combined with an alkali, alkaline earth metal bearing ceramic to form a cermet. The alloy was prepared from copper and tin (10 mass%) powders. Pre-sintered ceramic consisting of cesium, strontium, barium and rubidium alumino-silicates was mixed with unalloyed bronze precursor powders and cold pressed to 300 × 103 kPa, then sintered at 600 °C and 800 °C under hydrogen. Cermets were also prepared that incorporated molybdenum, which has a limited solubility in glass, under similar conditions. The cermet thermal conductivities were seven times that of the ceramic alone. These improved thermal properties can reduce thermal gradients within the waste forms thus lowering internal temperature gradients and thermal stresses, allowing for larger waste forms and higher waste loadings. These benefits can reduce the total number of waste packages necessary to immobilize a given amount of high level waste and immobilize troublesome elements.

Pressure-resisting cell for high-pressure, high-resolution nuclear magnetic resonance measurements at very high magnetic fields

NASA Astrophysics Data System (ADS)

Yamada, H.; Nishikawa, K.; Honda, M.; Shimura, T.; Akasaka, K.; Tabayashi, K.

2001-02-01

A pressure-resisting cell system has been developed for high-pressure high-resolution nuclear magnetic resonance (NMR) measurements up to a maximum pressure of 600 MPa. This cell system is capable of performing high-pressure experiments with any standard spectrometer, including modern high field NMR machines. A full description of the high-pressure NMR assembly mounted on a 750 MHz spectrometer is presented along with a detailed explanation of the procedure for preparing the pressure-resisting quartz and glass cells.

Measurements of decreasing lattice thermal conductivity of ferropericlase across the high-spin to mixed-spin state.

NASA Astrophysics Data System (ADS)

Merkel, S.; Langrand, C.; Hilairet, N.; Konopkova, Z.; Andrault, D.

2016-12-01

The thermal conductivity of lower mantle minerals depends on crystal structure and phase, with important implications for the style of convection in the mantle and the heat flow across the core-mantle boundary. In this study, we demonstrate how measurements of temperature in the laser-heated diamond anvil cell (LHDAC) can be used to determine relative changes in thermal conductivity across a pressure-induced phase change. A finite-element 3D heat flow model of the LHDAC is used to simulate experimental conditions. Results from modeling show that the peak temperature in the cell is primarily controlled by the geometry, sample thermal conductivity and heat input due to laser heating. Controlling for geometry, the model can output expected temperature versus laser-power curves for an increase or decrease in thermal conductivity with pressure. The modeled temperature differences indicate that we can experimentally distinguish the sign and magnitude of a thermal conductivity change due to a pressure-induced phase change. We perform a series of experiments to test our models. In one set of experiments, we measure temperature versus laser-power as a function of pressure for the NaCl B1-B2 phase transition, over the pressure range 18 to 54 GPa. A decrease in thermal conductivity across the NaCl B1-B2 phase transition (dκ/dP = -1.6 +/- 0.2 W/(mK GPa)) is needed to explain our measurements. This result is consistent with thermal conductivity measurements of other ionic salts, which undergo the B1-B2 phase transition at much lower pressure. We apply this experiment design to investigate the effect of spin transition on an iron-bearing magnesium oxide sample. In a series of experiments, we measure temperature vs. laser power for (Mg,Fe)O with 24 mol% Fe, loaded in Ne, over a pressure range from 22 to 60 GPa. We observe an increase in thermal conductivity between 22 and 42 GPa. But between 42 and 60 GPa, a pressure range consistent with previously reported mixed-spin state phase of (Mg,Fe)O, we observe a decrease in thermal conductivity. This result suggests that there may be a broad zone, in the depth range of 1000 - 1500 km, of reduced thermal transport properties in the mantle.

Measurements of decreasing lattice thermal conductivity of ferropericlase across the high-spin to mixed-spin state.

NASA Astrophysics Data System (ADS)

McGuire, C. P.; Sawchuk, K. L. S.; Kavner, A.

2017-12-01

The thermal conductivity of lower mantle minerals depends on crystal structure and phase, with important implications for the style of convection in the mantle and the heat flow across the core-mantle boundary. In this study, we demonstrate how measurements of temperature in the laser-heated diamond anvil cell (LHDAC) can be used to determine relative changes in thermal conductivity across a pressure-induced phase change. A finite-element 3D heat flow model of the LHDAC is used to simulate experimental conditions. Results from modeling show that the peak temperature in the cell is primarily controlled by the geometry, sample thermal conductivity and heat input due to laser heating. Controlling for geometry, the model can output expected temperature versus laser-power curves for an increase or decrease in thermal conductivity with pressure. The modeled temperature differences indicate that we can experimentally distinguish the sign and magnitude of a thermal conductivity change due to a pressure-induced phase change. We perform a series of experiments to test our models. In one set of experiments, we measure temperature versus laser-power as a function of pressure for the NaCl B1-B2 phase transition, over the pressure range 18 to 54 GPa. A decrease in thermal conductivity across the NaCl B1-B2 phase transition (dκ/dP = -1.6 +/- 0.2 W/(mK GPa)) is needed to explain our measurements. This result is consistent with thermal conductivity measurements of other ionic salts, which undergo the B1-B2 phase transition at much lower pressure. We apply this experiment design to investigate the effect of spin transition on an iron-bearing magnesium oxide sample. In a series of experiments, we measure temperature vs. laser power for (Mg,Fe)O with 24 mol% Fe, loaded in Ne, over a pressure range from 22 to 60 GPa. We observe an increase in thermal conductivity between 22 and 42 GPa. But between 42 and 60 GPa, a pressure range consistent with previously reported mixed-spin state phase of (Mg,Fe)O, we observe a decrease in thermal conductivity. This result suggests that there may be a broad zone, in the depth range of 1000 - 1500 km, of reduced thermal transport properties in the mantle.

Effect of high hydrostatic pressure and whey proteins on the disruption of casein micelle isolates.

PubMed

Harte, Federico M; Gurram, Subba Rao; Luedecke, Lloyd O; Swanson, Barry G; Barbosa-Cánovas, Gustavo V

2007-11-01

High hydrostatic pressure disruption of casein micelle isolates was studied by analytical ultracentrifugation and transmission electron microscopy. Casein micelles were isolated from skim milk and subjected to combinations of thermal treatment (85 degrees C, 20 min) and high hydrostatic pressure (up to 676 MPa) with and without whey protein added. High hydrostatic pressure promoted extensive disruption of the casein micelles in the 250 to 310 MPa pressure range. At pressures greater than 310 MPa no further disruption was observed. The addition of whey protein to casein micelle isolates protected the micelles from high hydrostatic pressure induced disruption only when the mix was thermally processed before pressure treatment. The more whey protein was added (up to 5 g/l) the more the protection against high hydrostatic pressure induced micelle disruption was observed in thermally treated samples subjected to 310 MPa.

Laser-machined microcavities for simultaneous measurement of high-temperature and high-pressure.

PubMed

Ran, Zengling; Liu, Shan; Liu, Qin; Huang, Ya; Bao, Haihong; Wang, Yanjun; Luo, Shucheng; Yang, Huiqin; Rao, Yunjiang

2014-08-07

Laser-machined microcavities for simultaneous measurement of high-temperature and high-pressure are demonstrated. These two cascaded microcavities are an air cavity and a composite cavity including a section of fiber and an air cavity. They are both placed into a pressure chamber inside a furnace to perform simultaneous pressure and high-temperature tests. The thermal and pressure coefficients of the short air cavity are ~0.0779 nm/°C and ~1.14 nm/MPa, respectively. The thermal and pressure coefficients of the composite cavity are ~32.3 nm/°C and ~24.4 nm/MPa, respectively. The sensor could be used to separate temperature and pressure due to their different thermal and pressure coefficients. The excellent feature of such a sensor head is that it can withstand high temperatures of up to 400 °C and achieve precise measurement of high-pressure under high temperature conditions.

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

Measurement of thermal diffusivity of depleted uranium metal microspheres

NASA Astrophysics Data System (ADS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Thermal conductivity measurements via time-domain thermoreflectance for the characterization of radiation induced damage

DOE PAGES

Cheaito, Ramez; Gorham, Caroline S.; Carnegie Mellon Univ., Pittsburgh, PA; ...

2015-05-01

The progressive build up of displacement damage and fission products inside different systems and components of a nuclear reactor can lead to significant defect formation, degradation, and damage of the constituent materials. This structural modification can highly influence the thermal transport mechanisms and various mechanical properties of solids. In this paper we demonstrate the use of time-domain thermoreflectance (TDTR), a non-destructive method capable of measuring the thermal transport in material systems from nano to bulk scales, to study the effect of radiation damage and the subsequent changes in the thermal properties of materials. We use TDTR to show that displacementmore » damage from ion irradiation can significantly reduce the thermal conductivity of Optimized ZIRLO, a material used as fuel cladding in several current nuclear reactors. We find that the thermal conductivity of copper-niobium nanostructured multilayers does not change with helium ion irradiation doses of up to 10 15 cm -2 and ion energy of 200 keV suggesting that these structures can be used and radiation tolerant materials in nuclear reactors. We compare the effect of ion doses and ion beam energies on the measured thermal conductivity of bulk silicon. Results demonstrate that TDTR thermal measurements can be used to quantify depth dependent damage.« less

Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

DOEpatents

Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

2009-05-05

A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

DOEpatents

Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

2009-01-27

A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

DOEpatents

Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

2009-01-06

A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

Effect of air confinement on thermal contact resistance in nanoscale heat transfer

NASA Astrophysics Data System (ADS)

Pratap, Dheeraj; Islam, Rakibul; Al-Alam, Patricia; Randrianalisoa, Jaona; Trannoy, Nathalie

2018-03-01

Here, we report a detailed analysis of thermal contact resistance (R c) of nano-size contact formed between a Wollaston wire thermal probe and the used samples (fused silica and titanium) as a function of air pressure (from 1 Pa to 105 Pa). Moreover, we suggest an analytical model using experimental data to extract R c. We found that for both samples, the thermal contact resistance decreases with increasing air pressure. We also showed that R c strongly depends on the thermal conductivity of materials keeping other parameters the same, such as roughness of the probe and samples, as well as the contact force. We provide a physical explanation of the R c trend with pressure and thermal conductivity of the materials: R c is ascribed to the heat transfer through solid-solid (probe-sample) contact and confined air at nanoscale cavities, due to the rough nature of the materials in contact. The contribution of confined air on heat transfer through the probe sample contact is significant at atmospheric pressure but decreases as the pressure decreases. In vacuum, only the solid-solid contact contributes to R c. In addition, theoretical calculations using the well-known acoustic and diffuse mismatch models showed a high thermal conductivity material that exhibits high heat transmission and consequently low R c, supporting our findings.

Thermal diffusivity and nuclear spin relaxation: a continuous wave free precession NMR study.

PubMed

Venâncio, Tiago; Engelsberg, Mario; Azeredo, Rodrigo B V; Colnago, Luiz A

2006-07-01

Continuous wave free precession (CWFP) nuclear magnetic resonance is capable of yielding quantitative and easily obtainable information concerning the kinetics of processes that change the relaxation rates of the nuclear spins through the action of some external agent. In the present application, heat flow from a natural rubber sample to a liquid nitrogen thermal bath caused a large temperature gradient leading to a non-equilibrium temperature distribution. The ensuing local changes in the relaxation rates could be monitored by the decay of the CWFP signals and, from the decays, it was possible to ascertain the prevalence of a diffusive process and to obtain an average value for the thermal diffusivity.

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

Fathi, Nima; McDaniel, Patrick; Vorobieff, Peter

The aim of this paper is evaluating the efficiency of a novel combined solar-nuclear cycle. CFD-Thermal analysis is performed to apply the available surplus heat from the nuclear cycle and measure the available kinetic energy of air for the turbine of a solar chimney power plant system (SCPPS). The presented idea helps to decrease the thermal pollution and handle the water shortage supply for water plant by replacing the cooling tower by solar chimney power plant to get the surplus heat from the available warm air in the secondary loop of the reactor. By applying this idea to a typicalmore » 1000 MW nuclear power plant with a 0.33 thermal efficiency, we can increase it to 0.39.« less

Bimodal Nuclear Thermal Rocket Analysis Developments

NASA Technical Reports Server (NTRS)

Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

2014-01-01

Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

Nuclear thermal rocket nozzle testing and evaluation program

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

Davidian, K.O.; Kacynski, K.J.

Performance characteristics of the Nuclear Thermal Rocket can be enhanced through the use of unconventional nozzles as part of the propulsion system. In this report, the Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis Research Center is outlined and the advantages of a plug nozzle are described. A facility description, experimental designs and schematics are given. Results of pretest performance analyses show that high nozzle performance can be attained despite substantial nozzle length reduction through the use of plug nozzles as compared to a convergent-divergent nozzle. Pretest measurement uncertainty analyses indicate that specific impulsemore » values are expected to be within plus or minus 1.17%.« less

The nuclear thermal electric rocket: a proposed innovative propulsion concept for manned interplanetary missions

NASA Astrophysics Data System (ADS)

Dujarric, C.; Santovincenzo, A.; Summerer, L.

2013-03-01

Conventional propulsion technology (chemical and electric) currently limits the possibilities for human space exploration to the neighborhood of the Earth. If farther destinations (such as Mars) are to be reached with humans on board, a more capable interplanetary transfer engine featuring high thrust, high specific impulse is required. The source of energy which could in principle best meet these engine requirements is nuclear thermal. However, the nuclear thermal rocket technology is not yet ready for flight application. The development of new materials which is necessary for the nuclear core will require further testing on ground of full-scale nuclear rocket engines. Such testing is a powerful inhibitor to the nuclear rocket development, as the risks of nuclear contamination of the environment cannot be entirely avoided with current concepts. Alongside already further matured activities in the field of space nuclear power sources for generating on-board power, a low level investigation on nuclear propulsion has been running since long within ESA, and innovative concepts have already been proposed at an IAF conference in 1999 [1, 2]. Following a slow maturation process, a new concept was defined which was submitted to a concurrent design exercise in ESTEC in 2007. Great care was taken in the selection of the design parameters to ensure that this quite innovative concept would in all respects likely be feasible with margins. However, a thorough feasibility demonstration will require a more detailed design including the selection of appropriate materials and the verification that these can withstand the expected mechanical, thermal, and chemical environment. So far, the predefinition work made clear that, based on conservative technology assumptions, a specific impulse of 920 s could be obtained with a thrust of 110 kN. Despite the heavy engine dry mass, a preliminary mission analysis using conservative assumptions showed that the concept was reducing the required Initial Mass in Low Earth Orbit compared to conventional nuclear thermal rockets for a human mission to Mars. Of course, the realization of this concept still requires proper engineering and the dimensioning of quite unconventional machinery. A patent was filed on the concept. Because of the operating parameters of the nuclear core, which are very specific to this type of concept, it seems possible to test on ground this kind of engine at full scale in close loop using a reasonable size test facility with safe and clean conditions. Such tests can be conducted within fully confined enclosure, which would substantially increase the associated inherent nuclear safety levels. This breakthrough removes a showstopper for nuclear rocket engines development. The present paper will disclose the NTER (Nuclear Thermal Electric Rocket) engine concept, will present some of the results of the ESTEC concurrent engineering exercise, and will explain the concept for the NTER on-ground testing facility. Regulations and safety issues related to the development and implementation of the NTER concept will be addressed as well.

Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.; Doughty, Glen E.; Bhattacharyya, Samit K.

2013-01-01

A number of recent assessments have confirmed the results of several earlier studies that Nuclear Thermal Propulsion (NTP) is a leading technology for human exploration of Mars. It is generally acknowledged that NTP provides the best prospects for the transportation of humans to Mars in the 2030's. Its high Isp coupled with the high thrusts achievable, allow reasonable trip times, thereby alleviating concerns about space radiation and "claustrophobia" effects. NASA has embarked on the latest phase of the development of NTP systems, and is adopting an affordable approach in response to the pressure of the times. The affordable strategy is built on maximizing the use of the large NTP technology base developed in the 1950's and 60's. The fact that the NTP engines were actually demonstrated to work as planned, is a great risk reduction feature in its development. The strategy utilizes non-nuclear testing to the fullest extent possible, and uses focused nuclear tests for the essential qualification and certification tests. The perceived cost risk of conducting the ground tests is being addressed by considering novel testing approaches. This includes the use of boreholes to contain radioactive effluents, and use of fuel with very high retention capability for fission products. The use of prototype flight tests is being considered as final steps in the development prior to undertaking human flight missions. In addition to the technical issues, plans are being prepared to address the institutional and political issues that need to be considered in this major venture. While the development and deployment of NTP system is not expected to be cheap, the value of the system will be very high, and amortized over the many missions that it enables and enhances, the imputed costs will be very reasonable. Using the approach outlined, NASA and its partners, currently the DOE, and subsequently industry, have a good chance of creating a sustained development program leading to human missions to Mars within the next few decades.

Current Development of Nuclear Thermal Propulsion technologies at the Center for Space Nuclear Research

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

Robert C. O'Brien; Steven K. Cook; Nathan D. Jerred

Nuclear power and propulsion has been considered for space applications since the 1950s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors / rocket engines in the Rover/NERVA programs1. The Aerojet Corporation was the prime contractor for the NERVA program. Modern changes in environmental laws present challenges for the redevelopment of the nuclear rocket. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel composition that is significantly different from those of the NERVA project can be engineered; this may be needed to ensure public support and compliance with safetymore » requirements. The Center for Space Nuclear Research (CSNR) is pursuing a number of technologies, modeling and testing processes to further the development of safe, practical and affordable nuclear thermal propulsion systems.« less

Irradiation of nuclear track emulsions with thermal neutrons, heavy ions, and muons

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

Artemenkov, D. A., E-mail: artemenkov@lhe.jinr.ru; Bradnova, V.; Zaitsev, A. A.

Exposures of test samples of nuclear track emulsion were analyzed. Angular and energy correlations of products originating from the thermal-neutron-induced reaction n{sub th} +{sup 10} B → {sup 7} Li + (γ)+ α were studied in nuclear track emulsions enriched in boron. Nuclear track emulsions were also irradiated with {sup 86}Kr{sup +17} and {sup 124}Xe{sup +26} ions of energy about 1.2 MeV per nucleon. Measurements of ranges of heavy ions in nuclear track emulsionsmade it possible to determine their energies on the basis of the SRIM model. The formation of high-multiplicity nuclear stars was observed upon irradiating nuclear track emulsionsmore » with ultrarelativistic muons. Kinematical features studied in this exposure of nuclear track emulsions for events of the muon-induced splitting of carbon nuclei to three alpha particles are indicative of the nucleardiffraction interaction mechanism.« less

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Safe, Affordable, Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Houts, M. G.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Doughty, G. E.

2014-01-01

The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced Nuclear Electric Propulsion (NEP).

Nuclear Thermal Propulsion for Advanced Space Exploration

NASA Technical Reports Server (NTRS)

Houts, M. G.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

2012-01-01

The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced Nuclear Electric Propulsion (NEP).

Ceramic Borehole Seals for Nuclear Waste Disposal Applications

NASA Astrophysics Data System (ADS)

Lowry, B.; Coates, K.; Wohletz, K.; Dunn, S.; Patera, E.; Duguid, A.; Arnold, B.; Zyvoloski, G.; Groven, L.; Kuramyssova, K.

2015-12-01

Sealing plugs are critical features of the deep borehole system design. They serve as structural platforms to bear the weight of the backfill column, and as seals through their low fluid permeability and bond to the borehole or casing wall. High hydrostatic and lithostatic pressures, high mineral content water, and elevated temperature due to the waste packages and geothermal gradient challenge the long term performance of seal materials. Deep borehole nuclear waste disposal faces the added requirement of assuring performance for thousands of years in large boreholes, requiring very long term chemical and physical stability. A high performance plug system is being developed which capitalizes on the energy of solid phase reactions to form a ceramic plug in-situ. Thermites are a family of self-oxidized metal/oxide reactions with very high energy content and the ability to react under water. When combined with engineered additives the product exhibits attractive structural, sealing, and corrosion properties. In the initial phase of this research, exploratory and scaled tests demonstrated formulations that achieved controlled, fine grained, homogeneous, net shape plugs composed predominantly of ceramic material. Laboratory experiments produced plug cores with confined fluid permeability as low as 100 mDarcy, compressive strength as high as 70 MPa (three times the strength of conventional well cement), with the inherent corrosion resistance and service temperature of ceramic matrices. Numerical thermal and thermal/structural analyses predicted the in-situ thermal performance of the reacted plugs, showing that they cooled to ambient temperature (and design strength) within 24 to 48 hours. The current development effort is refining the reactant formulations to achieve desired performance characteristics, developing the system design and emplacement processes to be compatible with conventional well service practices, and understanding the thermal, fluid, and structural effects the plug will have on surrounding media. This paper will report on the state of the development effort and plans for a field demonstration in early 2016 in a cased well with traditional plug seal and strength measurements.

Low-thrust chemical propulsion system propellant expulsion and thermal conditioning study. Executive summary

NASA Technical Reports Server (NTRS)

Merino, F.; Wakabayashi, I.; Pleasant, R. L.; Hill, M.

1982-01-01

Preferred techniques for providing abort pressurization and engine feed system net positive suction pressure (NPSP) for low thrust chemical propulsion systems (LTPS) were determined. A representative LTPS vehicle configuration is presented. Analysis tasks include: propellant heating analysis; pressurant requirements for abort propellant dump; and comparative analysis of pressurization techniques and thermal subcoolers.

Vibrational and thermal properties of β-HMX and TATB from dispersion corrected density functional theory

NASA Astrophysics Data System (ADS)

Landerville, Aaron C.; Oleynik, Ivan I.

2017-01-01

Dispersion Corrected Density Functional Theory (DFT+vdW) calculations are performed to predict vibrational and thermal properties of the bulk energetic materials (EMs) β-octahydrocyclotetramethylene-tetranitramine (β-HMX) and triaminotrinitrobenzene (TATB). DFT+vdW calculations of pressure-dependent crystal structure and the hydrostatic equation of state are followed by frozen-phonon calculations of their respective vibration spectra at each pressure. These are then used under the quasi-harmonic approximation to obtain zero-point and thermal free energy contributions to the pressure, resulting in pressure-volume-temperature (PVT) EOS for each material that are in excellent agreement with experiment. Heat capacities, and coefficients of thermal expansion as functions of temperature are also calculated and compared with experiment.

Diffusive mass transport in agglomerated glassy fallout from a near-surface nuclear test

NASA Astrophysics Data System (ADS)

Weisz, David G.; Jacobsen, Benjamin; Marks, Naomi E.; Knight, Kim B.; Isselhardt, Brett H.; Matzel, Jennifer E.

2018-02-01

Aerodynamically-shaped glassy fallout is formed when vapor phase constituents from the nuclear device are incorporated into molten carriers (i.e. fallout precursor materials derived from soil or other near-field environmental debris). The effects of speciation and diffusive transport of condensing constituents are not well defined in models of fallout formation. Previously we reported observations of diffuse micrometer scale layers enriched in Na, Fe, Ca, and 235U, and depleted in Al and Ti, at the interfaces of agglomerated fallout objects. Here, we derive the timescales of uranium mass transport in such fallout as it cools from 2500 K to 1500 K by applying a 1-dimensional planar diffusion model to the observed 235U/30Si variation at the interfaces. By modeling the thermal transport between the fireball and the carrier materials, the time of mass transport is calculated to be <0.6 s, <1 s, <2 s, and <3.5 s for fireball yields of 0.1 kt, 1 kt, 10 kt, and 100 kt respectively. Based on the calculated times of mass transport, a maximum temperature of deposition of uranium onto the carrier material of ∼2200 K is inferred (1σ uncertainty of ∼200 K). We also determine that the occurrence of micrometer scale layers of material enriched in relatively volatile Na-species as well as more refractory Ca-species provides evidence for an oxygen-rich fireball based on the vapor pressure of the two species under oxidizing conditions. These results represent the first application of diffusion-based modeling to derive material transport, thermal environments, and oxidation-speciation in near-surface nuclear detonation environments.

Diffusive mass transport in agglomerated glassy fallout from a near-surface nuclear test

DOE PAGES

Weisz, David G.; Jacobsen, Benjamin; Marks, Naomi E.; ...

2017-12-15

Aerodynamically-shaped glassy fallout is formed when vapor phase constituents from the nuclear device are incorporated into molten carriers (i.e. fallout precursor materials derived from soil or other near-field environmental debris). The effects of speciation and diffusive transport of condensing constituents are not well defined in models of fallout formation. Previously we reported observations of diffuse micrometer scale layers enriched in Na, Fe, Ca, and 235U, and depleted in Al and Ti, at the interfaces of agglomerated fallout objects. Here in this paper, we derive the timescales of uranium mass transport in such fallout as it cools from 2500 K tomore » 1500 K by applying a 1-dimensional planar diffusion model to the observed 235U/ 30Si variation at the interfaces. By modeling the thermal transport between the fireball and the carrier materials, the time of mass transport is calculated to be <0.6 s, <1 s, <2 s, and <3.5 s for fireball yields of 0.1 kt, 1 kt, 10 kt, and 100 kt respectively. Based on the calculated times of mass transport, a maximum temperature of deposition of uranium onto the carrier material of ~2200 K is inferred (1σ uncertainty of ~200 K). We also determine that the occurrence of micrometer scale layers of material enriched in relatively volatile Na-species as well as more refractory Ca-species provides evidence for an oxygen-rich fireball based on the vapor pressure of the two species under oxidizing conditions. These results represent the first application of diffusion-based modeling to derive material transport, thermal environments, and oxidation-speciation in near-surface nuclear detonation environments.« less

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

Diffusive mass transport in agglomerated glassy fallout from a near-surface nuclear test

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

Weisz, David G.; Jacobsen, Benjamin; Marks, Naomi E.

Aerodynamically-shaped glassy fallout is formed when vapor phase constituents from the nuclear device are incorporated into molten carriers (i.e. fallout precursor materials derived from soil or other near-field environmental debris). The effects of speciation and diffusive transport of condensing constituents are not well defined in models of fallout formation. Previously we reported observations of diffuse micrometer scale layers enriched in Na, Fe, Ca, and 235U, and depleted in Al and Ti, at the interfaces of agglomerated fallout objects. Here in this paper, we derive the timescales of uranium mass transport in such fallout as it cools from 2500 K tomore » 1500 K by applying a 1-dimensional planar diffusion model to the observed 235U/ 30Si variation at the interfaces. By modeling the thermal transport between the fireball and the carrier materials, the time of mass transport is calculated to be <0.6 s, <1 s, <2 s, and <3.5 s for fireball yields of 0.1 kt, 1 kt, 10 kt, and 100 kt respectively. Based on the calculated times of mass transport, a maximum temperature of deposition of uranium onto the carrier material of ~2200 K is inferred (1σ uncertainty of ~200 K). We also determine that the occurrence of micrometer scale layers of material enriched in relatively volatile Na-species as well as more refractory Ca-species provides evidence for an oxygen-rich fireball based on the vapor pressure of the two species under oxidizing conditions. These results represent the first application of diffusion-based modeling to derive material transport, thermal environments, and oxidation-speciation in near-surface nuclear detonation environments.« less

Thermally Driven Josephson Effect

NASA Technical Reports Server (NTRS)

Penanen, Konstantin; Chui, Talso

2008-01-01

A concept is proposed of the thermally driven Josephson effect in superfluid helium. Heretofore, the Josephson effect in a superfluid has been recognized as an oscillatory flow that arises in response to a steady pressure difference between two superfluid reservoirs separated by an array of submicron-sized orifices, which act in unison as a single Josephson junction. Analogously, the thermally driven Josephson effect is an oscillatory flow that arises in response to a steady temperature difference. The thermally driven Josephson effect is partly a consequence of a quantum- mechanical effect known as the fountain effect, in which a temperature difference in a superfluid is accompanied by a pressure difference. The thermally driven Josephson effect may have significance for the development of a high-resolution gyroscope based on the Josephson effect in a superfluid: If the pressure-driven Josephson effect were used, then the fluid on the high-pressure side would become depleted, necessitating periodic interruption of operation to reverse the pressure difference. If the thermally driven Josephson effect were used, there would be no net flow and so the oscillatory flow could be maintained indefinitely by maintaining the required slightly different temperatures on both sides of the junction.

Thermally induced processes in mixtures of aluminum with organic acids after plastic deformations under high pressure

NASA Astrophysics Data System (ADS)

Zhorin, V. A.; Kiselev, M. R.; Roldugin, V. I.

2017-11-01

DSC is used to measure the thermal effects of processes in mixtures of solid organic dibasic acids with powdered aluminum, subjected to plastic deformation under pressures in the range of 0.5-4.0 GPa using an anvil-type high-pressure setup. Analysis of thermograms obtained for the samples after plastic deformation suggests a correlation between the exothermal peaks observed around the temperatures of degradation of the acids and the thermally induced chemical reactions between products of acid degradation and freshly formed surfaces of aluminum particles. The release of heat in the mixtures begins at 30-40°C. The thermal effects in the mixtures of different acids change according to the order of acid reactivity in solutions. The extreme baric dependences of enthalpies of thermal effects are associated with the rearrangement of the electron subsystem of aluminum upon plastic deformation at high pressures.

Mechanical properties and microstructure of long term thermal aged WWER 440 RPV steel

NASA Astrophysics Data System (ADS)

Kolluri, M.; Kryukov, A.; Magielsen, A. J.; Hähner, P.; Petrosyan, V.; Sevikyan, G.; Szaraz, Z.

2017-04-01

The integrity assessment of the Reactor Pressure Vessel (RPV) is essential for the safe and Long Term Operation (LTO) of a Nuclear Power Plant (NPP). Hardening and embrittlement of RPV caused by neutron irradiation and thermal ageing are main reasons for mechanical properties degradation during the operation of an NPP. The thermal ageing-induced degradation of RPV steels becomes more significant with extended operational lives of NPPs. Consequently, the evaluation of thermal ageing effects is important for the structural integrity assessments required for the lifetime extension of NPPs. As a part of NRG's research programme on Structural Materials for safe-LTO of Light Water Reactor (LWR) RPVs, WWER-440 surveillance specimens, which have been thermal aged for 27 years (∼200,000 h) at 290 °C in a surveillance channel of Armenian-NPP, are investigated. Results from the mechanical and microstructural examination of these thermal aged specimens are presented in this article. The results indicate the absence of significant long term thermal ageing effect of 15Cr2MoV-A steel. No age hardening was detected in aged tensile specimens compared with the as-received condition. There is no difference between the impact properties of as-received and thermal aged weld metals. The upper shelf energy of the aged steel remains the same as for the as-received material at a rather high level of about 120 J. The T41 value did not change and was found to be about 10 °C. The microstructure of thermal aged weld, consisting carbides, carbonitrides and manganese-silicon inclusions, did not change significantly compared to as-received state. Grain-boundary segregation of phosphorus in long term aged weld is not significant either which has been confirmed by the absence of intergranular fracture increase in the weld. Negligible hardening and embrittlement observed after such long term thermal ageing is attributed to the optimum chemical composition of 15Cr2MoV-A for high thermal stability.

Activating the nuclear piston mechanism of 3D migration in tumor cells

PubMed Central

2017-01-01

Primary human fibroblasts have the remarkable ability to use their nucleus like a piston, switching from low- to high-pressure protrusions in response to the surrounding three-dimensional (3D) matrix. Although migrating tumor cells can also change how they migrate in response to the 3D matrix, it is not clear if they can switch between high- and low-pressure protrusions like primary fibroblasts. We report that unlike primary fibroblasts, the nuclear piston is not active in fibrosarcoma cells. Protease inhibition rescued the nuclear piston mechanism in polarized HT1080 and SW684 cells and generated compartmentalized pressure. Achieving compartmentalized pressure required the nucleoskeleton–cytoskeleton linker protein nesprin 3, actomyosin contractility, and integrin-mediated adhesion, consistent with lobopodia-based fibroblast migration. In addition, this activation of the nuclear piston mechanism slowed the 3D movement of HT1080 cells. Together, these data indicate that inhibiting protease activity during polarized tumor cell 3D migration is sufficient to restore the nuclear piston migration mechanism with compartmentalized pressure characteristic of nonmalignant cells. PMID:27998990

Non-thermal pressure in the outskirts of Abell 2142

NASA Astrophysics Data System (ADS)

Fusco-Femiano, Roberto; Lapi, Andrea

2018-03-01

Clumping and turbulence are expected to affect the matter accreted on to the outskirts of galaxy clusters. To determine their impact on the thermodynamic properties of Abell 2142, we perform an analysis of the X-ray temperature data from XMM-Newton via our SuperModel, a state-of-the-art tool for investigating the astrophysics of the intracluster medium already tested on many individual clusters (since Cavaliere, Lapi & Fusco-Femiano 2009). Using the gas density profile corrected for clumpiness derived by Tchernin et al. (2016), we find evidence for the presence of a non-thermal pressure component required to sustain gravity in the cluster outskirts of Abell 2142, that amounts to about 30 per cent of the total pressure at the virial radius. The presence of the non-thermal component implies the gas fraction to be consistent with the universal value at the virial radius and the electron thermal pressure profile to be in good agreement with that inferred from the SZ data. Our results indicate that the presence of gas clumping and of a non-thermal pressure component are both necessary to recover the observed physical properties in the cluster outskirts. Moreover, we stress that an alternative method often exploited in the literature (included Abell 2142) to determine the temperature profile kBT = Pe/ne basing on a combination of the Sunyaev-Zel'dovich (SZ) pressure Pe and of the X-ray electron density ne does not allow us to highlight the presence of non-thermal pressure support in the cluster outskirts.

Multidecadal global cooling and unprecedented ozone loss following a regional nuclear conflict

NASA Astrophysics Data System (ADS)

Mills, Michael J.; Toon, Owen B.; Lee-Taylor, Julia; Robock, Alan

2014-04-01

We present the first study of the global impacts of a regional nuclear war with an Earth system model including atmospheric chemistry, ocean dynamics, and interactive sea ice and land components. A limited, regional nuclear war between India and Pakistan in which each side detonates 50 15 kt weapons could produce about 5 Tg of black carbon (BC). This would self-loft to the stratosphere, where it would spread globally, producing a sudden drop in surface temperatures and intense heating of the stratosphere. Using the Community Earth System Model with the Whole Atmosphere Community Climate Model, we calculate an e-folding time of 8.7 years for stratospheric BC compared to 4-6.5 years for previous studies. Our calculations show that global ozone losses of 20%-50% over populated areas, levels unprecedented in human history, would accompany the coldest average surface temperatures in the last 1000 years. We calculate summer enhancements in UV indices of 30%-80% over midlatitudes, suggesting widespread damage to human health, agriculture, and terrestrial and aquatic ecosystems. Killing frosts would reduce growing seasons by 10-40 days per year for 5 years. Surface temperatures would be reduced for more than 25 years due to thermal inertia and albedo effects in the ocean and expanded sea ice. The combined cooling and enhanced UV would put significant pressures on global food supplies and could trigger a global nuclear famine. Knowledge of the impacts of 100 small nuclear weapons should motivate the elimination of more than 17,000 nuclear weapons that exist today.

Factors defining value and direction of thermal pressure between the mine shafts and impact of the general mine natural draught on ventilation process of underground mining companies

NASA Astrophysics Data System (ADS)

Nikolaev, A. V.; Alymenko, N. I.; Kamenskikh, A. A.; Alymenko, D. N.; Nikolaev, V. A.; Petrov, A. I.

2017-10-01

The article specifies measuring data of air parameters and its volume flow in the shafts and on the surface, collected in BKPRU-2 (Berezniki potash plant and mine 2) («Uralkali» PJSC) in normal operation mode, after shutdown of the main mine fan (GVU) and within several hours. As a result of the test it has been established that thermal pressure between the mine shafts is active continuously regardless of the GVU operation mode or other draught sources. Also it has been discovered that depth of the mine shafts has no impact on thermal pressure value. By the same difference of shaft elevation marks and parameters of outer air between the shafts, by their different depth, thermal pressure of the same value will be active. Value of the general mine natural draught defined as an algebraic sum of thermal pressure values between the shafts depends only on the difference of temperature and pressure of outer air and air in the shaft bottoms on condition of shutdown of the air handling system (unit-heaters, air conditioning systems).

Bonded foil pressure transducers

NASA Astrophysics Data System (ADS)

Daube, Bernie W.

The design of bonded-foil pressure transducers is discussed, with consideration given to individual components of both the electrical and the mechanical sections of the bonded-foil pressure transducers, as well as to the temperature control and the accuracy specification of these devices. Particular attention is given to applications of bonded foil pressure transducers, which include solid and liquid rocket engine testing for fuel and exhaust pressures, fuel and oil pressure monitoring on jet engines, and nuclear underground safety system pressure monitoring and nuclear test monitoring. A diagram of a transducer cutaway view is included.

Space disposal of nuclear wastes

NASA Technical Reports Server (NTRS)

Priest, C. C.; Nixon, R. F.; Rice, E. E.

1980-01-01

The DOE has been studying several options for nuclear waste disposal, among them space disposal, which NASA has been assessing. Attention is given to space disposal destinations noting that a circular heliocentric orbit about halfway between Earth and Venus is the reference option in space disposal studies. Discussion also covers the waste form, showing that parameters to be considered include high waste loading, high thermal conductivity, thermochemical stability, resistance to leaching, fabrication, resistance to oxidation and to thermal shock. Finally, the Space Shuttle nuclear waste disposal mission profile is presented.

Nuclear thermal rocket workshop reference system Rover/NERVA

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1991-01-01

The Rover/NERVA engine system is to be used as a reference, against which each of the other concepts presented in the workshop will be compared. The following topics are reviewed: the operational characteristics of the nuclear thermal rocket (NTR); the accomplishments of the Rover/NERVA programs; and performance characteristics of the NERVA-type systems for both Mars and lunar mission applications. Also, the issues of ground testing, NTR safety, NASA's nuclear propulsion project plans, and NTR development cost estimates are briefly discussed.

Near-zero thermal expansion in magnetically ordered state in dysprosium at high pressures and low temperatures

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

Hope, Kevin M.; Samudrala, Gopi K.; Vohra, Yogesh K.

The atomic volume of rare earth metal Dysprosium (Dy) has been measured up to high pressures of 35 GPa and low temperatures between 200 K and 7 K in a diamond anvil cell using angle dispersive x-ray diffraction at a synchrotron source. The hexagonal close-packed (hcp), alpha-Samarium (α-Sm), and double hexagonal close packed (dhcp) phases are observed to be stable in Dy under high-pressure and low-temperature conditions achieved in our experiments. Dy is known to undergo magnetic ordering below 176 K at ambient pressure with magnetic ordering Néel temperature (T N) that changes rapidly with increasing pressure. Our experimental measurementmore » shows that Dy has near-zero thermal expansion in the magnetically ordered state and normal thermal expansion in the paramagnetic state for all the three known high pressure phases (hcp, α-Sm, and dhcp) to 35 GPa. This near-zero thermal expansion behavior in Dy is observed below the magnetic ordering temperature T N at all pressures up to 35 GPa.« less

Near-zero thermal expansion in magnetically ordered state in dysprosium at high pressures and low temperatures

DOE PAGES

Hope, Kevin M.; Samudrala, Gopi K.; Vohra, Yogesh K.

2017-01-01

The atomic volume of rare earth metal Dysprosium (Dy) has been measured up to high pressures of 35 GPa and low temperatures between 200 K and 7 K in a diamond anvil cell using angle dispersive x-ray diffraction at a synchrotron source. The hexagonal close-packed (hcp), alpha-Samarium (α-Sm), and double hexagonal close packed (dhcp) phases are observed to be stable in Dy under high-pressure and low-temperature conditions achieved in our experiments. Dy is known to undergo magnetic ordering below 176 K at ambient pressure with magnetic ordering Néel temperature (T N) that changes rapidly with increasing pressure. Our experimental measurementmore » shows that Dy has near-zero thermal expansion in the magnetically ordered state and normal thermal expansion in the paramagnetic state for all the three known high pressure phases (hcp, α-Sm, and dhcp) to 35 GPa. This near-zero thermal expansion behavior in Dy is observed below the magnetic ordering temperature T N at all pressures up to 35 GPa.« less

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Evaluation of critical nuclear power plant electrical cable response to severe thermal fire conditions

NASA Astrophysics Data System (ADS)

Taylor, Gabriel James

The failure of electrical cables exposed to severe thermal fire conditions are a safety concern for operating commercial nuclear power plants (NPPs). The Nuclear Regulatory Commission (NRC) has promoted the use of risk-informed and performance-based methods for fire protection which resulted in a need to develop realistic methods to quantify the risk of fire to NPP safety. Recent electrical cable testing has been conducted to provide empirical data on the failure modes and likelihood of fire-induced damage. This thesis evaluated numerous aspects of the data. Circuit characteristics affecting fire-induced electrical cable failure modes have been evaluated. In addition, thermal failure temperatures corresponding to cable functional failures have been evaluated to develop realistic single point thermal failure thresholds and probability distributions for specific cable insulation types. Finally, the data was used to evaluate the prediction capabilities of a one-dimension conductive heat transfer model used to predict cable failure.

Modeling and Simulation of a Nuclear Fuel Element Test Section

NASA Technical Reports Server (NTRS)

Moran, Robert P.; Emrich, William

2011-01-01

"The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

Anorthite: Thermal equation of state to high pressures. [for comparison with Earth interior and cratering properties of lunar surface

NASA Technical Reports Server (NTRS)

Jeanloz, R.; Ahrens, T. J.

1979-01-01

The shock wave (Hugoniot) data on single crystal and porous anorthite (CaAl2Si208) to pressures of 120 GPa are presented. These data are inverted to yield high pressure values of the Grueneisen parameter, adiabatic bulk modulus, and coefficient of thermal expansion over a broad range of pressures and temperatures which in turn are used to reduce the raw Hugoniot data and construct an experimentally based, high pressure thermal equation of state for anorthite. The hypothesis that higher order anharmonic contributions to the thermal properties decrease more rapidly upon compression than the lowest order anharmonicities is supported. The properties of anorthite corrected to lower mantle conditions show that although the density of anorthite is comparable to that of the lower most mantle, its bulk modulus is considerably less, hence making enrichment in the mantle implausible except perhaps near its base.

Horizontal baffle for nuclear reactors

DOEpatents

Rylatt, John A.

1978-01-01

A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

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

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

30 CFR 250.521 - How do I manage the thermal effects caused by initial production on a newly completed or...

Code of Federal Regulations, 2011 CFR

2011-07-01

... 30 Mineral Resources 2 2011-07-01 2011-07-01 false How do I manage the thermal effects caused by... Casing Pressure Management § 250.521 How do I manage the thermal effects caused by initial production on... a normal and necessary operation to manage thermal casing pressure; therefore, you do not need to...

An FBG Optical Approach to Thermal Expansion Measurements under Hydrostatic Pressure.

PubMed

Rosa, Priscila F S; Thomas, Sean M; Balakirev, Fedor F; Betts, Jon; Seo, Soonbeom; Bauer, Eric D; Thompson, Joe D; Jaime, Marcelo

2017-11-04

We report on an optical technique for measuring thermal expansion and magnetostriction at cryogenic temperatures and under applied hydrostatic pressures of 2.0 GPa. Optical fiber Bragg gratings inside a clamp-type pressure chamber are used to measure the strain in a millimeter-sized sample of CeRhIn₅. We describe the simultaneous measurement of two Bragg gratings in a single optical fiber using an optical sensing instrument capable of resolving changes in length [dL/L = (L- L₀)/L₀] on the order of 10 -7 . Our results demonstrate the possibility of performing high-resolution thermal expansion measurements under hydrostatic pressure, a capability previously hindered by the small working volumes typical of pressure cells.

Production of stable, non-thermal atmospheric pressure rf capacitive plasmas using gases other than helium or neon

DOEpatents

Park, Jaeyoung; Henins, Ivars

2005-06-21

The present invention enables the production of stable, steady state, non-thermal atmospheric pressure rf capacitive .alpha.-mode plasmas using gases other than helium and neon. In particular, the current invention generates and maintains stable, steady-state, non-thermal atmospheric pressure rf .alpha.-mode plasmas using pure argon or argon with reactive gas mixtures, pure oxygen or air. By replacing rare and expensive helium with more readily available gases, this invention makes it more economical to use atmospheric pressure rf .alpha.-mode plasmas for various materials processing applications.

Mass Dependency of Isotope Fractionation of Gases Under Thermal Gradient and Its Possible Implications for Planetary Atmosphere Escaping Process

NASA Technical Reports Server (NTRS)

Sun, Tao; Niles, Paul; Bao, Huiming; Socki, Richard

2014-01-01

Physical processes that unmix elements/isotopes of gas molecules involve phase changes, diffusion (chemical or thermal), effusion and gravitational settling. Some of those play significant roles for the evolution of chemical and isotopic compositions of gases in planetary bodies which lead to better understanding of surface paleoclimatic conditions, e.g. gas bubbles in Antarctic ice, and planetary evolution, e.g. the solar-wind erosion induced gas escaping from exosphere on terrestrial planets.. A mass dependent relationship is always expected for the kinetic isotope fractionations during these simple physical processes, according to the kinetic theory of gases by Chapman, Enskog and others [3-5]. For O-bearing (O16, -O17, -O18) molecules the alpha O-17/ alpha O-18 is expected at 0.5 to 0.515, and for S-bearing (S32,-S33. -S34, -S36) molecules, the alpha S-33/ alpha S-34 is expected at 0.5 to 0.508, where alpha is the isotope fractionation factor associated with unmixing processes. Thus, one isotope pair is generally proxied to yield all the information for the physical history of the gases. However, we recently] reported the violation of mass law for isotope fractionation among isotope pairs of multiple isotope system during gas diffusion or convection under thermal gradient (Thermal Gradient Induced Non-Mass Dependent effect, TGI-NMD). The mechanism(s) that is responsible to such striking observation remains unanswered. In our past studies, we investigated polyatomic molecules, O2 and SF6, and we suggested that nuclear spin effect could be responsible to the observed NMD effect in a way of changing diffusion coefficients of certain molecules, owing to the fact of negligible delta S-36 anomaly for SF6.. On the other hand, our results also showed that for both diffusion and convection under thermal gradient, this NMD effect is increased by lower gas pressure, bigger temperature gradient and lower average temperature, which indicate that the nuclear spin effect may not be the significant contributor as the energies involved in the hyperfine effect are much smaller than those with molecular collisions, especially under convective conditions.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

NASA Astrophysics Data System (ADS)

Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

2016-12-01

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

Peculiarities of FeSi phonon spectrum induced by a change of atomic volume

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

Parshin, P. P., E-mail: Parshin-PP@nrcki.ru, E-mail: neupar45@yandex.ru; Chumakov, A. I.; Alekseev, P. A.

2016-12-15

We analyze in detail the results of experimental investigations of the evolution of the thermal vibration spectra for iron atoms in iron monosilicide FeSi depending on two external parameters, viz., temperature T (in the range 46–297 K at pressure P = 0.1 MPa) and pressure P (in the range 0.1 MPa–43 GPa at temperature T = 297 K), obtained by nuclear inelastic scattering of synchrotron radiation. The decrease of the atomic volume is accompanied by a rearrangement of the phonon spectrum, which is manifested, in particular, in the splitting of the low-energy peak in the spectrum and in an increasemore » of the energy for all phonons. The changes of the average energy of the iron atom vibrational spectrum and of the Debye energy with decreasing atomic volume are analyzed. Different versions of FeSi electron spectrum variation, which can be used to explain the observed phonon anomalies, are considered.« less

2003 NASA Seal/Secondary Air System Workshop. Volume 1

NASA Technical Reports Server (NTRS)

Steinetz, Bruce M. (Editor); Hendricks, Robert C. (Editor)

2004-01-01

The following reports were included in the 2003 NASA Seal/Secondary Air System Workshop:Low Emissions Alternative Power (LEAP); Overview of NASA Glenn Seal Developments; NASA Ultra Efficient Engine Technology Project Overview; Development of Higher Temperature Abradable Seals for Industrial Gas Turbines; High Misalignment Carbon Seals for the Fan Drive Gear System Technologies; Compliant Foil Seal Investigations; Test Rig for Evaluating Active Turbine Blade Tip Clearance Control Concepts; Controls Considerations for Turbine Active Clearance Control; Non-Contacting Finger Seal Developments and Design Considerations; Effect of Flow-Induced Radial Load on Brush Seal/Rotor Contact Mechanics; Seal Developments at Flowserve Corporation; Investigations of High Pressure Acoustic Waves in Resonators With Seal-Like Features; Numerical Investigations of High Pressure Acoustic Waves in Resonators; Feltmetal Seal Material Through-Flow; "Bimodal" Nuclear Thermal Rocket (BNTR) Propulsion for Future Human Mars Exploration Missions; High Temperature Propulsion System Structural Seals for Future Space Launch Vehicles; Advanced Control Surface Seal Development for Future Space Vehicles; High Temperature Metallic Seal Development for Aero Propulsion and Gas Turbine Applications; and BrazeFoil Honeycomb.

High-Pressure Geophysical Properties of Fcc Phase FeHX

NASA Astrophysics Data System (ADS)

Thompson, E. C.; Davis, A. H.; Bi, W.; Zhao, J.; Alp, E. E.; Zhang, D.; Greenberg, E.; Prakapenka, V. B.; Campbell, A. J.

2018-01-01

Face centered cubic (fcc) FeHX was synthesized at pressures of 18-68 GPa and temperatures exceeding 1,500 K. Thermally quenched samples were evaluated using synchrotron X-ray diffraction (XRD) and nuclear resonant inelastic X-ray scattering (NRIXS) to determine sample composition and sound velocities to 82 GPa. To aid in the interpretation of nonideal (X ≠ 1) stoichiometries, two equations of state for fcc FeHX were developed, combining an empirical equation of state for iron with two distinct synthetic compression curves for interstitial hydrogen. Matching the density deficit of the Earth's core using these equations of state requires 0.8-1.1 wt % hydrogen at the core-mantle boundary and 0.2-0.3 wt % hydrogen at the interface of the inner and outer cores. Furthermore, a comparison of Preliminary Reference Earth Model (PREM) to a Birch's law extrapolation of our experimental results suggests that an iron alloy containing ˜0.8-1.3 wt % hydrogen could reproduce both the density and compressional velocity (VP) of the Earth's outer core.

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

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

Yun, D.; Taiwo, T. A.; Kim, T. K.

2010-10-01

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluatemore » the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.« less

Pulse thermal energy transport/storage system

DOEpatents

Weislogel, Mark M.

1992-07-07

A pulse-thermal pump having a novel fluid flow wherein heat admitted to a closed system raises the pressure in a closed evaporator chamber while another interconnected evaporator chamber remains open. This creates a large pressure differential, and at a predetermined pressure the closed evaporator is opened and the opened evaporator is closed. This difference in pressure initiates fluid flow in the system.

Investigation of natural circulation instability and transients in passively safe novel modular reactor

NASA Astrophysics Data System (ADS)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal-hydraulic and nuclear coupled startup transients are performed to investigate the flow instabilities at low pressure and low power conditions. Two different power ramps are chosen to study the effect of power density on the flow instability. The experimental startup transient tests show the existence of three different flow instability mechanisms during the low pressure startup transients, i.e., flashing instability, condensation induced instability, and density wave oscillations. Flashing instability in the chimney section of the test loop and density wave oscillation are the main flow instabilities observed when the system pressure is below 0.5 MPa. They show completely different type of oscillations, i.e., intermittent oscillation and sinusoidal oscillation, in void fraction profile during the startup transients. In order to perform nuclear-coupled startup transients with void reactivity feedback, the Point Kinetics model is utilized to calculate the transient power during the startup transients. In addition, the differences between the electric resistance heaters and typical fuel element are taken into account. The reactor power calculated shows some oscillations due to flashing instability during the transients. However, the void reactivity feedback does not have significant influence on the flow instability during the startup procedure for the NMR-50. Further investigation of very small power ramp on the startup transients is carried out for the thermal-hydraulic startup transients. It is found that very small power density can eliminate the flashing oscillation in the single phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. Furthermore, initially pressurized startup procedure is investigated to eliminate the main flow instabilities. The results show that the pressurized startup procedure can suppress the flashing instability at low pressure and low power conditions. In order to have a deep understanding of natural circulation flow instability, the quasi-steady tests are performed using the test facility installed with preheater and subcooler. The effects of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback are investigated in the quasi-steady state tests. The stability boundaries are determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. In order to predict the stability boundary theoretically, linear stability analysis in the frequency domain is performed at four sections of the loop. The flashing in the chimney is considered as an axially uniform heat source. The dimensionless characteristic equation of the pressure drop perturbation is obtained by considering the void fraction effect and outlet flow resistance in the chimney section. The flashing boundary shows some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium is recommended to improve the accuracy of flashing instability boundary.

A Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications

NASA Astrophysics Data System (ADS)

Grunloh, Timothy P.

The objective of this dissertation is to develop a 3-D domain-overlapping coupling method that leverages the superior flow field resolution of the Computational Fluid Dynamics (CFD) code STAR-CCM+ and the fast execution of the System Thermal Hydraulic (STH) code TRACE to efficiently and accurately model thermal hydraulic transport properties in nuclear power plants under complex conditions of regulatory and economic importance. The primary contribution is the novel Stabilized Inertial Domain Overlapping (SIDO) coupling method, which allows for on-the-fly correction of TRACE solutions for local pressures and velocity profiles inside multi-dimensional regions based on the results of the CFD simulation. The method is found to outperform the more frequently-used domain decomposition coupling methods. An STH code such as TRACE is designed to simulate large, diverse component networks, requiring simplifications to the fluid flow equations for reasonable execution times. Empirical correlations are therefore required for many sub-grid processes. The coarse grids used by TRACE diminish sensitivity to small scale geometric details such as Reactor Pressure Vessel (RPV) internals. A CFD code such as STAR-CCM+ uses much finer computational meshes that are sensitive to the geometric details of reactor internals. In turbulent flows, it is infeasible to fully resolve the flow solution, but the correlations used to model turbulence are at a low level. The CFD code can therefore resolve smaller scale flow processes. The development of a 3-D coupling method was carried out with the intention of improving predictive capabilities of transport properties in the downcomer and lower plenum regions of an RPV in reactor safety calculations. These regions are responsible for the multi-dimensional mixing effects that determine the distribution at the core inlet of quantities with reactivity implications, such as fluid temperature and dissolved neutron absorber concentration.

Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

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

Weigl, M.

2012-07-01

Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactormore » with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible. The successful completion of two big D and D projects (HDR and KKN), which reached green field conditions, are showing quite the contrary. Moreover, research on D and D technologies offers the possibility to educate students on a field of nuclear technology, which will be very important in the future. In these days D and D companies are seeking for a lot of young engineers and this will not change in the coming years. (authors)« less

Laser diagnostics for NTP fuel corrosion studies

NASA Technical Reports Server (NTRS)

Wantuck, Paul J.; Butt, D. P.; Sappey, A. D.

1993-01-01

Viewgraphs and explanations on laser diagnostics for nuclear thermal propulsion (NTP) fuel corrosion studies are presented. Topics covered include: NTP fuels; U-Zr-C system corrosion products; planar laser-induced fluorescence (PLIF); utilization of PLIF for corrosion product characterization of nuclear thermal rocket fuel elements under test; ZrC emission spectrum; and PLIF imaging of ZrC plume.

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

Cycle Trades for Nuclear Thermal Rocket Propulsion Systems

NASA Technical Reports Server (NTRS)

White, C.; Guidos, M.; Greene, W.

2003-01-01

Nuclear fission has been used as a reliable source for utility power in the United States for decades. Even in the 1940's, long before the United States had a viable space program, the theoretical benefits of nuclear power as applied to space travel were being explored. These benefits include long-life operation and high performance, particularly in the form of vehicle power density, enabling longer-lasting space missions. The configurations for nuclear rocket systems and chemical rocket systems are similar except that a nuclear rocket utilizes a fission reactor as its heat source. This thermal energy can be utilized directly to heat propellants that are then accelerated through a nozzle to generate thrust or it can be used as part of an electricity generation system. The former approach is Nuclear Thermal Propulsion (NTP) and the latter is Nuclear Electric Propulsion (NEP), which is then used to power thruster technologies such as ion thrusters. This paper will explore a number of indirect-NTP engine cycle configurations using assumed performance constraints and requirements, discuss the advantages and disadvantages of each cycle configuration, and present preliminary performance and size results. This paper is intended to lay the groundwork for future efforts in the development of a practical NTP system or a combined NTP/NEP hybrid system.

Thermal-barrier coatings for utility gas turbines

NASA Technical Reports Server (NTRS)

Levine, S. R.; Miller, R. A.

1982-01-01

The potential of thermal barrier coatings for use in utility gas turbines was assessed. Pressurized passage and ambient pressure doped fuel burner rig tests revealed that thermal barrier coatings are not resistant to dirty combustion environments. However, present thermal barrier coatings, such as duplex partially stabilized zirconia and duplex Ca2SiO4 have ample resistance to the thermo-mechanical stress and temperature levels anticipated for heavy duty gas turbines firing clean fuel as revealed by clean fuel pressurized passage and ambient pressure burner rig tests. Thus, it is appropriate to evaluate such coatings on blades, vanes and combustors in the field. However, such field tests should be backed up with adequate effort in the areas of coating application technology and design analysis so that the field tests yield unequivocal results.

Coherent thermodynamic model for solid, liquid and gas phases of elements and simple compounds in wide ranges of pressure and temperature

NASA Astrophysics Data System (ADS)

Holzapfel, Wilfried B.

2018-06-01

Thermodynamic modeling of fluids (liquids and gases) uses mostly series expansions which diverge at low temperatures and do not fit to the behavior of metastable quenched fluids (amorphous, glass like solids). These divergences are removed in the present approach by the use of reasonable forms for the "cold" potential energy and for the thermal pressure of the fluid system. Both terms are related to the potential energy and to the thermal pressure of the crystalline phase in a coherent way, which leads to simpler and non diverging series expansions for the thermal pressure and thermal energy of the fluid system. Data for solid and fluid argon are used to illustrate the potential of the present approach.

Extension of the thermal porosimetry method to high gas pressure for nanoporosimetry estimation

NASA Astrophysics Data System (ADS)

Jannot, Y.; Degiovanni, A.; Camus, M.

2018-04-01

Standard pore size determination methods like mercury porosimetry, nitrogen sorption, microscopy, or X-ray tomography are not suited to highly porous, low density, and thus very fragile materials. For this kind of materials, a method based on thermal characterization has been developed in a previous study. This method has been used with air pressure varying from 10-1 to 105 Pa for materials having a thermal conductivity less than 0.05 W m-1 K-1 at atmospheric pressure. It enables the estimation of pore size distribution between 100 nm and 1 mm. In this paper, we present a new experimental device enabling thermal conductivity measurement under gas pressure up to 106 Pa, enabling the estimation of the volume fraction of pores having a 10 nm diameter. It is also demonstrated that the main thermal conductivity models (parallel, series, Maxwell, Bruggeman, self-consistent) lead to the same estimation of the pore size distribution as the extended parallel model (EPM) presented in this paper and then used to process the experimental data. Three materials with thermal conductivities at atmospheric pressure ranging from 0.014 W m-1 K-1 to 0.04 W m-1 K-1 are studied. The thermal conductivity measurement results obtained with the three materials are presented, and the corresponding pore size distributions between 10 nm and 1 mm are presented and discussed.

Thermal conductivity of tungsten: Effects of plasma-related structural defects from molecular-dynamics simulations

NASA Astrophysics Data System (ADS)

Hu, Lin; Wirth, Brian D.; Maroudas, Dimitrios

2017-08-01

We report results on the lattice thermal conductivities of tungsten single crystals containing nanoscale-sized pores or voids and helium (He) nanobubbles as a function of void/bubble size and gas pressure in the He bubbles based on molecular-dynamics simulations. For reference, we calculated lattice thermal conductivities of perfect tungsten single crystals along different crystallographic directions at room temperature and found them to be about 10% of the overall thermal conductivity of tungsten with a weak dependence on the heat flux direction. The presence of nanoscale voids in the crystal causes a significant reduction in its lattice thermal conductivity, which decreases with increasing void size. Filling the voids with He to form He nanobubbles and increasing the bubble pressure leads to further significant reduction of the tungsten lattice thermal conductivity, down to ˜20% of that of the perfect crystal. The anisotropy in heat conduction remains weak for tungsten single crystals containing nanoscale-sized voids and He nanobubbles throughout the pressure range examined. Analysis of the pressure and atomic displacement fields in the crystalline region that surrounds the He nanobubbles reveals that the significant reduction of tungsten lattice thermal conductivity in this region is due to phonon scattering from the nanobubbles, as well as lattice deformation around the nanobubbles and formation of lattice imperfections at higher bubble pressure.

Nuclear Thermal Propulsion (NTP) Development Activities at the NASA Marshall Space Flight Center - 2006 Accomplishments

NASA Technical Reports Server (NTRS)

Ballard, Richard O.

2007-01-01

In 2005-06, the Prometheus program funded a number of tasks at the NASA-Marshall Space Flight Center (MSFC) to support development of a Nuclear Thermal Propulsion (NTP) system for future manned exploration missions. These tasks include the following: 1. NTP Design Develop Test & Evaluate (DDT&E) Planning 2. NTP Mission & Systems Analysis / Stage Concepts & Engine Requirements 3. NTP Engine System Trade Space Analysis and Studies 4. NTP Engine Ground Test Facility Assessment 5. Non-Nuclear Environmental Simulator (NTREES) 6. Non-Nuclear Materials Fabrication & Evaluation 7. Multi-Physics TCA Modeling. This presentation is a overview of these tasks and their accomplishments

NASA's Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Mitchell, Doyce P.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Gerrish, Harold P.; Doughty, Glen; Belvin, Anthony; Clement, Steven; Borowski, Stanley K.;

Measurements of mineral thermal conductivity at high pressures and temperatures with the laser-heated diamond anvil cell

NASA Astrophysics Data System (ADS)

McGuire, C. P.; Rainey, E.; Kavner, A.

2016-12-01

The high-pressure, high-temperature thermal conductivities of lower mantle oxides and silicates play an important role in governing the heat flow across the core-mantle boundary, and the thermal conductivity of core materials determines, at first order, the power required to run the geodynamo. Uncertainties in the pressure-dependence and compositional-dependence of thermal conductivities has complicated our understanding of the heat flow in the deep earth and has implications for the geodynamo mechanism (Buffett, 2012). The goal of this study is to measure how thermal conductivity varies with pressure and composition using a technique that combines temperature measurements as a function of power input in the laser-heated diamond anvil cell (LHDAC) with a model of three-dimensional heat flow (Rainey & Kavner, 2014). In one set of experiments, we measured temperature versus laser-power for iron, iron silicide, and stainless steel (Fe:Cr:Ni = 70:19:11 wt%), using a variety of insulating layers. In another set of experiments, we measured temperature vs. laser power for a series of Fe-bearing periclase (Mg1-x,FexO) samples, with compositions ranging from x = .24 to x = .78. These experiments were conducted up to pressures of 25 GPa and temperatures of 2800 K. A numerical model for heat conduction in the LHDAC is used to forward model the temperature versus laser power curves at successive pressures, solving for the change in thermal conductivity of the material required to best reproduce the measurements. The heat flow model is implemented using a finite element full-approximation storage (FAS) multi-grid solver, which allows for efficient computation with flexible inputs for geometry and material properties in the diamond anvil cell (Rainey et al., 2013). We use the results of our experiments and model to extract pressure and compositional dependencies of thermal conductivity for the materials described herein. The results are used to help constrain models of the thermal properties of core and mantle materials.

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

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

Feldkamp, Martin; Nehrig, Marko; Bletzer, Claus

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues ismore » provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)« less

Shock tube studies of thermal radiation of diesel-spray combustion under a range of spray conditions

NASA Astrophysics Data System (ADS)

Tsuboi, T.; Kurihara, Y.; Takasaki, M.; Katoh, R.; Ishii, K.

2007-05-01

A tailored interface shock tube and an over-tailored interface shock tube were used to measure the thermal energy radiated during diesel-spray combustion of light oil, α-methylnaphthalene and cetane by changing the injection pressure. The ignition delay of methanol and the thermal radiation were also measured. Experiments were performed in a steel shock tube with a 7 m low-pressure section filled with air and a 6 m high-pressure section. Pre-compressed fuel was injected through a throttle nozzle into air behind a reflected shock wave. Monochromatic emissive power and the power emitted across all infrared wavelengths were measured with IR-detectors set along the central axis of the tube. Time-dependent radii where soot particles radiated were also determined, and the results were as follows. For diesel spray combustion with high injection pressures (from 10 to 80 MPa), the thermal radiation energy of light oil per injection increased with injection pressure from 10 to 30 MPa. The energy was about 2% of the heat of combustion of light oil at P inj = about 30 MPa. At injection pressure above 30 MPa the thermal radiation decreased with increasing injection pressure. This profile agreed well with the combustion duration, the flame length, the maximum amount of soot in the flame, the time-integrated soot volume and the time-integrated flame volume. The ignition delay of light oil was observed to decrease monotonically with increasing fuel injection pressure. For diesel spray combustion of methanol, the thermal radiation including that due to the gas phase was 1% of the combustion heat at maximum, and usually lower than 1%. The thermal radiation due to soot was lower than 0.05% of the combustion heat. The ignition delays were larger (about 50%) than those of light oil. However, these differences were within experimental error.

Effects of Nuclear Weapons.

ERIC Educational Resources Information Center

Sartori, Leo

1983-01-01

Fundamental principles governing nuclear explosions and their effects are discussed, including three components of a nuclear explosion (thermal radiation, shock wave, nuclear radiation). Describes how effects of these components depend on the weapon's yield, its height of burst, and distance of detonation point. Includes effects of three…

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Structure–property relations of orthorhombic [(CH{sub 3}){sub 3}NCH{sub 2}COO]{sub 2}(CuCl{sub 2}){sub 3}·2H{sub 2}O

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

Haussühl, Eiken, E-mail: haussuehl@kristall.uni-frankfurt.de; Schreuer, Jürgen; Wiehl, Leonore

2014-04-01

Large single crystals of orthorhombic [(CH{sub 3}){sub 3}NCH{sub 2}COO]{sub 2}(CuCl{sub 2}){sub 3}·2H{sub 2}O with dimensions up to 40×40×30 mm{sup 3} were grown from aqueous solutions. The elastic and piezoelastic coefficients were derived from ultrasonic resonance frequencies and their shifts upon variation of pressure, respectively, using the plate-resonance technique. Additionally, the coefficients of thermal expansion were determined between 95 K and 305 K by dilatometry. The elastic behaviour at ambient conditions is dominated by the 2-dimensional network of strong hydrogen bonds within the (001) plane leading to a corresponding pseudo-tetragonal anisotropy of the longitudinal elastic stiffness. The variation of elastic propertiesmore » with pressure, however, as well as the thermal expansion shows strong deviations from the pseudo-tetragonal symmetry. These deviations are probably correlated with tilts of the elongated tri-nuclear betaine–CuCl{sub 2}–water complexes. Neither the thermal expansion nor the specific heat capacity gives any hint on a phase transition in the investigated temperature range. - Graphical abstract: Single crystal of orthorhombic [(CH{sub 3}){sub 3}NCH{sub 2}COO]{sub 2}(CuCl{sub 2}){sub 3}·2H{sub 2}O. - Highlights: • Large single crystals (40 ×40 ×30 mm{sup 3}) of [(CH{sub 3}){sub 3}NCH{sub 2}COO]{sub 2}(CuCl{sub 2}){sub 3}·2H{sub 2}O were grown. • The elastic and piezoelastic coefficients were derived from ultrasonic resonance frequencies. • Thermal expansion (95 K–305 K) and heat capacity (113 K–323 K) were determined. • The orthorhombic structure shows pseudo-tetragonal elastic anisotropy at ambient conditions. • The crystal structure is stable in the investigated range (1–1600 bar, 95–303 K)« less

Data book for 12.5-inch diameter SRB thermal model water flotation test; 1.29 psia, series P022

NASA Technical Reports Server (NTRS)

Allums, S. L.

1974-01-01

Data acquired from tests conducted to determine how thermal conditions affect SRB (Space Shuttle Solid Rocket Booster) flotation at a scaled pressure of 1.29 psia are presented. Included are acceleration, pressure, and temperature data recorded from initial water impact to final flotation position using a 12.5-inch diameter thermal model of the SRB. Nineteen valid tests were conducted. These thermal tests indicated the following basic differences relative to the ambient temperature and pressure model tests: (1) more water was taken on board during penetration and (2) model flotation/sinking was temperature sensitive.

Pressure Ratio to Thermal Environments

NASA Technical Reports Server (NTRS)

Lopez, Pedro; Wang, Winston

2012-01-01

A pressure ratio to thermal environments (PRatTlE.pl) program is a Perl language code that estimates heating at requested body point locations by scaling the heating at a reference location times a pressure ratio factor. The pressure ratio factor is the ratio of the local pressure at the reference point and the requested point from CFD (computational fluid dynamics) solutions. This innovation provides pressure ratio-based thermal environments in an automated and traceable method. Previously, the pressure ratio methodology was implemented via a Microsoft Excel spreadsheet and macro scripts. PRatTlE is able to calculate heating environments for 150 body points in less than two minutes. PRatTlE is coded in Perl programming language, is command-line-driven, and has been successfully executed on both the HP and Linux platforms. It supports multiple concurrent runs. PRatTlE contains error trapping and input file format verification, which allows clear visibility into the input data structure and intermediate calculations.

Effects of pressure rise on cw laser ablation of tissue

NASA Astrophysics Data System (ADS)

LeCarpentier, Gerald L.; Motamedi, Massoud; Welch, Ashley J.

1991-06-01

The objectives of this research were to identify mechanisms responsible for the initiation of continuous wave (cw) laser ablation of tissue and investigate the role of pressure in the ablation process. Porcine aorta samples were irradiated in a chamber pressurized from 1 X 10-4 to 12 atmospheres absolute pressure. Acrylic and Zn-Se windows in the experimental pressure chamber allowed video and infrared cameras to simultaneously record mechanical and thermal events associated with cw argon laser ablation of these samples. Video and thermal images of tissue slabs documented the explosive nature of cw laser ablation of soft biological media and revealed similar ablation threshold temperatures and ablation onset times under different environmental pressures; however, more violent initiation explosions with decreasing environmental pressures were observed. These results suggest that ablation initiates with thermal alterations in the mechanical strength of the tissue and proceeds with an explosion induced by the presence superheated liquid within the tissue.

Nonlinear ultrasonic characterization of precipitation in 17-4PH stainless steel

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

Matlack, Kathryn; Bradley, Harrison A.; Thiele, Sebastian

2015-04-01

The extension of operational lifetime of most US nuclear reactors will cause reactor pressure vessel to be exposed to increased levels of neutron radiation damage. This research is part of a broader effort to develop a nondestructive evaluation technique to monitor radiation damage in reactor pressure vessel steels. The main contributor to radiation embrittlement in these steels is the formation of copper-rich precipitates. In this work, a precipitate hardenable martensitic alloy, 17-4PH stainless steel is exposed to thermal aging treatments, and used as a surrogate material to study the effects of copper precipitates on the measured acoustic nonlinearity parameter. Previousmore » work has demonstrated the effectiveness of these nonlinear ultrasonic (NLU) measurements in the characterization of radiation-induced microstructural changes in neutron irradiated reactor pressure vessel steels. NLU measurements using Rayleigh surface waves are performed on 17-4PH samples subjected to isothermal aging. NLU measurements are interpreted with hardness, thermo-electric power, TEM, and atom probe tomography measurements. The Rayleigh wave measurements showed a decrease in the acoustic nonlinearity parameter with increasing aging time, consistent with evidence of increasing number density of nucleated precipitates.« less

An adaptive load-following control system for a space nuclear power system

NASA Astrophysics Data System (ADS)

Metzger, John D.; El-Genk, Mohamed S.

An adaptive load-following control system is proposed for a space nuclear power system. The conceptual design of the SP-100 space nuclear power system proposes operating the nuclear reactor at a base thermal power and accommodating changes in the electrical power demand with a shunt regulator. It is necessary to increase the reactor thermal power if the payload electrical demand exceeds the peak system electrical output for the associated reactor power. When it is necessary to change the nuclear reactor power to meet a change in the power demand, the power ascension or descension must be accomplished in a predetermined manner to avoid thermal stresses in the system and to achieve the desired reactor period. The load-following control system described has the ability to adapt to changes in the system and to changes in the satellite environment. The application is proposed of the model reference adaptive control (MRAC). The adaptive control system has the ability to control the dynamic response of nonlinear systems. Three basic subsets of adaptive control are: (1) gain scheduling, (2) self-tuning regulators, and (3) model reference adaptive control.

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

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

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

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

75 FR 69711 - STP Nuclear Operating Company, South Texas Project Nuclear Power Plant, Units 3 and 4; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-15

... Rocks for Engineering Analysis and Design of Nuclear Power Plants.'' In this exemption request, the... the design of the wall. Hence, the staff concludes that, the resulting static and dynamic earth pressures will be bounded by the lateral earth pressures used in design. Bearing Capacity The applicant...

Thermal margin protection system for a nuclear reactor

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

Musick, C.R.

1974-02-12

A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

Dose distributions in phantoms irradiated in thermal columns of two different nuclear reactors.

PubMed

Gambarini, G; Agosteo, S; Altieri, S; Bortolussi, S; Carrara, M; Gay, S; Nava, E; Petrovich, C; Rosi, G; Valente, M

2007-01-01

In-phantom dosimetry studies have been carried out at the thermal columns of a thermal- and a fast-nuclear reactor for investigating: (a) the spatial distribution of the gamma dose and the thermal neutron fluence and (b) the accuracy at which the boron concentration should be estimated in an explanted organ of a boron neutron capture therapy patient. The phantom was a cylinder (11 cm in diameter and 12 cm in height) of tissue-equivalent gel. Dose images were acquired with gel dosemeters across the axial section of the phantom. The thermal neutron fluence rate was measured with activation foils in a few positions of this phantom. Dose and fluence rate profiles were also calculated with Monte Carlo simulations. The trend of these profiles do not show significant differences for the thermal columns considered in this work.

Radiation heat transfer in multitube, alkaline-metal thermal-to-electric converter

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

Tournier, J.M.P.; El-Genk, M.S.

Vapor anode, multitube Alkali-Metal Thermal-to-Electric Converters (AMTECs) are being considered for a number of space missions, such as the NASA Pluto/Express (PX) and Europa missions, scheduled for the years 2004 and 2005, respectively. These static converters can achieve a high fraction of Carnot efficiency at relatively low operating temperatures. An optimized cell can potentially provide a conversion efficiency between 20 and 30 percent, when operated at a hot-side temperature of 1000--1200 K and a cold-side temperature of 550--650 K. A comprehensive modeling and testing program of vapor anode, multitube AMTEC cells has been underway for more than three years atmore » the Air Force Research Laboratory`s Power and Thermal Group (AFRL/VSDVP), jointly with the University of New Mexico`s Institute for Space and Nuclear Power Studies. The objective of this program is to demonstrate the readiness of AMTECs for flight on future US Air Force space missions. A fast, integrated AMTEC Performance and Evaluation Analysis Model (APEAM) has been developed to support ongoing vacuum tests at AFRL and perform analyses and investigate potential design changes to improve the PX-cell performance. This model consists of three major components (Tournier and El-Genk 1998a, b): (a) a sodium vapor pressure loss model, which describes continuum, transition and free-molecule flow regimes in the low-pressure cavity of the cell; (b) an electrochemical and electrical circuit model; and (c) a radiation/conduction heat transfer model, for calculating parasitic heat losses. This Technical Note describes the methodology used to calculate the radiation view factors within the enclosure of the PX-cells, and the numerical procedure developed in this work to determine the radiation heat transport and temperatures within the cell cavity.« less

Evaluation of Used Fuel Disposition in Clay-Bearing Rock

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

Jové Colón, Carlos F.; Weck, Philippe F.; Sassani, David H.

2014-08-01

Radioactive waste disposal in shale/argillite rock formations has been widely considered given its desirable isolation properties (low permeability), geochemically reduced conditions, anomalous groundwater pressures, and widespread geologic occurrence. Clay/shale rock formations are characterized by their high content of clay minerals such as smectites and illites where diffusive transport and chemisorption phenomena predominate. These, in addition to low permeability, are key attributes of shale to impede radionuclide mobility. Shale host-media has been comprehensively studied in international nuclear waste repository programs as part of underground research laboratories (URLs) programs in Switzerland, France, Belgium, and Japan. These investigations, in some cases a decademore » or more long, have produced a large but fundamental body of information spanning from site characterization data (geological, hydrogeological, geochemical, geomechanical) to controlled experiments on the engineered barrier system (EBS) (barrier clay and seals materials). Evaluation of nuclear waste disposal in shale formations in the USA was conducted in the late 70’s and mid 80’s. Most of these studies evaluated the potential for shale to host a nuclear waste repository but not at the programmatic level of URLs in international repository programs. This report covers various R&D work and capabilities relevant to disposal of heat-generating nuclear waste in shale/argillite media. Integration and cross-fertilization of these capabilities will be utilized in the development and implementation of the shale/argillite reference case planned for FY15. Disposal R&D activities under the UFDC in the past few years have produced state-of-the-art modeling capabilities for coupled Thermal-Hydrological-Mechanical-Chemical (THMC), used fuel degradation (source term), and thermodynamic modeling and database development to evaluate generic disposal concepts. The THMC models have been developed for shale repository leveraging in large part on the information garnered in URLs and laboratory data to test and demonstrate model prediction capability and to accurately represent behavior of the EBS and the natural (barrier) system (NS). In addition, experimental work to improve our understanding of clay barrier interactions and TM couplings at high temperatures are key to evaluate thermal effects as a result of relatively high heat loads from waste and the extent of sacrificial zones in the EBS. To assess the latter, experiments and modeling approaches have provided important information on the stability and fate of barrier materials under high heat loads. This information is central to the assessment of thermal limits and the implementation of the reference case when constraining EBS properties and the repository layout (e.g., waste package and drift spacing). This report is comprised of various parts, each one describing various R&D activities applicable to shale/argillite media. For example, progress made on modeling and experimental approaches to analyze physical and chemical interactions affecting clay in the EBS, NS, and used nuclear fuel (source term) in support of R&D objectives. It also describes the development of a reference case for shale/argillite media. The accomplishments of these activities are summarized as follows: Development of a reference case for shale/argillite; Investigation of Reactive Transport and Coupled THM Processes in EBS: FY14; Update on Experimental Activities on Buffer/Backfill Interactions at elevated Pressure and Temperature; and Thermodynamic Database Development: Evaluation Strategy, Modeling Tools, First-Principles Modeling of Clay, and Sorption Database Assessment;ANL Mixed Potential Model For Used Fuel Degradation: Application to Argillite and Crystalline Rock Environments.« less

An FBG Optical Approach to Thermal Expansion Measurements under Hydrostatic Pressure

DOE PAGES

Rosa, Priscila Ferrari Silveira; Thomas, Sean Michael; Balakirev, Fedor Fedorovich; ...

2017-11-04

We report on an optical technique for measuring thermal expansion and magnetostriction at cryogenic temperatures and under applied hydrostatic pressures of 2.0 GPa. Optical fiber Bragg gratings inside a clamp-type pressure chamber are used to measure the strain in a millimeter-sized sample of CeRhIn 5. We describe the simultaneous measurement of two Bragg gratings in a single optical fiber using an optical sensing instrument capable of resolving changes in length [dL/L = (L- L 0)/L 0] on the order of 10 -7. Our results demonstrate the possibility of performing high-resolution thermal expansion measurements under hydrostatic pressure, a capability previously hinderedmore » by the small working volumes typical of pressure cells.« less

An FBG Optical Approach to Thermal Expansion Measurements under Hydrostatic Pressure

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

Rosa, Priscila Ferrari Silveira; Thomas, Sean Michael; Balakirev, Fedor Fedorovich

We report on an optical technique for measuring thermal expansion and magnetostriction at cryogenic temperatures and under applied hydrostatic pressures of 2.0 GPa. Optical fiber Bragg gratings inside a clamp-type pressure chamber are used to measure the strain in a millimeter-sized sample of CeRhIn 5. We describe the simultaneous measurement of two Bragg gratings in a single optical fiber using an optical sensing instrument capable of resolving changes in length [dL/L = (L- L 0)/L 0] on the order of 10 -7. Our results demonstrate the possibility of performing high-resolution thermal expansion measurements under hydrostatic pressure, a capability previously hinderedmore » by the small working volumes typical of pressure cells.« less

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

Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.

The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue frommore » J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.« less

Pressure and temperature induced elastic properties of rare earth chalcogenides

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

Shriya, S.; Sapkale, R., E-mail: sapkale.raju@rediffmail.com; Varshney, Dinesh, E-mail: vdinesh33@rediffmail.com

2016-05-06

The pressure and temperature dependent mechanical properties as Young modulus, Thermal expansion coefficient of rare earth REX (RE = La, Pr, Eu; X = O, S, Se, and Te) chalcogenides are studied. The rare earth chalcogenides showed a structural phase transition (B1–B2). Pressure dependence of Young modulus discerns an increase in pressure inferring the hardening or stiffening of the lattice as a consequence of bond compression and bond strengthening. Suppressed Young modulus as functions of temperature infers the weakening of the lattice results in bond weakening in REX. Thermal expansion coefficient demonstrates that REX (RE = La, Pr, Eu; Xmore » = O, S, Se, and Te) chalcogenides is mechanically stiffened, and thermally softened on applied pressure and temperature.« less

Thermal Insulation Performance of Textile Structures for Spacesuit Applications at Martian Pressure and Temperature

NASA Technical Reports Server (NTRS)

Orndoff, Evelyne; Trevino, Luis A.

2000-01-01

Protection of astronauts from the extreme temperatures in the space environment has been provided in the past using multi-layer insulation in ultra-high vacuum environments of low earth orbit and the lunar surface. For planetary environments with residual gas atmospheres such as Mars with ambient pressures between 8 to 14 hPa (8 to 14 mbar), new protection techniques are required because of the dominating effect of the ambient gas on heat loss through the insulation. At Mars ambient pressure levels, the heat loss can be excessive at expected suit external temperatures of 172 K with state-of-the-art suit insulation, requiring an active heat source and its accompanying weight and volume penalties. Micro-fibers have been identified as one potential structure to reduce the heat losses, but existing fundamental data on fiber heat transfer at low pressure is lacking for integrated fabric structures. This baseline study presents insulation performance test data at different pressures and fabric loads for selected polyesters and aramids as a function of fiber density, fiber diameter, fabric density, and fabric construction. A set of trend data of thermal conductivity versus ambient pressure is presented for each fiber and fabric construction design to identify the design effects on thermal conductivity at various ambient pressures, and to select a fiber and fabric design for further development as a suit insulation. The trend data also shows the pressure level at which thermal conductivity approaches a minimum, below which no further improvement is possible for a given fiber and fabric design. The pressure levels and resulting thermal conductivities from the trend data can then be compared to the ambient pressure at a planetary surface, Mars for example, to determine if a particular fiber and fabric design has potential as a suit insulation.

Effects of thermal loading and hydrostatic pressure on reflecting wavelengths of double-coated fiber Bragg grating with different coating-layer thickness

NASA Astrophysics Data System (ADS)

Seraji, Faramarz E.; Toutian, Golnoush

2017-10-01

Fiber Bragg grating (FBG) of different configurations used as sensing devices are vulnerable to environmental factors, such as static pressures and thermal loading, which cause their characteristic Bragg reflecting wavelengths to up/down-shift. In this paper, by considering double-coated FBG with different primary and secondary coating materials, the effects of thermal loading and hydrostatic pressure on FBG with different coating-layer thicknesses are analyzed to find design criteria for controlling the Bragg wavelength shift. The obtained results of the analysis may be employed as criteria to design pressure and temperature sensors when using double-coated FBGs.

Study of cavity effect in micro-Pirani gauge chamber with improved sensitivity for high vacuum regime

NASA Astrophysics Data System (ADS)

Zhang, Guohe; Lai, Junhua; Kong, Yanmei; Jiao, Binbin; Yun, Shichang; Ye, Yuxin

2018-05-01

Ultra-low pressure application of Pirani gauge needs significant improvement of sensitivity and expansion of measureable low pressure limit. However, the performance of Pirani gauge in high vacuum regime remains critical concerns since gaseous thermal conduction with high percentage is essential requirement. In this work, the heat transfer mechanism of micro-Pirani gauge packaged in a non-hermetic chamber was investigated and analyzed compared with the one before wafer-level packaging. The cavity effect, extremely important for the efficient detection of low pressure, was numerically and experimentally analyzed considering the influence of the pressure, the temperature and the effective heat transfer area in micro-Pirani gauge chamber. The thermal conduction model is validated by experiment data of MEMS Pirani gauges with and without capping. It is found that nature gaseous convection in chamber, determined by the Rayleigh number, should be taken into consideration. The experiment and model calculated results show that thermal resistance increases in the molecule regime, and further increases after capping due to the suppression of gaseous convection. The gaseous thermal conduction accounts for an increasing percentage of thermal conduction at low pressure while little changes at high pressure after capping because of the existence of cavity effect improving the sensitivity of cavity-effect-influenced Pirani gauge for high vacuum regime.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-10-28

India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai Business Line...MW thermal or special nuclear material connected therewith. U.S. Nuclear Cooperation with India: Issues for Congress Congressional Research

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

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

Kahler, Albert Comstock

We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

Thermal and high pressure inactivation kinetics of blueberry peroxidase.

PubMed

Terefe, Netsanet Shiferaw; Delon, Antoine; Versteeg, Cornelis

2017-10-01

This study for the first time investigated the stability and inactivation kinetics of blueberry peroxidase in model systems (McIlvaine buffer, pH=3.6, the typical pH of blueberry juice) during thermal (40-80°C) and combined high pressure-thermal processing (0.1-690MPa, 30-90°C). At 70-80°C, the thermal inactivation kinetics was best described by a biphasic model with ∼61% labile and ∼39% stable fractions at temperature between 70 and 75°C. High pressure inhibited the inactivation of the enzyme with no inactivation at pressures as high as 690MPa and temperatures less than 50°C. The inactivation kinetics of the enzyme at 60-70°C, and pressures higher than 500MPa was best described by a first order biphasic model with ∼25% labile fraction and 75% stable fraction. The activation energy values at atmospheric pressure were 548.6kJ/mol and 324.5kJ/mol respectively for the stable and the labile fractions. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

A review of nuclear thermal propulsion carbide fuel corrosion and key issues

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.

1994-01-01

Corrosion (mass loss) of carbide nuclear fuels due to their exposure to hot hydrogen in nuclear thermal propulsion engine systems greatly impacts the performance, thrust-to-weight and life of such systems. This report provides an overview of key issues and processes associated with the corrosion of carbide materials. Additionally, past pertinent development reactor test observations, as well as related experimental work and analysis modeling efforts are reviewed. At the conclusion, recommendations are presented, which provide the foundation for future corrosion modeling and verification efforts.

Analysis of a Nuclear Enhanced Airbreathing Rocket for Earth to Orbit Applications

NASA Technical Reports Server (NTRS)

Adams, Robert B.; Landrum, D. Brian; Brown, Norman (Technical Monitor)

2001-01-01

The proposed engine concept is the Nuclear Enhanced Airbreathing Rocket (NEAR). The NEAR concept uses a fission reactor to thermally heat a propellant in a rocket plenum. The rocket is shrouded, thus the exhaust mixes with ingested air to provide additional thermal energy through combustion. The combusted flow is then expanded through a nozzle to provide thrust.

"Bimodal" Nuclear Thermal Rocket (BNTR) Propulsion for Future Human Mars Exploration Missions

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

2004-01-01

The Nuclear Thermal Rocket (NTR) Propulsion program is discussed. The Rover/NERVA program from 1959-1972 is compared with the current program. A key technology description, bimodal vehicle design for Mars Cargo and the crew transfer vehicle with inflatable module and artificial gravity capability, including diagrams are included. The LOX-Augmented NTR concept/operational features and characteristics are discussed.

Double copper sheath multiconductor instrumentation cable is durable and easily installed in high thermal or nuclear radiation area

NASA Technical Reports Server (NTRS)

Mc Crae, A. W., Jr.

1967-01-01

Multiconductor instrumentation cable in which the conducting wires are routed through two concentric copper tube sheaths, employing a compressed insulator between the conductors and between the inner and outer sheaths, is durable and easily installed in high thermal or nuclear radiation area. The double sheath is a barrier against moisture, abrasion, and vibration.

The Pen Branch Project: Restoration of a Forested Wetland in South Carolina

Treesearch

Randall K. Kolka; Eric A. Nelson; Ronald E. Bonar; Neil C. Dulohery; David Gartner

1998-01-01

The Pen Branch Project is a program to restore a forested riparian wetland that has been subject to thermal disturbance caused by nuclear reactor operations at the Department of Energy's (DOE) Savannah River Site (SRS), an 80,200-hectare nuclear facility located in South Carolina. Various levels of thermal discharges to streams located across the US. have occurred...

DSMC analysis of species separation in rarefied nozzle flows

NASA Technical Reports Server (NTRS)

Chung, Chan-Hong; De Witt, Kenneth J.; Jeng, Duen-Ren; Penko, Paul F.

1992-01-01

The direct-simulation Monte Carlo method has been used to investigate the behavior of a small amount of a harmful species in the plume and the backflow region of nuclear thermal propulsion rockets. Species separation due to pressure diffusion and nonequilibrium effects due to rapid expansion into a surrounding low-density environment are the most important factors in this type of flow. It is shown that a relatively large amount of the lighter species is scattered into the backflow region and the heavier species becomes negligible in this region due to the extreme separation between species. It is also shown that the type of molecular interaction between the species can have a substantial effect on separation of the species.

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness ofmore » the PIUS concept to severe off-normal conditions having a very low probability of occurrence.« less

The Physics of Energy

NASA Astrophysics Data System (ADS)

Jaffe, Robert L.; Taylor, Washington

2018-01-01

Part I. Basic Energy Physics and Uses: 1. Introduction; 2. Mechanical energy; 3. Electromagnetic energy; 4. Waves and light; 5. Thermodynamics I: heat and thermal energy; 6. Heat transfer; 7. Introduction to quantum physics; 8. Thermodynamics II: entropy and temperature; 9. Energy in matter; 10. Thermal energy conversion; 11. Internal combustion engines; 12. Phase-change energy conversion; 13. Thermal power and heat extraction cycles; Part II. Energy Sources: 14. The forces of nature; 15. Quantum phenomena in energy systems; 16. An overview of nuclear power; 17. Structure, properties and decays of nuclei; 18. Nuclear energy processes: fission and fusion; 19. Nuclear fission reactors and nuclear fusion experiments; 20. Ionizing radiation; 21. Energy in the universe; 22. Solar energy: solar production and radiation; 23. Solar energy: solar radiation on Earth; 24. Solar thermal energy; 25. Photovoltaic solar cells; 26. Biological energy; 27. Ocean energy flow; 28. Wind: a highly variable resource; 29. Fluids – the basics; 30. Wind turbines; 31. Energy from moving water: hydro, wave, tidal, and marine current power; 32. Geothermal energy; 33. Fossil fuels; Part III. Energy System Issues and Externalities: 34. Energy and climate; 35. Earth's climate: past, present, and future; 36. Energy efficiency, conservation, and changing energy sources; 37. Energy storage; 38. Electricity generation and transmission.

Vibration isolation and pressure compensation apparatus for sensitive instrumentation

NASA Technical Reports Server (NTRS)

Averill, R. D. (Inventor)

1983-01-01

A system for attenuating the inherent vibration associated with a mechanical refrigeration unit employed to cryogenically cool sensitive instruments used in measuring chemical constituents of the atmosphere is described. A modular system including an instrument housing and a reaction bracket with a refrigerator unit floated there between comprise the instrumentation system. A pair of evacuated bellows that "float' refrigerator unit and provide pressure compensation at all levels of pressure from seal level to the vacuum of space. Vibration isolators and when needed provide additional vibration damping for the refrigerator unit. A flexible thermal strap (20 K) serves to provide essentially vibration free thermal contact between cold tip of the refrigerator unit and the instrument component mounted on the IDL mount. Another flexible strap (77 K) serves to provide vibration free thermal contact between the TDL mount thermal shroud and a thermal shroud disposed about the thermal shaft.

High pressure elasticity and thermal properties of depleted uranium

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

Jacobsen, M. K., E-mail: mjacobsen@lanl.gov; Velisavljevic, N., E-mail: nenad@lanl.gov

2016-04-28

Studies of the phase diagram of uranium have revealed a wealth of high pressure and temperature phases. Under ambient conditions the crystal structure is well defined up to 100 gigapascals (GPa), but very little information on thermal conduction or elasticity is available over this same range. This work has applied ultrasonic interferometry to determine the elasticity, mechanical, and thermal properties of depleted uranium to 4.5 GPa. Results show general strengthening with applied load, including an overall increase in acoustic thermal conductivity. Further implications are discussed within. This work presents the first high pressure studies of the elasticity and thermal properties ofmore » depleted uranium metal and the first real-world application of a previously developed containment system for making such measurements.« less

Numerical prediction of micro-channel LD heat sink operated with antifreeze based on CFD method

NASA Astrophysics Data System (ADS)

Liu, Gang; Liu, Yang; Wang, Chao; Wang, Wentao; Wang, Gang; Tang, Xiaojun

2014-12-01

To theoretically study the feasibility of antifreeze coolants applied as cooling fluids for high power LD heat sink, detailed Computational Fluid Dynamics (CFD) analysis of liquid cooled micro-channels heat sinks is presented. The performance operated with antifreeze coolant (ethylene glycol aqueous solution) compared with pure water are numerical calculated for the heat sinks with the same micro-channels structures. The maximum thermal resistance, total pressure loss (flow resistance), thermal resistance vs. flow-rate, and pressure loss vs. flow-rate etc. characteristics are numerical calculated. The results indicate that the type and temperature of coolants plays an important role on the performance of heat sinks. The whole thermal resistance and pressure loss of heat sinks increase significantly with antifreeze coolants compared with pure water mainly due to its relatively lower thermal conductivity and higher fluid viscosity. The thermal resistance and pressure loss are functions of the flow rate and operation temperature. Increasing of the coolant flow rate can reduce the thermal resistance of heat sinks; meanwhile increase the pressure loss significantly. The thermal resistance tends to a limit with increasing flow rate, while the pressure loss tends to increase exponentially with increasing flow rate. Low operation temperature chiefly increases the pressure loss rather than thermal resistance due to the remarkable increasing of fluid viscosity. The actual working point of the cooling circulation system can be determined on the basis of the pressure drop vs. flow rate curve for the micro-channel heat sink and that for the circulation system. In the same system, if the type or/and temperature of the coolant is changed, the working point is accordingly influenced, that is, working flow rate and pressure is changed simultaneously, due to which the heat sink performance is influenced. According to the numerical simulation results, if ethylene glycol aqueous solution is applied instead of pure water as the coolant under the same or a higher working temperature, the available output of optical power will decrease due to the worse heat sink performance; if applied under a lower working temperature(0 °C, －20 °C), although the heat sink performance become worse, however the temperature difference of heat transfer rises more significantly, the available output of optical power will increase on the contrary.

Combined pressure-thermal inactivation kinetics of Bacillus amyloliquefaciens spores in egg patty mince.

PubMed

Rajan, S; Ahn, J; Balasubramaniam, V M; Yousef, A E

2006-04-01

Bacillus amyloliquefaciens is a potential surrogate for Clostridium botulinum in validation studies involving bacterial spore inactivation by pressure-assisted thermal processing. Spores of B. amyloliquefaciens Fad 82 were inoculated into egg patty mince (approximately 1.4 x 10(8) spores per g), and the product was treated with combinations of pressure (0.1 to 700 MPa) and heat (95 to 121 degrees C) in a custom-made high-pressure kinetic tester. The values for the inactivation kinetic parameter (D), temperature coefficient (zT), and pressure coefficient (zP) were determined with a linear model. Inactivation parameters from the nonlinear Weibull model also were estimated. An increase in process pressure decreased the D-value at 95, 105, and 110 degrees C; however, at 121 degrees C the contribution of pressure to spore lethality was less pronounced. The zP-value increased from 170 MPa at 95 degrees C to 332 MPa at 121 degrees C, suggesting that B. amyloliquefaciens spores became less sensitive to pressure changes at higher temperatures. Similarly, the zT-value increased from 8.2 degrees C at 0.1 MPa to 26.8 degrees C at 700 MPa, indicating that at elevated pressures, the spores were less sensitive to changes in temperature. The nonlinear Weibull model parameter b increased with increasing pressure or temperature and was inversely related to the D-value. Pressure-assisted thermal processing is a potential alternative to thermal processing for producing shelf-stable egg products.

Fabrication of amorphous InGaZnO thin-film transistor-driven flexible thermal and pressure sensors

NASA Astrophysics Data System (ADS)

Park, Ick-Joon; Jeong, Chan-Yong; Cho, In-Tak; Lee, Jong-Ho; Cho, Eou-Sik; Kwon, Sang Jik; Kim, Bosul; Cheong, Woo-Seok; Song, Sang-Hun; Kwon, Hyuck-In

2012-10-01

In this work, we present the results concerning the use of amorphous indium-gallium-zinc-oxide (a-IGZO) thin-film transistor (TFT) as a driving transistor of the flexible thermal and pressure sensors which are applicable to artificial skin systems. Although the a-IGZO TFT has been attracting much attention as a driving transistor of the next-generation flat panel displays, no study has been performed about the application of this new device to the driving transistor of the flexible sensors yet. The proposed thermal sensor pixel is composed of the series-connected a-IGZO TFT and ZnO-based thermistor fabricated on a polished metal foil, and the ZnO-based thermistor is replaced by the pressure sensitive rubber in the pressure sensor pixel. In both sensor pixels, the a-IGZO TFT acts as the driving transistor and the temperature/pressure-dependent resistance of the ZnO-based thermistor/pressure-sensitive rubber mainly determines the magnitude of the output currents. The fabricated a-IGZO TFT-driven flexible thermal sensor shows around a seven times increase in the output current as the temperature increases from 20 °C to 100 °C, and the a-IGZO TFT-driven flexible pressure sensors also exhibit high sensitivity under various pressure environments.

n+235U resonance parameters and neutron multiplicities in the energy region below 100 eV

NASA Astrophysics Data System (ADS)

Pigni, Marco T.; Capote, Roberto; Trkov, Andrej; Pronyaev, Vladimir G.

2017-09-01

In August 2016, following the recent effort within the Collaborative International Evaluated Library Organization (CIELO) pilot project to improve the neutron cross sections of 235U, Oak Ridge National Laboratory (ORNL) collaborated with the International Atomic Energy Agency (IAEA) to release a resonance parameter evaluation. This evaluation restores the performance of the evaluated cross sections for the thermal- and above-thermal-solution benchmarks on the basis of newly evaluated thermal neutron constants (TNCs) and thermal prompt fission neutron spectra (PFNS). Performed with support from the US Nuclear Criticality Safety Program (NCSP) in an effort to provide the highest fidelity general purpose nuclear database for nuclear criticality applications, the resonance parameter evaluation was submitted as an ENDF-compatible file to be part of the next release of the ENDF/B-VIII.0 nuclear data library. The resonance parameter evaluation methodology used the Reich-Moore approximation of the R-matrix formalism implemented in the code SAMMY to fit the available time-of-flight (TOF) measured data for the thermal induced cross section of n+235U up to 100 eV. While maintaining reasonably good agreement with the experimental data, the validation analysis focused on restoring the benchmark performance for 235U solutions by combining changes to the resonance parameters and to the prompt resonance v̅

Thermal conductivity measurements in porous mixtures of methane hydrate and quartz sand

USGS Publications Warehouse

Waite, W.F.; deMartin, B.J.; Kirby, S.H.; Pinkston, J.; Ruppel, C.D.

2002-01-01

Using von Herzen and Maxwell's needle probe method, we measured thermal conductivity in four porous mixtures of quartz sand and methane gas hydrate, with hydrate composing 0, 33, 67 and 100% of the solid volume. Thermal conductivities were measured at a constant methane pore pressure of 24.8 MPa between -20 and +15??C, and at a constant temperature of -10??C between 3.5 and 27.6 MPa methane pore pressure. Thermal conductivity decreased with increasing temperature and increased with increasing methane pore pressure. Both dependencies weakened with increasing hydrate content. Despite the high thermal conductivity of quartz relative to methane hydrate, the largest thermal conductivity was measured in the mixture containing 33% hydrate rather than in hydrate-free sand. This suggests gas hydrate enhanced grain-to-grain heat transfer, perhaps due to intergranular contact growth during hydrate synthesis. These results for gas-filled porous mixtures can help constrain thermal conductivity estimates in porous, gas hydrate-bearing systems.

Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory

NASA Astrophysics Data System (ADS)

Bertani, C.; Falcone, N.; Bersano, A.; Caramello, M.; Matsushita, T.; De Salve, M.; Panella, B.

2017-11-01

High safety and reliability of advanced nuclear reactors, Generation IV and Small Modular Reactors (SMR), have a crucial role in the acceptance of these new plants design. Among all the possible safety systems, particular efforts are dedicated to the study of passive systems because they rely on simple physical principles like natural circulation, without the need of external energy source to operate. Taking inspiration from the second Decay Heat Removal system (DHR2) of ALFRED, the European Generation IV demonstrator of the fast lead cooled reactor, an experimental facility has been built at the Energy Department of Politecnico di Torino (PROPHET facility) to study single and two-phase flow natural circulation. The facility behavior is simulated using the thermal-hydraulic system code RELAP5-3D, which is widely used in nuclear applications. In this paper, the effect of the initial water inventory on natural circulation is analyzed. The experimental time behaviors of temperatures and pressures are analyzed. The experimental matrix ranges between 69 % and 93%; the influence of the opposite effects related to the increase of the volume available for the expansion and the pressure raise due to phase change is discussed. Simulations of the experimental tests are carried out by using a 1D model at constant heat power and fixed liquid and air mass; the code predictions are compared with experimental results. Two typical responses are observed: subcooled or two phase saturated circulation. The steady state pressure is a strong function of liquid and air mass inventory. The numerical results show that, at low initial liquid mass inventory, the natural circulation is not stable but pulsated.

Nuclear thermal propulsion workshop overview

NASA Technical Reports Server (NTRS)

Clark, John S.

1991-01-01

NASA is planning an Exploration Technology Program as part of the Space Exploration Initiative to return U.S. astronauts to the moon, conduct intensive robotic exploration of the moon and Mars, and to conduct a piloted mission to Mars by 2019. Nuclear Propulsion is one of the key technology thrust for the human mission to Mars. The workshop addresses NTP (Nuclear Thermal Rocket) technologies with purpose to: assess the state-of-the-art of nuclear propulsion concepts; assess the potential benefits of the concepts for the mission to Mars; identify critical, enabling technologies; lay-out (first order) technology development plans including facility requirements; and estimate the cost of developing these technologies to flight-ready status. The output from the workshop will serve as a data base for nuclear propulsion project planning.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-05-27

in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai Business Line Online...than 5 MW thermal or special nuclear material connected therewith. . U.S. Nuclear Cooperation with India: Issues for Congress Congressional

Electrical and thermal conductivity of Fe-C alloy at high pressure: implications for effects of carbon on the geodynamo of the Earth's core

NASA Astrophysics Data System (ADS)

Zhang, C.; Lin, J. F.; Liu, Y.; Feng, S.; Jin, C.; Yoshino, T.

2017-12-01

Thermal conductivity of iron alloy in the Earth's core plays a crucial role in constraining the energetics of the geodynamo and the thermal evolution of the planet. Studies on the thermal conductivity of iron reveal the importance of the effects of light elements and high temperature. Carbon has been proposed to be a candidate light element in Earth's core for its meteoritic abundance and high-pressure velocity-density profiles of iron carbides (e.g., Fe7C3). In this study, we employed four-probe van der Pauw method in a diamond anvil cell to measure the electrical resistivity of pure iron, iron carbon alloy, and iron carbides at high pressures. These studies were complimented with synchrotron X-ray diffraction and focused ion beam (FIB) analyses. Our results show significant changes in the electrical conductivity of these iron-carbon alloys that are consistent previous reports with structural and electronic transitions at high pressures, indicating that these transitions should be taken into account in evaluating the electrical and thermal conductivity at high pressure. To apply our results to understand the thermal conduction in the Earth's core, we have compared our results with literature values for the electrical and thermal conductivity of iron alloyed with light elements (C, Si) at high pressures. These comparisons permit the validity of the Wiedemann-Franz law and Matthiessen's rule for the effects of light elements on the thermal conductivity of the Earth's core. We found that an addition of a light element such as carbon has an strong effect on the reducing the thermal conductivity of Earth's core, but the magnitude of the alloying effect strongly depends on the identity of the light element and the crystal and electronic structures. Based on our results and literature values, we have modelled the electrical and thermal conductivity of iron-carbon alloy at Earth's core pressure-temperature conditions to the effects on the heat flux in the Earth's core. In this presentation, we will address how carbon as a potential light element in the Earth's core can significantly affect our view of the heat flux across the core-mantle boundary and geodynamo of our planet.

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.

(Ca,Mg)-Carbonate and Mg-Carbonate at the Phoenix Landing Site: Evaluation of the Phoenix Lander's Thermal Evolved Gas Analyzer (TEGA) Data Using Laboratory Simulations

NASA Technical Reports Server (NTRS)

Sutter, B.; Ming, D. W.; Boynton, W. V.; Niles, P. B.; Morris, R. V.

2011-01-01

Calcium carbonate (4.5 wt. %) was detected in the soil at the Phoenix Landing site by the Phoenix Lander s The Thermal and Evolved Gas Analyzer [1]. TEGA operated at 12 mbar pressure, yet the detection of calcium carbonate is based on interpretations derived from thermal analysis literature of carbonates measured under ambient (1000 mbar) and vacuum (10(exp -3) mbar) conditions [2,3] as well as at 100 and 30 mbar [4,5] and one analysis at 12 mbar by the TEGA engineering qualification model (TEGA-EQM). Thermodynamics (Te = H/ S) dictate that pressure affects entropy ( S) which causes the temperature (Te) of mineral decomposition at one pressure to differ from Te obtained at another pressure. Thermal decomposition analyses of Fe-, Mg-, and Ca-bearing carbonates at 12 mbar is required to enhance the understanding of the TEGA results at TEGA operating pressures. The objectives of this work are to (1) evaluate the thermal and evolved gas behavior of a suite of Fe-, Mg-, Ca-carbonate minerals at 1000 and 12 mbar and (2) discuss possible emplacement mechanisms for the Phoenix carbonate.

Pain Sensitivity Subgroups in Individuals With Spine Pain: Potential Relevance to Short-Term Clinical Outcome

PubMed Central

Bialosky, Joel E.; Robinson, Michael E.

2014-01-01

Background Cluster analysis can be used to identify individuals similar in profile based on response to multiple pain sensitivity measures. There are limited investigations into how empirically derived pain sensitivity subgroups influence clinical outcomes for individuals with spine pain. Objective The purposes of this study were: (1) to investigate empirically derived subgroups based on pressure and thermal pain sensitivity in individuals with spine pain and (2) to examine subgroup influence on 2-week clinical pain intensity and disability outcomes. Design A secondary analysis of data from 2 randomized trials was conducted. Methods Baseline and 2-week outcome data from 157 participants with low back pain (n=110) and neck pain (n=47) were examined. Participants completed demographic, psychological, and clinical information and were assessed using pain sensitivity protocols, including pressure (suprathreshold pressure pain) and thermal pain sensitivity (thermal heat threshold and tolerance, suprathreshold heat pain, temporal summation). A hierarchical agglomerative cluster analysis was used to create subgroups based on pain sensitivity responses. Differences in data for baseline variables, clinical pain intensity, and disability were examined. Results Three pain sensitivity cluster groups were derived: low pain sensitivity, high thermal static sensitivity, and high pressure and thermal dynamic sensitivity. There were differences in the proportion of individuals meeting a 30% change in pain intensity, where fewer individuals within the high pressure and thermal dynamic sensitivity group (adjusted odds ratio=0.3; 95% confidence interval=0.1, 0.8) achieved successful outcomes. Limitations Only 2-week outcomes are reported. Conclusions Distinct pain sensitivity cluster groups for individuals with spine pain were identified, with the high pressure and thermal dynamic sensitivity group showing worse clinical outcome for pain intensity. Future studies should aim to confirm these findings. PMID:24764070

Thermal investigation of nuclear waste disposal in space

NASA Technical Reports Server (NTRS)

Wilkinson, C. L.

1981-01-01

A thermal analysis has been conducted to determine the allowable size and response of bare and shielded nuclear waste forms in both low earth orbit and at 0.85 astronomical units. Contingency conditions of re-entry with a 45 deg and 60 deg aeroshell are examined as well as re-entry of a spherical shielded waste form. A variety of shielded schemes were examined and the waste form thermal response for each determined. Two optimum configurations were selected. The thermal response of these two shielded waste configurations to indefinite exposure to ground conditions following controlled and uncontrolled re-entry is determined. In all cases the prime criterion is that waste containment must be maintained.

American Nuclear Society 1994 student conference eastern region

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

NONE

This report contains abstracts from the 1994 American Nuclear Society Student Conference. The areas covered by these abstracts are: fusion and plasma physics; nuclear chemistry; radiation detection; reactor physics; thermal hydraulics; and corrosion science and waste issues.

Mission design considerations for nuclear risk mitigation

NASA Technical Reports Server (NTRS)

Stancati, Mike; Collins, John

1993-01-01

Strategies for the mitigation of the nuclear risk associated with two specific mission operations are discussed. These operations are the safe return of nuclear thermal propulsion reactors to earth orbit and the disposal of lunar/Mars spacecraft reactors.

Lightweight Damage Tolerant Radiators for In-Space Nuclear Electric Power and Propulsion

NASA Technical Reports Server (NTRS)

Craven, Paul; SanSoucie, Michael P.; Tomboulian, Briana; Rogers, Jan; Hyers, Robert

2014-01-01

Nuclear electric propulsion (NEP) is a promising option for high-speed in-space travel due to the high energy density of nuclear power sources and efficient electric thrusters. Advanced power conversion technologies for converting thermal energy from the reactor to electrical energy at high operating temperatures would benefit from lightweight, high temperature radiator materials. Radiator performance dictates power output for nuclear electric propulsion systems. Pitch-based carbon fiber materials have the potential to offer significant improvements in operating temperature and mass. An effort at the NASA Marshall Space Flight Center to show that woven high thermal conductivity carbon fiber mats can be used to replace standard metal and composite radiator fins to dissipate waste heat from NEP systems is ongoing. The goals of this effort are to demonstrate a proof of concept, to show that a significant improvement of specific power (power/mass) can be achieved, and to develop a thermal model with predictive capabilities. A description of this effort is presented.

Carbide fuels for nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

Matthews, R. B.; Blair, H. T.; Chidester, K. M.; Davidson, K. V.; Stark, W. E.; Storms, E. K.

1991-09-01

A renewed interest in manned exploration of space has revitalized interest in the potential for advancing nuclear rocket technology developed during the 1960's. Carbide fuel performance, melting point, stability, fabricability and compatibility are key technology issues for advanced Nuclear Thermal Propulsion reactors. The Rover fuels development ended with proven carbide fuel forms with demonstrated operating temperatures up to 2700 K for over 100 minutes. The next generation of nuclear rockets will start where the Rover technology ended, but with a more rigorous set of operating requirements including operating lifetime to 10 hours, operating temperatures greater that 3000 K, low fission product release, and compatibility. A brief overview of Rover/NERVA carbide fuel development is presented. A new fuel form with the highest potential combination of operating temperature and lifetime is proposed that consists of a coated uranium carbide fuel sphere with built-in porosity to contain fission products. The particles are dispersed in a fiber reinforced ZrC matrix to increase thermal shock resistance.

Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

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

Weber, William J.; Zhang, Yanwen

This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effectsmore » of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.« less

Magnetic cooling for microkelvin nanoelectronics on a cryofree platform.

PubMed

Palma, M; Maradan, D; Casparis, L; Liu, T-M; Froning, F N M; Zumbühl, D M

2017-04-01

We present a parallel network of 16 demagnetization refrigerators mounted on a cryofree dilution refrigerator aimed to cool nanoelectronic devices to sub-millikelvin temperatures. To measure the refrigerator temperature, the thermal motion of electrons in a Ag wire-thermalized by a spot-weld to one of the Cu nuclear refrigerators-is inductively picked-up by a superconducting gradiometer and amplified by a SQUID mounted at 4 K. The noise thermometer as well as other thermometers are used to characterize the performance of the system, finding magnetic field independent heat-leaks of a few nW/mol, cold times of several days below 1 mK, and a lowest temperature of 150 μK of one of the nuclear stages in a final field of 80 mT, close to the intrinsic SQUID noise of about 100 μK. A simple thermal model of the system capturing the nuclear refrigerator, heat leaks, and thermal and Korringa links describes the main features very well, including rather high refrigerator efficiencies typically above 80%.

Thermoelectric power measurement under hydrostatic pressure using a self-clamped pressure cell

NASA Astrophysics Data System (ADS)

Choi, E. S.; Kang, Haeyong; Jo, Y. J.; Kang, W.

2002-08-01

A thermoelectric power (TEP) measurement technique in a self-clamped pressure cell is presented. Thermal and electrical contacts were glued to heaters by Stycast epoxy, which enhances thermal integration. The pressure effect of Chromel-Constantan and Chromel-AuFe0.07% thermocouples are compared to Chromel-Alumel thermocouples, which are known to be pressure insensitive between 4.2 and 300 K. The investigated thermocouples are found to have a small pressure effect; approx][plus-or-minus4% at maximum in the measured temperature and pressure range. Any pressure effect on Au wires was also found to be very small from the pressure-dependent TEP measurement of YBCO superconductor below Tc.

Historical Material Analysis of DC745U Pressure Pads

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

Ortiz-Acosta, Denisse

As part of the Enhance Surveillance mission, it is the goal to provide suitable lifetime assessment of stockpile materials. This report is an accumulation of historical publication on the DC745U material and their findings. It is the intention that the B61 LEP program uses this collection of data to further develop their understanding and potential areas of study. DC745U is a commercially available silicone elastomer consisting of dimethyl, methyl-phenyl, and methyl-vinyl siloxane repeat units. Originally, this material was manufactured by Dow Corning as Silastic{reg_sign} DC745U at their manufacturing facility in Kendallville, IN. Recently, Dow Corning shifted this material to themore » Xiameter{reg_sign} brand product line. Currently, DC745U is available through Xiameter{reg_sign} or Dow Corning's distributor R. D. Abbott Company. DC745U is cured using 0.5 wt% vinyl-specific peroxide curing agent known as Luperox 101 or Varox DBPH-50. This silicone elastomer is used in numerous parts, including two major components (outer pressure pads and aft cap support) in the W80 and as pressure pads on the B61. DC745U is a proprietary formulation, thus Dow Corning provides limited information on its composition and properties. Based on past experience with Dow Corning, DC745U is at risk of formulation changes without notification to the costumer. A formulation change for DC745U may have a significant impact because the network structure is a key variable in determining material properties. The purpose of this report is to provide an overview of historical DC745U studies and identify gaps that need to be addressed in future work. Some of the previous studies include the following: 1. Spectroscopic characterization of raw gum stock. 2. Spectroscopic, thermal, and mechanical studies on cured DC745U. 3. Nuclear Magnetic Resonance (NMR) and solvent swelling studies on DC745U with different crosslink densities. 4. NMR, solvent swelling, thermal, and mechanical studies on thermally aged DC745U. 5. NMR, solvent swelling, thermal, and mechanical studies on radiolytically aged DC745U. Each area is reviewed and further work is suggested to improve our understanding of DC745U for systems engineering, surveillance, aging assessments, and lifetime assessment.« less

RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

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

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumptionmore » of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

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

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. Themore » assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

NASA Astrophysics Data System (ADS)

Bersano, Andrea

With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be compared to the values predicted by the system code and differences will be discussed with the ultimate goal to qualify RELAP5-3D for the analysis of decay heat removal systems in natural circulation. The numerical data will be also used to understand the key parameters related to the heat transfer in natural circulation and to optimize the operation of the system.

REVIEW OF PROPOSED METHODOLOGY FOR A RISK- INFORMED RELAXATION TO ASME SECTION XI APPENDIX G

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

Dickson, Terry L; Kirk, Mark

2010-01-01

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclearmore » industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.« less

The Influence of Temperature on Time-Dependent Deformation and Failure in Granite: A Mesoscale Modeling Approach

NASA Astrophysics Data System (ADS)

Xu, T.; Zhou, G. L.; Heap, Michael J.; Zhu, W. C.; Chen, C. F.; Baud, Patrick

2017-09-01

An understanding of the influence of temperature on brittle creep in granite is important for the management and optimization of granitic nuclear waste repositories and geothermal resources. We propose here a two-dimensional, thermo-mechanical numerical model that describes the time-dependent brittle deformation (brittle creep) of low-porosity granite under different constant temperatures and confining pressures. The mesoscale model accounts for material heterogeneity through a stochastic local failure stress field, and local material degradation using an exponential material softening law. Importantly, the model introduces the concept of a mesoscopic renormalization to capture the co-operative interaction between microcracks in the transition from distributed to localized damage. The mesoscale physico-mechanical parameters for the model were first determined using a trial-and-error method (until the modeled output accurately captured mechanical data from constant strain rate experiments on low-porosity granite at three different confining pressures). The thermo-physical parameters required for the model, such as specific heat capacity, coefficient of linear thermal expansion, and thermal conductivity, were then determined from brittle creep experiments performed on the same low-porosity granite at temperatures of 23, 50, and 90 °C. The good agreement between the modeled output and the experimental data, using a unique set of thermo-physico-mechanical parameters, lends confidence to our numerical approach. Using these parameters, we then explore the influence of temperature, differential stress, confining pressure, and sample homogeneity on brittle creep in low-porosity granite. Our simulations show that increases in temperature and differential stress increase the creep strain rate and therefore reduce time-to-failure, while increases in confining pressure and sample homogeneity decrease creep strain rate and increase time-to-failure. We anticipate that the modeling presented herein will assist in the management and optimization of geotechnical engineering projects within granite.

Thermal Propulsion Capture System Heat Exchanger Design

NASA Technical Reports Server (NTRS)

Richard, Evan M.

2016-01-01

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

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-04-08

Kazakhstan might start uranium exports to India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor...reactors producing more than 5 MW thermal or special nuclear material connected therewith. U.S. Nuclear Cooperation with India: Issues for Congress

Nuclear Hybrid Energy Systems FY16 Modeling Efforts at ORNL

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

Cetiner, Sacit M.; Greenwood, Michael Scott; Harrison, Thomas J.

A nuclear hybrid system uses a nuclear reactor as the basic power generation unit. The power generated by the nuclear reactor is utilized by one or more power customers as either thermal power, electrical power, or both. In general, a nuclear hybrid system will couple the nuclear reactor to at least one thermal power user in addition to the power conversion system. The definition and architecture of a particular nuclear hybrid system is flexible depending on local markets needs and opportunities. For example, locations in need of potable water may be best served by coupling a desalination plant to themore » nuclear system. Similarly, an area near oil refineries may have a need for emission-free hydrogen production. A nuclear hybrid system expands the nuclear power plant from its more familiar central power station role by diversifying its immediately and directly connected customer base. The definition, design, analysis, and optimization work currently performed with respect to the nuclear hybrid systems represents the work of three national laboratories. Idaho National Laboratory (INL) is the lead lab working with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory. Each laboratory is providing modeling and simulation expertise for the integration of the hybrid system.« less

High-Pressure Transport Properties Of Fluids: Theory And Data From Levitated Drops At Combustion-Relevant Temperatures

NASA Technical Reports Server (NTRS)

Bellan, Josette; Harstad, Kenneth; Ohsaka, Kenichi

2003-01-01

Although the high pressure multicomponent fluid conservation equations have already been derived and approximately validated for binary mixtures by this PI, the validation of the multicomponent theory is hampered by the lack of existing mixing rules for property calculations. Classical gas dynamics theory can provide property mixing-rules at low pressures exclusively. While thermal conductivity and viscosity high-pressure mixing rules have been documented in the literature, there is no such equivalent for the diffusion coefficients and the thermal diffusion factors. The primary goal of this investigation is to extend the low pressure mixing rule theory to high pressures and validate the new theory with experimental data from levitated single drops. The two properties that will be addressed are the diffusion coefficients and the thermal diffusion factors. To validate/determine the property calculations, ground-based experiments from levitated drops are being conducted.

Nuclear Thermal Propulsion Mars Mission Systems Analysis and Requirements Definition

NASA Technical Reports Server (NTRS)

Mulqueen, Jack; Chiroux, Robert C.; Thomas, Dan; Crane, Tracie

2007-01-01

This paper describes the Mars transportation vehicle design concepts developed by the Marshall Space Flight Center (MSFC) Advanced Concepts Office. These vehicle design concepts provide an indication of the most demanding and least demanding potential requirements for nuclear thermal propulsion systems for human Mars exploration missions from years 2025 to 2035. Vehicle concept options vary from large "all-up" vehicle configurations that would transport all of the elements for a Mars mission on one vehicle. to "split" mission vehicle configurations that would consist of separate smaller vehicles that would transport cargo elements and human crew elements to Mars separately. Parametric trades and sensitivity studies show NTP stage and engine design options that provide the best balanced set of metrics based on safety, reliability, performance, cost and mission objectives. Trade studies include the sensitivity of vehicle performance to nuclear engine characteristics such as thrust, specific impulse and nuclear reactor type. Tbe associated system requirements are aligned with the NASA Exploration Systems Mission Directorate (ESMD) Reference Mars mission as described in the Explorations Systems Architecture Study (ESAS) report. The focused trade studies include a detailed analysis of nuclear engine radiation shield requirements for human missions and analysis of nuclear thermal engine design options for the ESAS reference mission.

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

Thermal power transfer system using applied potential difference to sustain operating pressure difference

NASA Technical Reports Server (NTRS)

Bhandari, Pradeep (Inventor); Fujita, Toshio (Inventor)

1991-01-01

A thermal power transfer system using a phase change liquid gas fluid in a closed loop configuration has a heat exchanger member connected to a gas conduit for inputting thermal energy into the fluid. The pressure in the gas conduit is higher than a liquid conduit that is connected to a heat exchanger member for outputting thermal energy. A solid electrolyte member acts as a barrier between the gas conduit and the liquid conduit adjacent to a solid electrolyte member. The solid electrolyte member has the capacity of transmitting ions of a fluid through the electrolyte member. The ions can be recombined with electrons with the assistance of a porous electrode. An electrical field is applied across the solid electrolyte member to force the ions of the fluid from a lower pressure liquid conduit to the higher pressure gas conduit.

Thermal conductivity of H2O-CH3OH mixtures at high pressures: Implications for the dynamics of icy super-Earths outer shells

NASA Astrophysics Data System (ADS)

Hsieh, Wen-Pin; Deschamps, Frédéric

2015-10-01

Thermal conductivity of H2O-volatile mixtures at extreme pressure-temperature conditions is a key factor to determine the heat flux and profile of the interior temperature in icy bodies. We use time domain thermoreflectance and stimulated Brillouin scattering combined with diamond anvil cells to study the thermal conductivity and sound velocity of water (H2O)-methanol (CH3OH) mixtures to pressures as high as 12 GPa. Compared to pure H2O, the presence of 5-20 wt % CH3OH significantly reduces the thermal conductivity and sound velocity when the mixture becomes ice VI-CH3OH and ice VII-CH3OH phases at high pressures, indicating that the heat transfer is hindered within the icy body. We then apply these results to model the heat transfer through the icy mantles of super-Earths, assuming that these mantles are animated by thermal convection. Our calculations indicate that the decrease of thermal conductivity due to the presence of 10 wt % CH3OH induces a twofold decrease of the power transported by convection.

Nuclear Power Plant environment`s surveillance by satellite remote sensing and in-situ monitoring data

NASA Astrophysics Data System (ADS)

Zoran, Maria

The main environmental issues affecting the broad acceptability of nuclear power plant are the emission of radioactive materials, the generation of radioactive waste, and the potential for nuclear accidents. All nuclear fission reactors, regardless of design, location, operator or regulator, have the potential to undergo catastrophic accidents involving loss of control of the reactor core, failure of safety systems and subsequent widespread fallout of hazardous fission products. Risk is the mathematical product of probability and consequences, so lowprobability and high-consequence accidents, by definition, have a high risk. NPP environment surveillance is a very important task in frame of risk assessment. Satellite remote sensing data had been applied for dosimeter levels first time for Chernobyl NPP accident in 1986. Just for a normal functioning of a nuclear power plant, multitemporal and multispectral satellite data in complementarily with field data are very useful tools for NPP environment surveillance and risk assessment. Satellite remote sensing is used as an important technology to help environmental research to support research analysis of spatio-temporal dynamics of environmental features nearby nuclear facilities. Digital processing techniques applied to several LANDSAT, MODIS and QuickBird data in synergy with in-situ data are used to assess the extent and magnitude of radiation and non-radiation effects on the water, near field soil, vegetation and air. As a test case the methodology was applied for for Nuclear Power Plant (NPP) Cernavoda, Romania. Thermal discharge from nuclear reactors cooling is dissipated as waste heat in Danube-Black -Sea Canal and Danube River. Water temperatures captured in thermal IR imagery are correlated with meteorological parameters. If during the winter thermal plume is localized to an area of a few km of NPP, the temperature difference between the plume and non-plume areas being about 1.5 oC, during summer and fall , is a larger thermal plume up to 5-6 km far along Danube Black Sea Canal ,the temperature change is about 1.0 oC.

Self-contained high-pressure cell, apparatus, and procedure for the preparation of encapsulated proteins dissolved in low viscosity fluids for nuclear magnetic resonance spectroscopy

NASA Astrophysics Data System (ADS)

Peterson, Ronald W.; Wand, A. Joshua

2005-09-01

The design of a sample cell for high-performance nuclear magnetic resonance (NMR) at elevated pressure is described. The cell has been optimized for the study of encapsulated proteins dissolved in low viscosity fluids but is suitable for more general nuclear magnetic resonance (NMR) spectroscopy of biomolecules at elevated pressure. The NMR cell is comprised of an alumina-toughened zirconia tube mounted on a self-sealing nonmagnetic metallic valve. The cell has several advantages, including relatively low cost, excellent NMR performance, high-pressure tolerance, chemical inertness, and a relatively large active volume. Also described is a low volume sample preparation device that allows for the preparation of samples under high hydrostatic pressure and their subsequent transfer to the NMR cell.

Study on hot melt pressure sensitive coil material for removing surface nuclear pollution dust

NASA Astrophysics Data System (ADS)

Wang, Jing; Li, Jiao; Wang, Jianhui; Zheng, Li; Li, Jian; Lv, Linmei

2018-02-01

A new method for removing surface nuclear pollution by using hot melt pressure sensitive membrane was presented. The hot melt pressure sensitive membrane was designed and prepared by screening hot melt pressure sensitive adhesive and substrate. The simulated decontamination test of the hot melt pressure sensitive membrane was performed by using 100 mesh and 20 mesh standard sieve dust for simulation of nuclear explosion fall ash and radioactive contaminated particles, respectively. It was found that the single decontamination rate of simulated fall ash and contaminated particles were both above 80% under pressure conditions of 25kPa or more at 140°C. And the maximum single decontamination rate was 92.5%. The influence of heating temperature and pressure on the decontamination rate of the membrane was investigated at the same time. The results showed that higher heating temperature could increase the decontamination rate by increasing the viscosity of the adhesive. When the adhesive amount of the adhesive layer reached saturation, a higher pressure could increase the single decontamination rate also.

Improving Prompt Temperature Feedback by Stimulating Doppler Broadening in Heterogeneous Composite Nuclear Fuel Forms

NASA Astrophysics Data System (ADS)

Morrison, Christopher

Nuclear fuels with similar aggregate material composition, but with different millimeter and micrometer spatial configurations of the component materials can have very different safety and performance characteristics. This research focuses on modeling and attempting to engineer heterogeneous combinations of nuclear fuels to improve negative prompt temperature feedback in response to reactivity insertion accidents. Improvements in negative prompt temperature feedback are proposed by developing a tailored thermal resistance in the nuclear fuel. In the event of a large reactivity insertion, the thermal resistance allows for a faster negative Doppler feedback by temporarily trapping heat in material zones with strong absorption resonances. A multi-physics simulation framework was created that could model large reactivity insertions. The framework was then used to model a comparison of a heterogeneous fuel with a tailored thermal resistance and a homogeneous fuel without the tailored thermal resistance. The results from the analysis confirmed the fundamental premise of prompt temperature feedback and provide insights into the neutron spectrum dynamics throughout the transient process. A trade study was conducted on infinite lattice fuels to help map a design space to study and improve prompt temperature feedback with many results. A multi-scale fuel pin analysis was also completed to study more realistic geometries. The results of this research could someday allow for novel nuclear fuels that would behave differently than current fuels. The idea of having a thermal barrier coating in the fuel is contrary to most current thinking. Inherent resistance to reactivity insertion accidents could enable certain reactor types once considered vulnerable to reactivity insertion accidents to be reevaluated in light of improved negative prompt temperature feedback.

Heat Transfer Enhancement of Metal Hydride Particle Bed for Heat Driven Type Refrigerator by Carbon Fiber

NASA Astrophysics Data System (ADS)

Bae, Sang-Chul; Tanae, Takayuki; Monde, Masanori; Katsuta, Masafumi

A series of study has been performed on the metal hydride particle beds of Ti0.15Zr0.85Cr0.9Fe0.6Ni0.2Mn0.3Cu0.05 (MH-1, using for heat source), Ti0.73Zr0.27Cr1.2Fe0.3Ni0.1Mn0.4Cu0.05 (MH-2, using for cooling load) to measure the effective thermal conductivities. The effective thermal conductivities of activated and oxidized MH particle bed in helium have been examined. Experiment results show that pressure has great influence on effective thermal conductivity in low pressure range (<0.5 MPa). And that influence decreases rapidly with increase of gas pressure. The reason of pressure dependence at low pressure range is that the mean free path of gas becomes greater than effective thickness of gas film which is important to the heat transfer mechanism of particle bed. In order to enhance the poor thermal conductivity of metal hydride particle bed, carbon fiber mixing method has been used in this study. Three types, two insert methods and five mass percentages of carbon fiber have been examined and compared. The highest effective thermal conductivity of MH particle bed has been reached with Type B carbon fiber which has second higher thermal conductivity, and 2 weight percentage. This method has acquired 5-6 times higher thermal conductivity than pure metal hydride particle beds with quite low quantity of additives, only 2 mass% of carbon fiber. This is a good result comparing to other method which can reach higher effective thermal conductivity but needs much higher percentage of additives too.

Thermal-hydraulic analysis capabilities and methods development at NYPA

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

Feltus, M.A.

1987-01-01

The operation of a nuclear power plant must be regularly supported by various thermal-hydraulic (T/H) analyses that may include final safety analysis report (FSAR) design basis calculations and licensing evaluations and conservative and best-estimate analyses. The development of in-house T/H capabilities provides the following advantages: (a) it leads to a better understanding of the plant design basis and operating characteristics; (b) methods developed can be used to optimize plant operations and enhance plant safety; (c) such a capability can be used for design reviews, checking vendor calculations, and evaluating proposed plant modifications; and (d) in-house capability reduces the cost ofmore » analysis. This paper gives an overview of the T/H capabilities and current methods development activity within the engineering department of the New York Power Authority (NYPA) and will focus specifically on reactor coolant system (RCS) transients and plant dynamic response for non-loss-of-coolant accident events. This paper describes NYPA experience in performing T/H analyses in support of pressurized water reactor plant operation.« less

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

Ranjan, Devesh

Diffusion bonded heat exchangers are the leading candidates for the sCO 2 Brayton cycles in next generation nuclear power plants. Commercially available diffusion bonded heat exchangers utilize set of continuous semi-circular zigzag micro channels to increase the heat transfer area and enhance heat transfer through increased turbulence production. Such heat exchangers can lead to excessive pressure drop as well as flow maldistribution in the case of poorly designed flow distribution headers. The goal of the current project is to fabricate and test potential discontinuous fin patterns for diffusion bonded heat exchangers; which can achieve desired thermal performance at lower pressuremore » drops. Prototypic discontinuous offset rectangular and Airfoil fin surface geometries were chemically etched on to 316 stainless steel plate and sealed against an un-etched flat pate using O-ring seal emulating diffusion bonded heat exchangers. Thermal-hydraulic performance of these prototypic discontinuous fin geometries was experimentally evaluated and compared to the existing data for the continuous zigzag channels. The data generated from this project will serve as the database for future testing and validation of numerical models.« less

Battery management systems with thermally integrated fire suppression

DOEpatents

Bandhauer, Todd M.; Farmer, Joseph C.

2017-07-11

A thermal management system is integral to a battery pack and/or individual cells. It relies on passive liquid-vapor phase change heat removal to provide enhanced thermal protection via rapid expulsion of inert high pressure refrigerant during abnormal abuse events and can be integrated with a cooling system that operates during normal operation. When a thermal runaway event occurs and sensed by either active or passive sensors, the high pressure refrigerant is preferentially ejected through strategically placed passages within the pack to rapidly quench the battery.

Development, fabrication and evaluation of composite thermal engine insulation

NASA Technical Reports Server (NTRS)

1973-01-01

Foil enclosure configurations of 10 variations were fabricated and evaluated. A discussion of the thermal protection system panel design includes: (1) description of 3DSX/foil concept, (2) design environment, (3) material selection, (4) fabrication enclosure, (5) structural design, (6) thermal sizing, and (7) weight analysis. The structural design study includes foil evaluation, venting pressure loads, thermomechanical behavior, and enclosure venting (burst) pressure tests. Results of experimental demonstrations of performance and reuse capabilities are given for both thermal and acoustic testing.

The NASA Advanced Exploration Systems Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Mitchell, Doyce P.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Gerrish, Harold P.; Doughty, Glen; Belvin, Anthony; Clement, Steven; Borowski, Stanley K.;

An historical collection of papers on nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

The present volume of historical papers on nuclear thermal propulsion (NTP) encompasses NTP technology development regarding solid-core NTP technology, advanced concepts from the early years of NTP research, and recent activities in the field. Specific issues addressed include NERVA rocket-engine technology, the development of nuclear rocket propulsion at Los Alamos, fuel-element development, reactor testing for the Rover program, and an overview of NTP concepts and research emphasizing two decades of NASA research. Also addressed are the development of the 'nuclear light bulb' closed-cycle gas core and a demonstration of a fissioning UF6 gas in an argon vortex. The recent developments reviewed include the application of NTP to NASA's Lunar Space Transportation System, the use of NTP for the Space Exploration Initiative, and the development of nuclear rocket engines in the former Soviet Union.

The use of combined thermal/pressure polyvinylidene fluoride film airflow sensor in polysomnography.

PubMed

Kryger, Meir; Eiken, Todd; Qin, Li

2013-12-01

The technologies recommended by the American Academy of Sleep Medicine (AASM) to monitor airflow in polysomnography (PSG) include the simultaneous monitoring of two physical variables: air temperature (for thermal airflow) and air pressure (for nasal pressure). To comply with airflow monitoring standards in the sleep lab setting thus often requires the patient to wear two sensors under the nose during testing. We hypothesized that a single combined thermal/pressure sensor using polyvinylidene fluoride (PVDF) film responsive to both airflow temperature and pressure would be effective in documenting abnormal breathing events during sleep. Sixty patients undergoing routine PSG testing to rule out obstructive sleep apnea at two different sleep laboratories were asked to wear a third PVDF airflow sensor in addition to the traditional thermal sensor and pressure sensor. Apnea and hypopnea events were scored by the sleep lab technologists using the AASM guidelines (CMS option) using the thermal sensor for apnea and the pressure sensor for hypopnea (scorer 1). The digital PSG data were also forwarded to an outside registered polysomnographic technologist for scoring of respiratory events detected in the PVDF airflow channels (scorer 2). The Pearson correlation coefficient, r, between apnea and hypopnea indices obtained using the AASM sensors and the combined PVDF sensor was almost unity for the four calculated indices: apnea-hypopnea index (0.990), obstructive apnea index (0.992), hypopnea index (0.958), and central apnea index (1.0). The slope of the four relationships was virtually unity and the coefficient of determination (r (2)) was also close to 1. The results of intraclass correlation coefficients (>0.95) and Bland-Altman plots also provide excellent agreement between the combined PVDF sensor and the AASM sensors. The indices used to calculate apnea severity obtained with the combined PVDF thermal and pressure sensor were equivalent to those obtained using AASM-recommended sensors.

Large blast and thermal simulator advanced concept driver design by computational fluid dynamics. Final report, 1987-1989

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

Opalka, K.O.

1989-08-01

The construction of a large test facility has been proposed for simulating the blast and thermal environment resulting from nuclear explosions. This facility would be used to test the survivability and vulnerability of military equipment such as trucks, tanks, and helicopters in a simulated thermal and blast environment, and to perform research into nuclear blast phenomenology. The proposed advanced design concepts, heating of driver gas and fast-acting throat valves for wave shaping, are described and the results of CFD studies to advance these new technical concepts fro simulating decaying blast waves are reported.

Design and evaluation of candidate pressure ports for the HYFLITE experiment

NASA Technical Reports Server (NTRS)

Teter, John E., Jr.; Cleckner, Craig S.; Vontheumer, Alfred E.

1994-01-01

A concept for placing a pressure transducer directly in a shuttle type tile was developed at Langley Research Center. A 5 inch long quartz with a .020 inch inner diameter provides the thermal isolation necessary to allow 2800 F surface pressure measurements to be taken by pressure transducer rated at 250 F. The assembly is potted in place with RTV 560 in a piece of FRCI-12 thermal protection system insulation tile. The integrity of the thermal protection system is maintained even with the intrusion of the pressure port assembly and the pressure port does not disrupt the air flow across the lifting body. Approximately 200 of these pressure ports are to be used in each of the Hypersonic Flight Experiment (HYFLITE) flight tests. Initial vibroacoustic and aerothermal testing of the pressure port designs have been completed at Langley Research vibration laboratory and the 20 MWatt 2 x 9 turbulent duct facility at Ames Research Center. The performance of the pressure ports were found to be well within the required design limits for all cases. In addition, a failure mode in which the entire pressure port assembly was removed proved to be a begin case.

Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

NASA Astrophysics Data System (ADS)

Manfredini, A.; Mazzini, M.

2017-11-01

One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.