WorldWideScience

Sample records for bimodal nuclear thermal

  1. Bimodal Nuclear Thermal Rocket Analysis Developments

    Science.gov (United States)

    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.

  2. RSMASS-D nuclear thermal propulsion and bimodal system mass models

    Science.gov (United States)

    King, Donald B.; Marshall, Albert C.

    1997-01-01

    Two relatively simple models have been developed to estimate reactor, radiation shield, and balance of system masses for a particle bed reactor (PBR) nuclear thermal propulsion concept and a cermet-core power and propulsion (bimodal) concept. The approach was based on the methodology developed for the RSMASS-D models. The RSMASS-D approach for the reactor and shield sub-systems uses a combination of simple equations derived from reactor physics and other fundamental considerations along with tabulations of data from more detailed neutron and gamma transport theory computations. Relatively simple models are used to estimate the masses of other subsystem components of the nuclear propulsion and bimodal systems. Other subsystem components include instrumentation and control (I&C), boom, safety systems, radiator, thermoelectrics, heat pipes, and nozzle. The user of these models can vary basic design parameters within an allowed range to achieve a parameter choice which yields a minimum mass for the operational conditions of interest. Estimated system masses are presented for a range of reactor power levels for propulsion for the PBR propulsion concept and for both electrical power and propulsion for the cermet-core bimodal concept. The estimated reactor system masses agree with mass predictions from detailed calculations with xx percent for both models.

  3. Finite-thrust optimization of interplanetary transfers of space vehicle with bimodal nuclear thermal propulsion

    Science.gov (United States)

    Kharytonov, Oleksii M.; Kiforenko, Boris M.

    2011-08-01

    The nuclear thermal rocket (NTR) propulsion is one of the leading promising technologies for primary space propulsion for manned exploration of the solar system due to its high specific impulse capability and sufficiently high thrust-to-weight ratio. Another benefit of NTR is its possible bimodal design, when nuclear reactor is used for generation of a jet thrust in a high-thrust mode and (with an appropriate power conversion system) as a source of electric power to supply the payload and the electric engines in a low-thrust mode. The model of the NTR thrust control was developed considering high-thrust NTR as a propulsion system of limited power and exhaust velocity. For the proposed model the control of the thrust value is accomplished by the regulation of reactor thermal power and propellant mass flow rate. The problem of joint optimization of the combination of high- and low-thrust arcs and the parameters of bimodal NTR (BNTR) propulsion system is considered for the interplanetary transfers. The interplanetary trajectory of the space vehicle is formed by the high-thrust NTR burns, which define planet-centric maneuvers and by the low-thrust heliocentric arcs where the nuclear electric propulsion (NEP) is used. The high-thrust arcs are analyzed using finite-thrust approach. The motion of the corresponding dynamical system is realized in three phase spaces concerning the departure planet-centric maneuver by means of high-thrust NTR propulsion, the low-thrust NEP heliocentric maneuver and the approach high-thrust NTR planet-centric maneuver. The phase coordinates are related at the time instants of the change of the phase spaces due to the relations between the space vehicle masses. The optimal control analysis is performed using Pontryagin's maximum principle. The numerical results are analyzed for Earth-Mars "sprint" transfer. The optimal values of the parameters that define the masses of NTR and NEP subsystems have been evaluated. It is shown that the low

  4. Conventional and Bimodal Nuclear Thermal Rocket (NTR) Artificial Gravity Mars Transfer Vehicle Concepts

    Science.gov (United States)

    Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

    2016-01-01

    A variety of countermeasures have been developed to address the debilitating physiological effects of zero-gravity (0-g) experienced by cosmonauts and astronauts during their approximately 0.5 to 1.2 year long stays in low Earth orbit (LEO). Longer interplanetary flights, combined with possible prolonged stays in Mars orbit, could subject crewmembers to up to approximately 2.5 years of weightlessness. In view of known and recently diagnosed problems associated with 0-g, an artificial gravity (AG) spacecraft offers many advantages and may indeed be an enabling technology for human flights to Mars. A number of important human factors must be taken into account in selecting the rotation radius, rotation rate, and orientation of the habitation module or modules. These factors include the gravity gradient effect, radial and tangential Coriolis forces, along with cross-coupled acceleration effects. Artificial gravity Mars transfer vehicle (MTV) concepts are presented that utilize both conventional NTR, as well as, enhanced bimodal nuclear thermal rocket (BNTR) propulsion. The NTR is a proven technology that generates high thrust and has a specific impulse (Isp) capability of approximately 900 s-twice that of today's best chemical rockets. The AG/MTV concepts using conventional Nuclear Thermal Propulsion (NTP) carry twin cylindrical International Space Station (ISS)- type habitation modules with their long axes oriented either perpendicular or parallel to the longitudinal spin axis of the MTV and utilize photovoltaic arrays (PVAs) for spacecraft power. The twin habitat modules are connected to a central operations hub located at the front of the MTV via two pressurized tunnels that provide the rotation radius for the habitat modules. For the BNTR AG/MTV option, each engine has its own closed secondary helium(He)-xenon (Xe) gas loop and Brayton Rotating Unit (BRU) that can generate 10s of kilowatts (kWe) of spacecraft electrical power during the mission coast phase

  5. A Crewed Mission to Apophis Using a Hybrid Bimodal Nuclear Thermal Electric Propulsion (BNTEP) System

    Science.gov (United States)

    Mccurdy, David R.; Borowski, Stanley K.; Burke, Laura M.; Packard, Thomas W.

    2014-01-01

    A BNTEP system is a dual propellant, hybrid propulsion concept that utilizes Bimodal Nuclear Thermal Rocket (BNTR) propulsion during high thrust operations, providing 10's of kilo-Newtons of thrust per engine at a high specific impulse (Isp) of 900 s, and an Electric Propulsion (EP) system during low thrust operations at even higher Isp of around 3000 s. Electrical power for the EP system is provided by the BNTR engines in combination with a Brayton Power Conversion (BPC) closed loop system, which can provide electrical power on the order of 100's of kWe. High thrust BNTR operation uses liquid hydrogen (LH2) as reactor coolant propellant expelled out a nozzle, while low thrust EP uses high pressure xenon expelled by an electric grid. By utilizing an optimized combination of low and high thrust propulsion, significant mass savings over a conventional NTR vehicle can be realized. Low thrust mission events, such as midcourse corrections (MCC), tank settling burns, some reaction control system (RCS) burns, and even a small portion at the end of the departure burn can be performed with EP. Crewed and robotic deep space missions to a near Earth asteroid (NEA) are best suited for this hybrid propulsion approach. For these mission scenarios, the Earth return V is typically small enough that EP alone is sufficient. A crewed mission to the NEA Apophis in the year 2028 with an expendable BNTEP transfer vehicle is presented. Assembly operations, launch element masses, and other key characteristics of the vehicle are described. A comparison with a conventional NTR vehicle performing the same mission is also provided. Finally, reusability of the BNTEP transfer vehicle is explored.

  6. A One-year, Short-Stay Crewed Mars Mission Using Bimodal Nuclear Thermal Electric Propulsion (BNTEP) - A Preliminary Assessment

    Science.gov (United States)

    Burke, Laura A.; Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

    2013-01-01

    A crewed mission to Mars poses a signi cant challenge in dealing with the physiolog- ical issues that arise with the crew being exposed to a near zero-gravity environment as well as signi cant solar and galactic radiation for such a long duration. While long sur- face stay missions exceeding 500 days are the ultimate goal for human Mars exploration, short round trip, short surface stay missions could be an important intermediate step that would allow NASA to demonstrate technology as well as study the physiological e ects on the crew. However, for a 1-year round trip mission, the outbound and inbound hy- perbolic velocity at Earth and Mars can be very large resulting in a signi cant propellant requirement for a high thrust system like Nuclear Thermal Propulsion (NTP). Similarly, a low thrust Nuclear Electric Propulsion (NEP) system requires high electrical power lev- els (10 megawatts electric (MWe) or more), plus advanced power conversion technology to achieve the lower speci c mass values needed for such a mission. A Bimodal Nuclear Thermal Electric Propulsion (BNTEP) system is examined here that uses three high thrust Bimodal Nuclear Thermal Rocket (BNTR) engines allowing short departure and capture maneuvers. The engines also generate electrical power that drives a low thrust Electric Propulsion (EP) system used for ecient interplanetary transit. This combined system can help reduce the total launch mass, system and operational requirements that would otherwise be required for equivalent NEP or Solar Electric Propulsion (SEP) mission. The BNTEP system is a hybrid propulsion concept where the BNTR reactors operate in two separate modes. During high-thrust mode operation, each BNTR provides 10's of kilo- Newtons of thrust at reasonably high speci c impulse (Isp) of 900 seconds for impulsive trans-planetary injection and orbital insertion maneuvers. When in power generation / EP mode, the BNTR reactors are coupled to a Brayton power conversion system allowing each

  7. Dynamical and statistical bimodality in nuclear fragmentation

    Science.gov (United States)

    Mallik, S.; Chaudhuri, G.; Gulminelli, F.

    2018-02-01

    The origin of bimodal behavior in the residue distribution experimentally measured in heavy ion reactions is reexamined using Boltzmann-Uehling-Uhlenbeck simulations. We suggest that, depending on the incident energy and impact parameter of the reaction, both entrance channel and exit channel effects can be at the origin of the observed behavior. Specifically, fluctuations in the reaction mechanism induced by fluctuations in the collision rate, as well as thermal bimodality directly linked to the nuclear liquid-gas phase transition, are observed in our simulations. Both phenomenologies were previously proposed in the literature but presented as incompatible and contradictory interpretations of the experimental measurements. These results indicate that heavy ion collisions at intermediate energies can be viewed as a powerful tool to study both bifurcations induced by out-of-equilibrium critical phenomena, as well as finite-size precursors of thermal phase transitions.

  8. 'Bimodal' Nuclear Thermal Rocket (BNTR) propulsion for an artificial gravity HOPE mission to Callisto

    International Nuclear Information System (INIS)

    Borowski, Stanley K.; McGuire, Melissa L.; Mason, Lee M.; Gilland, James H.; Packard, Thomas W.

    2003-01-01

    This paper summarizes the results of a year long, multi-center NASA study which examined the viability of nuclear fission propulsion systems for Human Outer Planet Exploration (HOPE). The HOPE mission assumes a crew of six is sent to Callisto. Jupiter's outermost large moon, to establish a surface base and propellant production facility. The Asgard asteroid formation, a region potentially rich in water-ice, is selected as the landing site. High thrust BNTR propulsion is used to transport the crew from the Earth-Moon L1 staging node to Callisto then back to Earth in less than 5 years. Cargo and LH2 'return' propellant for the piloted Callisto transfer vehicle (PCTV) is pre-deployed at the moon (before the crew's departure) using low thrust, high power, nuclear electric propulsion (NEP) cargo and tanker vehicles powered by hydrogen magnetoplasmadynamic (MPD) thrusters. The PCTV is powered by three 25 klbf BNTR engines which also produce 50 kWe of power for crew life support and spacecraft operational needs. To counter the debilitating effects of long duration space flight (∼855 days out and ∼836 days back) under '0-gE' conditions, the PCTV generates an artificial gravity environment of '1-gE' via rotation of the vehicle about its center-of-mass at a rate of ∼4 rpm. After ∼123 days at Callisto, the 'refueled' PCTV leaves orbit for the trip home. Direct capsule re-entry of the crew at mission end is assumed. Dynamic Brayton power conversion and high temperature uranium dioxide (UO2) in tungsten metal ''cermet'' fuel is used in both the BNTR and NEP vehicles to maximize hardware commonality. Technology performance levels and vehicle characteristics are presented, and requirements for PCTV reusability are also discussed

  9. Nuclear bimodal new vision solar system missions

    International Nuclear Information System (INIS)

    Mondt, J.F.; Zubrin, R.M.

    1996-01-01

    This paper presents an analysis of the potential mission capability using space reactor bimodal systems for planetary missions. Missions of interest include the Main belt asteroids, Jupiter, Saturn, Neptune, and Pluto. The space reactor bimodal system, defined by an Air Force study for Earth orbital missions, provides 10 kWe power, 1000 N thrust, 850 s Isp, with a 1500 kg system mass. Trajectories to the planetary destinations were examined and optimal direct and gravity assisted trajectories were selected. A conceptual design for a spacecraft using the space reactor bimodal system for propulsion and power, that is capable of performing the missions of interest, is defined. End-to-end mission conceptual designs for bimodal orbiter missions to Jupiter and Saturn are described. All missions considered use the Delta 3 class or Atlas 2AS launch vehicles. The space reactor bimodal power and propulsion system offers both; new vision open-quote open-quote constellation close-quote close-quote type missions in which the space reactor bimodal spacecraft acts as a carrier and communication spacecraft for a fleet of microspacecraft deployed at different scientific targets and; conventional missions with only a space reactor bimodal spacecraft and its science payload. copyright 1996 American Institute of Physics

  10. Pluto/Charon exploration utilizing a bi-modal PBR nuclear propulsion/power system

    Science.gov (United States)

    Venetoklis, Peter S.

    1995-01-01

    The paper describes a Pluto/Charon orbiter utilizing a bi-modal nuclear propulsion and power system based on the Particle Bed Reactor. The orbiter is sized for launch to Nuclear-Safe orbit atop a Titan IV or equivalent launch veicle. The bi-modal system provides thermal propulsion for Earth orbital departure and Pluto orbital capture, and 10 kWe of electric power for payload functions and for in-system maneuvering with ion thrusters. Ion thrusters are used to perform inclination changes about Pluto, a transfer from low Pluto orbit to low Charon orbit, and inclination changes about charon. A nominal payload can be deliverd in as little as 15 years, 1000 kg in 17 years, and close to 2000 kg in 20 years. Scientific return is enormously aided by the availability of up to 10 kWe, due to greater data transfer rates and more/better instruments. The bi-modal system can provide power at Pluto/Charon for 10 or more years, enabling an extremely robust, scientifically rewarding, and cost-effective exploration mission.

  11. A simple hydrothermal route to bimodal mesoporous nanorod {gamma}-alumina with high thermal stability

    Energy Technology Data Exchange (ETDEWEB)

    Li, Xiang; Han, Dezhi; Xue, Hongxia; Liu, Xinmei; Yan, Zifeng [China Univ. of Petroleum, Qingdao (China). State Key Lab. of Heavy Oil Processing

    2011-12-15

    In the presence of polyethylene glycol, bimodal mesoporous nanorod {gamma}-alumina was successfully synthesized via the thermal decomposition of ammonium aluminium carbonate hydroxide precursor which was prepared via hydrothermal processing with inorganic aluminium salt. The alumina exhibits high surface area (494 m{sup 2}g{sup -1}), large porosity (1.1 m{sup 3}g{sup -1}) and a particular double-pore structure after calcination at 500 C. The smaller pore diameter is concentrated on about 3 nm and the larger one is exhibited in the range of 10 - 38 nm. The scaffold-like aggregation of {gamma}-alumina nanorods endows this novel material with excellent thermal stability. A possible formation mechanism of bimodal mesoporous structure is also proposed in this study. (orig.)

  12. Application of a bi-modal PBR nuclear propulsion and power system to military missions

    Science.gov (United States)

    Venetoklis, Peter S.

    1995-01-01

    The rapid proliferation of arms technology and space access combined with current economic realities in the United States are creating ever greater demands for more capable space-based military assets. The paper illustrates that bi-modal nuclear propulsion and power based on the Particle Bed Reactor (PBR) is a high-leverage tehcnology that can maximize utility while minimizing cost. Mission benefits offered by the bi-modal PBR, including enhanced maneuverability, lifetime, survivability, payload power, and operational flexibility, are discussed. The ability to deliver desired payloads on smaller boosters is also illustrated. System descriptions and parameters for 10 kWe and 100 kWe power output levels are summarized. It is demonstrated via design exercise that bi-modal PBR dramtically enhances performance of a military satellite in geosynchronous orbit, increasing payload mass, payload power, and maneuverability.

  13. Thermal Super-Pixels for Bimodal Stress Recognition

    DEFF Research Database (Denmark)

    Irani, Ramin; Nasrollahi, Kamal; Dhall, Abhinav

    2016-01-01

    Stress is a response to time pressure or negative environmental conditions. If its stimulus iterates or stays for a long time, it affects health conditions. Thus, stress recognition is an important issue. Traditional systems for this purpose are mostly contact-based, i.e., they require a sensor...... to be in touch with the body which is not always practical. Contact-free monitoring of the stress by a camera [1, 2] can be an alternative. These systems usually utilize only an RGB or a thermal camera to recognize stress. To the best of our knowledge, the only work on fusion of these two modalities for stress...... recognition is [3] which uses a feature level fusion of the two modalities. The features in [3] are extracted directly from pixel values. In this paper we show that extracting the features from super-pixels, followed by decision level fusion results in a system outperforming [3]. The experimental results...

  14. An examination of bimodal nuclear power and propulsion benefits for outer solar system missions

    International Nuclear Information System (INIS)

    Zubrin, R.; Mondt, J.

    1996-01-01

    This paper presents the results of an analysis of the capability of nuclear bimodal systems to perform outer solar system exploration missions. Missions of interest include orbiter missions to Jupiter, Saturn, Uranus, Neptune, and Pluto. An initial technology baseline consisting of the NEBA 10 kWe, 1000 N thrust, 850 s, 1500 kg bimodal system was selected, and its performance examined against a data base for trajectories to outer solar system planetary destinations to select optimal direct and gravity assisted trajectories for study. A conceptual design for a common bimodal spacecraft capable of performing missions to all the planetary destinations was developed and made the basis of end to end mission designs for orbiter missions to Jupiter, Saturn, and Neptune. All mission designs considered use the Atlas 2AS for launch. The radiological hazard associated with using Earth gravity assists on such missions was examined and shown to be small compared to that currently accepted on Earth fly-by missions involving RTGs. It is shown that the bimodal nuclear power and propulsion system offers many attractive options for planetary missions, including both conventional planetary missions in which all instruments are carried by a single primary orbiting spacecraft, and unconventional missions in which the primary spacecraft acts as a carrier, relay, and mother ship for a fleet of micro spacecraft deployed at the planetary destination. copyright 1996 American Institute of Physics

  15. Thermal decomposition behavior of nano/micro bimodal feedstock with different solids loading

    Science.gov (United States)

    Oh, Joo Won; Lee, Won Sik; Park, Seong Jin

    2018-01-01

    Debinding is one of the most critical processes for powder injection molding. The parts in debinding process are vulnerable to defect formation, and long processing time of debinding decreases production rate of whole process. In order to determine the optimal condition for debinding process, decomposition behavior of feedstock should be understood. Since nano powder affects the decomposition behavior of feedstock, nano powder effect needs to be investigated for nano/micro bimodal feedstock. In this research, nano powder effect on decomposition behavior of nano/micro bimodal feedstock has been studied. Bimodal powders were fabricated with different ratios of nano powder, and the critical solids loading of each powder was measured by torque rheometer. Three different feedstocks were fabricated for each powder depending on solids loading condition. Thermogravimetric analysis (TGA) experiment was carried out to analyze the thermal decomposition behavior of the feedstocks, and decomposition activation energy was calculated. The result indicated nano powder showed limited effect on feedstocks in lower solids loading condition than optimal range. Whereas, it highly influenced the decomposition behavior in optimal solids loading condition by causing polymer chain scission with high viscosity.

  16. Thermal stability of bimodal microstructure in magnesium alloy AZ91 processed by ECAP

    Energy Technology Data Exchange (ETDEWEB)

    Pantělejev, Libor, E-mail: pantelejev@fme.vutbr.cz [Institute of Materials Science and Engineering, Faculty of Mechanical Engineering, Brno University of Technology, Technická 2896/2, 616 69 Brno (Czech Republic); NETME Centre, Faculty of Mechanical Engineering, Brno University of Technology, Technická 2896/2, 616 69 Brno (Czech Republic); Štěpánek, Roman [Institute of Materials Science and Engineering, Faculty of Mechanical Engineering, Brno University of Technology, Technická 2896/2, 616 69 Brno (Czech Republic); NETME Centre, Faculty of Mechanical Engineering, Brno University of Technology, Technická 2896/2, 616 69 Brno (Czech Republic); Man, Ondřej [Institute of Materials Science and Engineering, Faculty of Mechanical Engineering, Brno University of Technology, Technická 2896/2, 616 69 Brno (Czech Republic)

    2015-09-15

    The changes in microstructure of equal channel angular pressing (ECAP) processed magnesium alloy AZ91 during thermal exposure were studied in this paper. The microstructure stability was investigated by means of electron backscatter diffraction (EBSD), which allowed to measure the changes in grain size, mutual ratio of low-angle boundaries (LABs) to high-angle ones (HABs) and local lattice distortion evaluated by the kernel average misorientation (KAM) parameter. It was found experimentally that the threshold temperature at which significant grain coarsening takes place is 350 °C. No modification to mean grain diameter occurs below this temperature, nonetheless, some changes in LAB and HAB fraction, as well as in local lattice distortion, can be observed. - Highlights: • Thermal stability of bimodal UFG AZ91 alloy was assessed by means of EBSD. • Threshold temperature for pronounced grain coarsening was found at 350 °C. • Below 350 °C increase in LAB fraction and local lattice distortion takes place. • Local lattice distortion (LLD) can be well described using KAM approach. • LLD is influenced by coarsening and precipitation of Mg{sub 17}Al{sub 12} particles.

  17. Nuclear thermal/nuclear electric hybrids

    Science.gov (United States)

    Reid, B. D.

    1991-01-01

    A description is given of the nuclear thermal and nuclear electric hybrid. The specifications are described along with its mission performance. Next, the technical status, development requirements, and some cost estimates are provided.

  18. Nuclear thermal propulsion program overview

    Science.gov (United States)

    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.

  19. NASA's Nuclear Thermal Propulsion Project

    Science.gov (United States)

    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.; hide

    2015-01-01

    The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation NTP system 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 a first generation NTP 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 NTP project could also help enable high performance fission power systems and Nuclear Electric Propulsion (NEP).

  20. A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

    Energy Technology Data Exchange (ETDEWEB)

    D’Auria, F; Rohatgi, Upendra S.

    2017-01-12

    The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

  1. Nuclear thermal propulsion workshop overview

    International Nuclear Information System (INIS)

    Clark, J.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

  2. Nuclear Thermal Propulsion Development Risks

    Science.gov (United States)

    Kim, Tony

    2015-01-01

    There are clear advantages of development of a Nuclear Thermal Propulsion (NTP) for a crewed mission to Mars. NTP for in-space propulsion enables more ambitious space missions by providing high thrust at high specific impulse ((is) approximately 900 sec) that is 2 times the best theoretical performance possible for chemical rockets. Missions can be optimized for maximum payload capability to take more payload with reduced total mass to orbit; saving cost on reduction of the number of launch vehicles needed. Or missions can be optimized to minimize trip time significantly to reduce the deep space radiation exposure to the crew. NTR propulsion technology is a game changer for space exploration to Mars and beyond. However, 'NUCLEAR' is a word that is feared and vilified by some groups and the hostility towards development of any nuclear systems can meet great opposition by the public as well as from national leaders and people in authority. The public often associates the 'nuclear' word with weapons of mass destruction. The development NTP is at risk due to unwarranted public fears and clear honest communication of nuclear safety will be critical to the success of the development of the NTP technology. Reducing cost to NTP development is critical to its acceptance and funding. In the past, highly inflated cost estimates of a full-scale development nuclear engine due to Category I nuclear security requirements and costly regulatory requirements have put the NTP technology as a low priority. Innovative approaches utilizing low enriched uranium (LEU). Even though NTP can be a small source of radiation to the crew, NTP can facilitate significant reduction of crew exposure to solar and cosmic radiation by reducing trip times by 3-4 months. Current Human Mars Mission (HMM) trajectories with conventional propulsion systems and fuel-efficient transfer orbits exceed astronaut radiation exposure limits. Utilizing extra propellant from one additional SLS launch and available

  3. Unique nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Culver, D.W.; Rochow, R.

    1993-06-01

    In January, 1992, a new, advanced nuclear thermal rocket engine (NTRE) concept intended for manned missions to the moon and to Mars was introduced (Culver, 1992). This NTRE promises to be both shorter and lighter in weight than conventionally designed engines, because its forward flowing reactor is located within an expansion-deflection rocket nozzle. The concept has matured during the year, and this paper discusses a nearer term version that resolves four open issues identified in the initial concept: (1) the reactor design and cooling scheme simplification while retaining a high pressure power balance option; (2) elimination need for a new, uncooled nozzle throat material suitable for long life application; (3) a practical provision for reactor power control; and (4) use of near-term, long-life turbopumps

  4. Philosophy for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Buden, D.; Madsen, W.; Redd, L.

    1993-01-01

    The philosophy used for development of nuclear thermal propulsion will determine the cost, schedule and risk associated with the activities. As important is the impression of the decision makers. If the development cost is higher than the product value, it is doubtful that funding will ever be available. On the other hand, if the development supports the economic welfare of the country with a high rate of return, the probability of funding greatly increases. The philosophy is divided into: realism, design, operations and qualification. ''Realism'' addresses such items as political acceptability, potential customers, robustness-flexibility, public acceptance, decisions as needed, concurrent engineering, and the possible role of the CIS. ''Design'' addresses ''minimum requirement,'' built in safety and reliability redundancy, emphasize on eliminating risk at lowest levels, and the possible inclusion of electric generation. ''Operations'' addresses sately, environment, operations, design margins and degradation modes. ''Qualification'' addresses testing needs and test facilities

  5. Space Nuclear Thermal Propulsion (SNTP) program

    Science.gov (United States)

    Bleeker, Gary A.

    1993-01-01

    An overview of the Space Nuclear Thermal Propulsion program is presented in graphic form. A program organizational chart is presented that shows the government and industry participants. Enabling technologies and test facilities and approaches are also addressed.

  6. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  7. Nuclear waste package thermal performance

    International Nuclear Information System (INIS)

    Lundberg, W.

    1985-01-01

    Given the geology, the corrosion of deep geologic nuclear waste packages depends largely on the package temperature history. Factors affecting package temperature are described, and predictions of package temperatures and resulting corrosion vs time relationships are presented and discussed for candidate geologies

  8. A development approach for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Buden, D.

    1992-01-01

    The cost and time to develop nuclear thermal propulsion systems are very approach dependent. The objectives addressed are the development of an ''acceptable'' nuclear thermal propulsion system that can be used as part of the transportation system for people to explore Mars and the enhancement performance of other missions, within highly constrained budgets and schedules. To accomplish this, it was necessary to identify the cost drivers considering mission parameters, safety of the crew, mission success, facility availability and time and cost to construct new facilities, qualification criteria, status of technologies, management structure, and use of such system engineering techniques as concurrent engineering

  9. Macroscopic dynamics of thermal nuclear excitations

    International Nuclear Information System (INIS)

    Bastrukov, S.I.; Deak, F.; Kiss, A.; Seres, Z.

    1989-11-01

    The concept of kinetic temperature as a local dynamical variable of thermal nuclear collective motion is formulated using long-mean-free-path approach based on the Landau-Vlasov kinetic equation. In the Fermi drop model the thermal fluid dynamics of the spherical nucleus is analyzed. It is shown that in a compressible Fermi liquid the temperature pulses propagate in the form of spherical wave in phase with the acoustic wave. The thermal and compressional excitations are caused by the isotropic harmonic oscillations of the Fermi sphere in momentum space. (author) 25 refs.; 2 figs

  10. Nuclear Thermal Propulsion for Advanced Space Exploration

    Science.gov (United States)

    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).

  11. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    Science.gov (United States)

    Emrich, William J.

    2008-01-01

    To support a potential future development of a nuclear thermal rocket engine, 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 test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

  12. MSFC nuclear thermal propulsion technology program

    Science.gov (United States)

    Swint, Shane

    1993-01-01

    Viewgraphs on non-nuclear materials assessment, nuclear thermal propulsion (NTP) turbomachinery technologies, and high temperature superconducting magnetic bearing technology are presented. The objective of the materials task is to identify and evaluate candidate materials for use in NTP turbomachinery and propellant feed system applications. The objective of the turbomachinery technology task is to develop and validate advanced turbomachinery technologies at the component and turbopump assembly levels. The objective of the high temperature superconductors (HTS) task is to develop and validate advanced technology for HTS passive magnetic/hydrostatic bearing.

  13. Nuclear thermal propulsion engine cost trade studies

    International Nuclear Information System (INIS)

    Paschall, R.K.

    1993-01-01

    The NASA transportation strategy for the Mars Exploration architecture includes the use of nuclear thermal propulsion as the primary propulsion system for Mars transits. It is anticipated that the outgrowth of the NERVA/ROVER programs will be a nuclear thermal propulsion (NTP) system capable of providing the propulsion for missions to Mars. The specific impulse (Isp) for such a system is expected to be in the 870 s range. Trade studies were conducted to investigate whether or not it may be cost effective to invest in a higher performance (Isp>870 s) engine for nuclear thermal propulsion for missions to Mars. The basic cost trades revolved around the amount of mass that must be transported to low-earth orbit prior to each Mars flight and the cost to launch that mass. The mass required depended on the assumptions made for Mars missions scenarios including piloted/cargo flights, number of Mars missions, and transit time to Mars. Cost parameters included launch cost, program schedule for development and operations, and net discount rate. The results were very dependent on the assumptions that were made. Under some assumptions, higher performance engines showed cost savings in the billions of dollars; under other assumptions, the additional cost to develop higher performance engines was not justified

  14. Nuclear reactor vessel fuel thermal insulating barrier

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-19

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

  15. Multiphase flow dynamics 5 nuclear thermal hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2015-01-01

    This Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step demons...

  16. Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2012-01-01

    The present Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step...

  17. Thermal imaging for the nuclear power industry

    International Nuclear Information System (INIS)

    Caruso, F.T.

    1986-01-01

    While, on its face, thermal imaging for the nuclear power industry bears little difference from infra-red imaging for the industrial complex, as a whole (in so far as equipment, trained personnel, and technique, are concerned), there are vast differences with regard to access, training, and movement within a nuclear facility. For the un-initiated, working inside of a nuclear power plant can be a series of frustrations, fraught with time wasting periods of training, classes, and seminars,--interspersed with an unending line of meetings and project planning sessions. For those used to working within the system, the experience can be of tremendous satisfaction in undertaking, and successfully completing a project under some very difficult circumstances

  18. Applied mathematical methods in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Ransom, V.H.; Trapp, J.A.

    1983-01-01

    Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated

  19. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  20. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Directory of Open Access Journals (Sweden)

    Seung Hyun Nam

    2015-10-01

    Full Text Available Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER, for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MWth and an electricity generation mode of 100 kWth, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  1. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon

    2015-01-01

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW th and an electricity generation mode of 100 kW th , equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics

  2. Shielding Development for Nuclear Thermal Propulsion

    Science.gov (United States)

    Caffrey, Jarvis A.; Gomez, Carlos F.; Scharber, Luke L.

    2015-01-01

    Radiation shielding analysis and development for the Nuclear Cryogenic Propulsion Stage (NCPS) effort is currently in progress and preliminary results have enabled consideration for critical interfaces in the reactor and propulsion stage systems. Early analyses have highlighted a number of engineering constraints, challenges, and possible mitigating solutions. Performance constraints include permissible crew dose rates (shared with expected cosmic ray dose), radiation heating flux into cryogenic propellant, and material radiation damage in critical components. Design strategies in staging can serve to reduce radiation scatter and enhance the effectiveness of inherent shielding within the spacecraft while minimizing the required mass of shielding in the reactor system. Within the reactor system, shield design is further constrained by the need for active cooling with minimal radiation streaming through flow channels. Material selection and thermal design must maximize the reliability of the shield to survive the extreme environment through a long duration mission with multiple engine restarts. A discussion of these challenges and relevant design strategies are provided for the mitigation of radiation in nuclear thermal propulsion.

  3. Oxygen Containment System Options for Nuclear Thermal Propulsion Testing

    Data.gov (United States)

    National Aeronautics and Space Administration — All nuclear thermal propulsion (NTP) ground testing conducted in the 1950s and 1960s during the ROVER/(Nuclear Engine Rocket Vehicle Application (NERVA) program...

  4. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

  5. Thermally-insulating layer for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The thermally-insulating layer has been designed both for insulating surfaces within the core of a nuclear reactor and transmitting loads such as the core-weight. Said layer comprises a layer of bricks and a layer of tiles with smaller clearance between the tiles than between the bricks, the latter having a reduced cross-section against the tiles so as to be surrounded by relatively large interconnected ducts forming a continuous chamber behind the tile-layer in order to induce a substantial decreases in the transverse flow of the reactor-core coolant. The core preferably comprises hexagonal columns supported by rhomb-shaped plates, with channels distributed so as to mix the coolant of twelve columns. The plates are separated from support-tiles by means of pillars [fr

  6. System model development for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Walton, J.T.; Perkins, K.R.; Buksa, J.J.; Worley, B.A.; Dobranich, D.

    1992-01-01

    A critical enabling technology in the evolutionary development of nuclear thermal propulsion (NTP) is the ability to predict the system performance under a variety of operating conditions. Since October 1991, US (DOE), (DOD) and NASA have initiated critical technology development efforts for NTP systems to be used on Space Exploration Initiative (SEI) missions to the Moon and Mars. This paper presents the strategy and progress of an interagency NASA/DOE/DOD team for NTP system modeling. It is the intent of the interagency team to develop several levels of computer programs to simulate various NTP systems. An interagency team was formed for this task to use the best capabilities available and to assure appropriate peer review. The vision and strategy of the interagency team for developing NTP system models will be discussed in this paper. A review of the progress on the Level 1 interagency model is also presented

  7. Nuclear Thermal Rocket Simulation in NPSS

    Science.gov (United States)

    Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

    2013-01-01

    Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

  8. Ultrahigh Specific Impulse Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Anne Charmeau; Brandon Cunningham; Samim Anghaie

    2009-02-09

    Research on nuclear thermal propulsion systems (NTP) have been in forefront of the space nuclear power and propulsion due to their design simplicity and their promise for providing very high thrust at reasonably high specific impulse. During NERVA-ROVER program in late 1950's till early 1970's, the United States developed and ground tested about 18 NTP systems without ever deploying them into space. The NERVA-ROVER program included development and testing of NTP systems with very high thrust (~250,000 lbf) and relatively high specific impulse (~850 s). High thrust to weight ratio in NTP systems is an indicator of high acceleration that could be achieved with these systems. The specific impulse in the lowest mass propellant, hydrogen, is a function of square root of absolute temperature in the NTP thrust chamber. Therefor optimizing design performance of NTP systems would require achieving the highest possible hydrogen temperature at reasonably high thrust to weight ratio. High hydrogen exit temperature produces high specific impulse that is a diret measure of propellant usage efficiency.

  9. Nuclear Thermal Rocket (NTR) Development Risk Communication

    Science.gov (United States)

    Kim, Tony

    2014-01-01

    There are clear advantages of development of a Nuclear Thermal Rocket (NTR) for a crewed mission to Mars. NTR for in-space propulsion enables more ambitious space missions by providing high thrust at high specific impulse (approximately 900 sec) that is 2 times the best theoretical performance possible for chemical rockets. Missions can be optimized for maximum payload capability to take more payload with reduced total mass to orbit; saving cost on reduction of the number of launch vehicles needed. Or missions can be optimized to minimize trip time significantly to reduce the deep space radiation exposure to the crew. NTR propulsion technology is a game changer for space exploration. However, "NUCLEAR" is a word that is feared and vilified by some groups and the hostility towards development of any nuclear systems can meet great opposition by the public as well as from national leaders and people in authority. Communication of nuclear safety will be critical to the success of the development of the NTR. Why is there a fear of nuclear? A bomb that can level a city is a scary weapon. The first and only times the Nuclear Bomb was used in a war was on Hiroshima and Nagasaki during World War 2. The "Little Boy" atomic bomb was dropped on Hiroshima on August 6, 1945 and the "Fat Man" on Nagasaki 3 days later on August 9th. Within the first 4 months of bombings, 90- 166 thousand people died in Hiroshima and 60-80 thousand died in Nagasaki. It is important to note for comparison that over 500 thousand people died and 5 million made homeless due to strategic bombing (approximately 150 thousand tons) of Japanese cities and war assets with conventional non-nuclear weapons between 1942- 1945. A major bombing campaign of "firebombing" of Tokyo called "Operation Meetinghouse" on March 9 and 10 consisting of 334 B-29's dropped approximately1,700 tons of bombs around 16 square mile area and over 100 thousand people have been estimated to have died. The declaration of death is very

  10. Thermal efficiency improvements - an imperative for nuclear generating stations

    International Nuclear Information System (INIS)

    Hassanien, S.; Rouse, S.

    1997-01-01

    A one and a half percent thermal performance improvement of Ontario Hydro's operating nuclear units (Bruce B, Pickering B, and Darlington) means almost 980 GWh are available to the transmission system (assuming an 80% capacity factor). This is equivalent to the energy consumption of 34,000 electrically-heated homes in Ontario, and worth more than $39 million in revenue to Ontario Hydro Nuclear Generation. Improving nuclear plant thermal efficiency improves profitability (more GWh per unit of fuel) and competitiveness (cost of unit energy), and reduces environmental impact (less spent fuel and nuclear waste). Thermal performance will naturally decrease due to the age of the units unless corrective action is taken. Most Ontario Hydro nuclear units are ten to twenty years old. Some common causes for loss of thermal efficiency are: fouling and tube plugging of steam generators, condensers, and heat exchangers; steam leaks in the condenser due to valve wear, steam trap and drain leaks; deposition, pitting, cracking, corrosion, etc., of turbine blades; inadequate feedwater metering resulting from corrosion and deposition. This paper stresses the importance of improving the nuclear units' thermal efficiency. Ontario Hydro Nuclear has demonstrated energy savings results are achievable and affordable. Between 1994 and 1996, Nuclear reduced its energy use and improved thermal efficiency by over 430,000 MWh. Efficiency improvement is not automatic - strategies are needed to be effective. This paper suggests practical strategies to systematically improve thermal efficiency. (author)

  11. Thermal analysis of spent nuclear fuels repository

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, F.; Salome, J.; Cardoso, F.; Velasquez, C.E.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Asa Norte, Brazilia (Brazil); Viana, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear-CNEN, Rua Gal Severiano, n 90 - Botafogo, 22290-901, Rio de Janeiro, RJ (Brazil)

    2016-07-01

    In the first part, Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) spent fuels (SF) were evaluated to the thermal of the spent fuel pool (SFP) without an external cooling system. The goal is to compare the water boiling time of the pool storing different types of spent nuclear fuels. This study used the software ANSYS Workbench 16.2 - student version. For the VHTR, two types of fuel were analyzed: (Th,TRU)O{sub 2} and UO{sub 2}. This part of the studies were performed for wet storage condition using a single type of SF and decay heat values at times t=0 and t=10 years after the reactor discharge. The ANSYS CFX module was used and the results show that the time that water takes to reach the boiling point varies from 2.4 minutes for the case of VHTR-(Th,TRU)O{sub 2} SF at time t=0 year after reactor discharge until 32.4 hours for the case of PWR SF at time t=10 years after the discharge reactor. The second part of this work consists of modeling a geological repository. Firstly, the temperature evaluation of the spent fuel from a PWR was analyzed. A PWR canister was simulated using the ANSYS transient thermal module. Then the temperature of canister could be computed during the time spent on a portion of a geological repository. The mean temperature on the canister surface increased during the first nine years, reaching a plateau at 35.5 C. degrees between the tenth and twentieth years after the geological disposal. The idea is to extend this study for the other systems analyzed in the first part. The idea is to include in the study, the spent fuels from VHTR and ADS and to compare the canister behavior using different spent fuels. (authors)

  12. Turbopump options for nuclear thermal rockets

    International Nuclear Information System (INIS)

    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. 10 refs

  13. Test facilities for evaluating nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Beck, D.F.; Allen, G.C.; Shipers, L.R.; Dobranich, D.; Ottinger, C.A.; Harmon, C.D.; Fan, W.C.; Todosow, M.

    1992-01-01

    Interagency panels evaluating nuclear thermal propulsion (NTP) development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and baseline performance of some of the major subsystems designed to support a proposed ground test complex for evaluating nuclear thermal propulsion fuel elements and engines being developed for the Space Nuclear Thermal Propulsion (SNTP) program. Some preliminary results of evaluating this facility for use in testing other NTP concepts are also summarized

  14. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as a propulsion concept which could provide the fastest trip times to Mars and as the preferred concept for...

  15. Superconducting Electric Boost Pump for Nuclear Thermal Propulsion, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — A submersible, superconducting electric boost pump sized to meet the needs of future Nuclear Thermal Propulsion systems in the 25,000 lbf thrust range is proposed....

  16. Corredor Bimodal Cafetero

    OpenAIRE

    Duque Escobar, Gonzalo

    2015-01-01

    El Corredor Bimodal Cafetero es un proyecto de infraestructura estratégica que articula la Hidrovía del Magdalena con el Corredor Férreo del río Cauca, inscrito en el Plan Nacional de Desarrollo 2014/2018 y financiable con la salida de 30 mil toneladas diarias de carbón andino a la cuenca del Pacífico. Incluye el Túnel Cumanday para cruzar la Cordillera Central, el Ferrocarril Cafetero de 150 km y 3% de pendiente entre La Dorada y el Km 41, y la Transversal Cafetera de 108 km para una vía de...

  17. On the thermal properties of polarized nuclear matter

    International Nuclear Information System (INIS)

    Hassan, M.Y.M.; Montasser, S.S.; Ramadan, S.

    1979-08-01

    The thermal properties of polarized nuclear matter are calculated using Skyrme III interaction modified by Dabrowski for polarized nuclear matter. The temperature dependence of the volume, isospin, spin and spin isospin pressure and energies are determined. The temperature, isospin, spin and spin isospin dependence of the equilibrium Fermi momentum is also discussed. (author)

  18. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  19. Application of thermal-hydraulic codes in the nuclear sector

    International Nuclear Information System (INIS)

    Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.

    2011-01-01

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

  20. Space Nuclear Thermal Propulsion (SNTP) Air Force facility

    Science.gov (United States)

    Beck, David F.

    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.

  1. Thermal coupling system analysis of a nuclear desalination plant

    International Nuclear Information System (INIS)

    Adak, A.K.; Srivastava, V.K.; Tewari, P.K.

    2010-01-01

    When a nuclear reactor is used to supply steam for desalination plant, the method of coupling has a significant technical and economic impact. The exact method of coupling depends upon the type of reactor and type of desalination plant. As a part of Nuclear Desalination Demonstration Project (NDDP), BARC has successfully commissioned a 4500 m 3 /day MSF desalination plant coupled to Madras Atomic Power Station (MAPS) at Kalpakkam. Desalination plant coupled to nuclear power plant of Pressurized Heavy Water Reactor (PHWR) type is a good example of dual-purpose nuclear desalination plant. This paper presents the thermal coupling system analysis of this plant along with technical and safety aspects. (author)

  2. Nuclear Thermal Propulsion (prior to FY15: Nuclear Cryogenic Propulsion Stage)

    Data.gov (United States)

    National Aeronautics and Space Administration — A key goal of the project is to address critical, long-term nuclear thermal propulsion (NTP) technology challenges and issues through development, analysis, and...

  3. Effluent treatment options for nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    Shipers, L.R.; Brockmann, J.E.

    1992-01-01

    A variety of approaches for handling effluent from nuclear thermal propulsion system ground tests in an environmentally acceptable manner are discussed. The functional requirements of effluent treatment are defined and concept options are presented within the framework of these requirements. System concepts differ primarily in the choice of fission-product retention and waste handling concepts. The concept options considered range from closed cycle (venting the exhaust to a closed volume or recirculating the hydrogen in a closed loop) to open cycle (real time processing and venting of the effluent). This paper reviews the strengths and weaknesses of different methods to handle effluent from nuclear thermal propulsion system ground tests

  4. Nuclear thermal rocket nozzle testing and evaluation program

    Science.gov (United States)

    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.

  5. Engine cycle design considerations for nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Collins, J.T.

    1993-01-01

    A top-level study was performed which addresses nuclear thermal propulsion system engine cycle options and their applicability to support future Space Exploration Initiative manned lunar and Mars missions. Technical and development issues associated with expander, gas generator, and bleed cycle near-term, solid core nuclear thermal propulsion engines are identified and examined. In addition to performance and weight the influence of the engine cycle type on key design selection parameters such as design complexity, reliability, development time, and cost are discussed. Representative engine designs are presented and compared. Their applicability and performance impact on typical near-term lunar and Mars missions are shown

  6. Coupled fast-thermal system at the 'RB' nuclear reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    1987-01-01

    The results of the analyses of the possibility of the coupled fast-thermal system (CFTS) design at the 'RB' nuclear reactor are shown. As the proof of the theoretical analyses the first stage CFTS-1 has been designed, realized, and tested. The excellent agreement between the results of the CFTS-1 studies and the theoretical predictions opens a straight way to the second, the final stage - realization of the designed CFST at the 'RB' nuclear reactor. (author)

  7. Coupled fast-thermal system at the 'RB' nuclear reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    1987-04-01

    The results of the analyses of the possibility of the coupled fast-thermal system (CFTS) design at the 'RB' nuclear reactor are shown. As the proof of the theoretical analyses the first stage CFTS-1 has been designed, realized, and tested. The excellent agreement between the results of the CFTS-1 studies and the theoretical predictions opens a straight way to the second, the final stage - realization of the designed CFST at the 'RB' nuclear reactor. (author)

  8. Tutorial on nuclear thermal propulsion safety for Mars

    International Nuclear Information System (INIS)

    Buden, D.

    1992-01-01

    Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments

  9. Thermal Analysis of Storage Cans Containing Special Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Jerrell, J.W.

    2000-11-17

    A series of thermal analyses have been completed for ten storage can configurations representing various cases of materials stored in F-Area. The analyses determine the temperatures of the cans, the special nuclear material, and the air sealed within the cans. Analyses to aid in understanding the effect of oxide accumulation and metal aging on temperatures are also included.

  10. Application of thermal neutrons in testing of nuclear facilities materials

    International Nuclear Information System (INIS)

    Milczarek, J.J.

    2008-01-01

    Recent applications of thermal neutrons in testing of materials used in nuclear facilities are presented. The neutron radiography technique, characterization of residual stresses with neutron diffraction and small angle neutron scattering is considered in some detail. The results of testing of fuel elements, steel used for pressurized reactor vessels and changes induced in control rods are discussed. (author)

  11. Some perspectives in nuclear astrophysics on non-thermal phenomena

    International Nuclear Information System (INIS)

    Tatischeff, V.

    2012-01-01

    In this HDR (Accreditation to Supervise Researches) report, the author presents and comments his research activities on nuclear phenomena in stellar eruptions (solar eruptions, lithium nucleosynthesis in stellar eruptions), on particle acceleration in shock waves of stellar explosions (diffusive acceleration by shock wave, particle acceleration in symbiotic novae, particle acceleration in radio-detected supernovae), of research on low energy cosmic rays (galactic emission of nuclear gamma rays, non thermal soft X rays as new tracer of accelerated particles), and on the origin of short period radioactivities in the primitive solar system (extinguished radio-activities and formation of the solar system, origin of berylium-10 in the primitive solar system). The author concludes with some perspectives on non thermal phenomena in nuclear astrophysics, and on research and development for the future of medium-energy gamma astronomy [fr

  12. Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs

    Energy Technology Data Exchange (ETDEWEB)

    James Werner

    2014-07-01

    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 be 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.

  13. Cycle Trades for Nuclear Thermal Rocket Propulsion Systems

    Science.gov (United States)

    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.

  14. An historical collection of papers on nuclear thermal propulsion

    Science.gov (United States)

    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.

  15. Innovative Approaches to Development and Ground Testing of Advanced Bimodal Space Power and Propulsion Systems

    International Nuclear Information System (INIS)

    Hill, T.; Noble, C.; Martinell, J.; Borowski, S.

    2000-01-01

    The last major development effort for nuclear power and propulsion systems ended in 1993. Currently, there is not an initiative at either the National Aeronautical and Space Administration (NASA) or the U.S. Department of Energy (DOE) that requires the development of new nuclear power and propulsion systems. Studies continue to show nuclear technology as a strong technical candidate to lead the way toward human exploration of adjacent planets or provide power for deep space missions, particularly a 15,000 lbf bimodal nuclear system with 115 kW power capability. The development of nuclear technology for space applications would require technology development in some areas and a major flight qualification program. The last major ground test facility considered for nuclear propulsion qualification was the U.S. Air Force/DOE Space Nuclear Thermal Propulsion Project. Seven years have passed since that effort, and the questions remain the same, how to qualify nuclear power and propulsion systems for future space flight. It can be reasonably assumed that much of the nuclear testing required to qualify a nuclear system for space application will be performed at DOE facilities as demonstrated by the Nuclear Rocket Engine Reactor Experiment (NERVA) and Space Nuclear Thermal Propulsion (SNTP) programs. The nuclear infrastructure to support testing in this country is aging and getting smaller, though facilities still exist to support many of the technology development needs. By renewing efforts, an innovative approach to qualifying these systems through the use of existing facilities either in the U.S. (DOE's Advance Test Reactor, High Flux Irradiation Facility and the Contained Test Facility) or overseas should be possible

  16. To MARS and Beyond with Nuclear Power - Design Concept of Korea Advanced Nuclear Thermal Engine Rocket

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Chang, Soon Heung

    2013-01-01

    The President Park of ROK has also expressed support for space program promotion, praising the success of NARO as evidence of a positive outlook. These events hint a strong signal that ROK's space program will be accelerated by the national eager desire. In this national eager desire for space program, the policymakers and the aerospace engineers need to pay attention to the advanced nuclear technology of ROK that is set to a major world nuclear energy country, even exporting the technology. The space nuclear application is a very much attractive option because its energy density is the most enormous among available energy sources in space. This paper presents the design concept of Korea Advanced Nuclear Thermal Engine Rocket (KANuTER) that is one of the advanced nuclear thermal rocket engine developing in Korea Advanced Institute of Science and Technology (KAIST) for space application. Solar system exploration relying on CRs suffers from long trip time and high cost. In this regard, nuclear propulsion is a very attractive option for that because of higher performance and already demonstrated technology. Although ROK was a late entrant into elite global space club, its prospect as a space racer is very bright because of the national eager desire and its advanced technology. Especially it is greatly meaningful that ROK has potential capability to launch its nuclear technology into space as a global nuclear energy leader and a soaring space adventurer. In this regard, KANuTER will be a kind of bridgehead for Korean space nuclear application

  17. Thermal performance test for steam turbine of nuclear power plants

    International Nuclear Information System (INIS)

    Bu Yubing; Xu Zongfu; Wang Shiyong

    2014-01-01

    Through study of steam turbine thermal performance test of CPR1000 nuclear power plant, we solve the enthalpy calculation problems of the steam turbine in wet steam zone using heat balance method which can help to figure out the real overall heat balance diagram for the first time, and we develop a useful software for thermal heat balance calculation. Ling'ao phase II as an example, this paper includes test instrument layout, system isolation, risk control, data acquisition, wetness measurement, heat balance calculation, etc. (authors)

  18. Interim report on nuclear waste depository thermal analysis

    International Nuclear Information System (INIS)

    Altenbach, T.J.

    1978-01-01

    A thermal analysis of a deep geologic depository for spent nuclear fuel is being conducted. The TRUMP finite difference heat transfer code is used to analyze a 3-dimensional model of the depository. The model uses a unit cell consisting of one spent fuel canister buried in salt beneath a ventilated room in the depository. A base case was studied along with several parametric variations. It is concluded that this method is appropriate for analyzing the thermal response of the system, and that the most important parameter in determining the maximum temperatures is the canister heat generation rate. The effects of room ventilation and different depository media are secondary

  19. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    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.

  20. Thermal properties of nuclear matter under the periodic boundary condition

    International Nuclear Information System (INIS)

    Otuka, Naohiko; Ohnishi, Akira

    1999-01-01

    We present the thermal properties of nuclear matter under the periodic boundary condition by the use of our hadronic nucleus-nucleus cascade model (HANDEL) which is developed to treat relativistic heavy-ion collisions from BNL-AGS to CERN-SPS. We first show some results of p-p scattering calculation in our new version which is improved in order to treat isospin ratio and multiplicity more accurately. We then display the results of calculation of nuclear matter with baryon density ρ b = 0.77 fm 3 at some energy densities. Time evolution of particle abundance and temperature are shown. (author)

  1. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  2. Performance Comparisons of Nanoaluminum, Coated Microaluminum and Their Bimodal Mixtures

    Science.gov (United States)

    Woody, D. L.; Dokhan, A.; Johnson, C. E.

    2004-07-01

    Comparison studies of materials containing standard nano aluminum (ultrafine) and micro aluminum coated with BaSO4 were performed. Differential thermal analysis and thermogravimetric analysis output were used to observe the effect of adding an unconventional coating to micron-sized aluminum particle materials. These results were compared to those of ultrafine aluminum particles. Bimodal combinations of ultrafine aluminum and micron-sized aluminum (coated and uncoated) were observed also. These preliminary results showed an interaction between the ultrafine aluminum (UFAL) and micron-sized aluminum in bimodal mixtures.

  3. Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code

    International Nuclear Information System (INIS)

    Hall, M.L.; Rider, W.J.; Cappiello, M.W.

    1992-01-01

    The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper

  4. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    Science.gov (United States)

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  5. Thermal effluents from nuclear power plant influences species distribution and thermal tolerance of fishes in reservoirs

    International Nuclear Information System (INIS)

    Pal, A.K.; Das, T.; Dalvi, R.S.; Bagchi, S.; Manush, S.M.; Ayyappan, S.; Chandrachoodan, P.P.; Apte, S.K.; Ravi, P.M.

    2007-01-01

    During electricity generation water bodies like reservoir act as a heat sink for thermal effluent discharges from nuclear power plant. We hypothesized that the fish fauna gets distributed according to their temperature preference in the thermal gradient. In a simulated environment using critical thermal methodology (CTM), we assessed thermal tolerance and metabolic profile of fishes (Puntius filamentosus, Parluciosoma daniconius, Ompok malabaricus, Mastacembelus armatus, Labeo calbasu, Horabragrus brachysoma, Etroplus suratensis, Danio aequipinnatus and Gonoproktopterus curmuca) collected from Kadra reservoir in Karnataka state. Results of CTM tests agrees with the species abundance as per the temperature gradient formed in the reservoir due to thermal effluent discharge. E. suratensis and H. brachysoma) appear to be adapted to high temperature (with high CTMax and CTMin values) and are in abundance at point of thermal discharge. Similarly, P. daniconius, appear to be adapted to cold (low CTM values) is in abundance in lower stretches of Kadra reservoir. Overall results indicate that discharge form nuclear power plant influences the species biodiversity in enclosed water bodies. (author)

  6. Thermal investigation of nuclear waste disposal in space

    Science.gov (United States)

    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.

  7. IMPULSE---an advanced, high performance nuclear thermal propulsion system

    International Nuclear Information System (INIS)

    Petrosky, L.J.; Disney, R.K.; Mangus, J.D.; Gunn, S.A.; Zweig, H.R.

    1993-01-01

    IMPULSE is an advanced nuclear propulsion engine for future space missions based on a novel conical fuel. Fuel assemblies are formed by stacking a series of truncated (U, Zr)C cones with non-fueled lips. Hydrogen flows radially inward between the cones to a central plenum connected to a high performance bell nozzle. The reference IMPULSE engine rated at 75,000 lb thrust and 1800 MWt weighs 1360 kg and is 3.65 meters in height and 81 cm in diameter. Specific impulse is estimated to be 1000 for a 15 minute life at full power. If longer life times are required, the operating temperature can be reduced with a concomitant decrease in specific impulse. Advantages of this concept include: well defined coolant paths without outlet flow restrictions; redundant orificing; very low thermal gradients and hence, thermal stresses, across the fuel elements; and reduced thermal stresses because of the truncated conical shape of the fuel elements

  8. Fatigue design of nuclear class 1 piping considering thermal stratification

    International Nuclear Information System (INIS)

    Kweon, Hyeong Do; Kim, Jong Sung; Lee, Kang Yong

    2008-01-01

    In ASME B and PV Code, Section III, Subsection NB-3600, thermal stratification is not taken into account to determine the peak stress intensity range for fatigue design of nuclear class 1 piping. Therefore, the effects of several parameters such as boundary layer thickness, temperature difference, stratification length, wall thickness, inner diameter and material properties on peak temperature and peak stress intensity due to non-linear temperature distribution of thermal stratification in a pipe cross-section are studied through the numerical parametric study. The results of the parametric study are closely examined and consolidated to introduce an additional term into the equation of ASME so that the modified equation can be used to determine the peak stress intensity range due to all loads including thermal stratification

  9. Design considerations for Mars transfer vehicles using nuclear thermal propulsion

    Science.gov (United States)

    Emrich, William J.

    1995-01-01

    The design of a Mars Transfer Vehicle (MTV) utilizing nuclear propulsion will require that careful consideration be given to the nuclear radiation environment in which it will operate. The extremely high neutron and gamma fluxes characteristic of nuclear thermal propulsion systems will cause significant heating of the fluid systems in close proximity to the reactor, especially in the lower propellant tanks. Crew radiation doses are also a concern particularly late in a mission when there is less shielding from the propellant tanks. In this study, various vehicle configuration and shielding strategies were examined and the resulting time dependent radiation fields evaluated. A common cluster of three particle bed reactor (PBR) engines were used in all configurations examined. In general, it appears that long, relatively narrow vehicles perform the best from a radiation standpoint, however, good shield optimization will be critical in maintaining a low radiation environment while minimizing the shield weight penalty.

  10. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

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

    Science.gov (United States)

    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.

  12. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  13. Nuclear Thermal Propulsion (NTP): A Proven Growth Technology for Human NEO/Mars Exploration Missions

    Science.gov (United States)

    Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

    2012-01-01

    The nuclear thermal rocket (NTR) represents the next "evolutionary step" in high performance rocket propulsion. Unlike conventional chemical rockets that produce their energy through combustion, the NTR derives its energy from fission of Uranium-235 atoms contained within fuel elements that comprise the engine s reactor core. Using an "expander" cycle for turbopump drive power, hydrogen propellant is raised to a high pressure and pumped through coolant channels in the fuel elements where it is superheated then expanded out a supersonic nozzle to generate high thrust. By using hydrogen for both the reactor coolant and propellant, the NTR can achieve specific impulse (Isp) values of 900 seconds (s) or more - twice that of today s best chemical rockets. From 1955 - 1972, twenty rocket reactors were designed, built and ground tested in the Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) programs. These programs demonstrated: (1) high temperature carbide-based nuclear fuels; (2) a wide range of thrust levels; (3) sustained engine operation; (4) accumulated lifetime at full power; and (5) restart capability - all the requirements needed for a human Mars mission. Ceramic metal "cermet" fuel was pursued as well, as a backup option. The NTR also has significant "evolution and growth" capability. Configured as a "bimodal" system, it can generate its own electrical power to support spacecraft operational needs. Adding an oxygen "afterburner" nozzle introduces a variable thrust and Isp capability and allows bipropellant operation. In NASA s recent Mars Design Reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, versatile vehicle design, simple assembly, and growth potential. In contrast to other advanced propulsion options, no large technology scale-ups are required for NTP either. In fact, the smallest engine tested during the Rover program

  14. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  15. NERVA-Derived Nuclear Thermal Propulsion Dual Mode Operation

    Science.gov (United States)

    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.

  16. Nuclear thermal propulsion transportation systems for lunar/Mars exploration

    International Nuclear Information System (INIS)

    Clark, J.S.; Borowski, S.K.; Mcilwain, M.C.; Pellaccio, D.G.

    1992-09-01

    Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the next generation of space propulsion systems - the key to space exploration

  17. Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

    Science.gov (United States)

    Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

    2015-01-01

    The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at

  18. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering; Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    2016-09-20

    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 effects 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.

  19. Discussion on several problems in thermal pollution control of coastal nuclear power plants

    International Nuclear Information System (INIS)

    Li Jingjing; Zhang Yongxing; Jiang Ziying; Yang Hongwei; Cheng Weiya; Jiao Zhijuan; Liu Ruirui; Zhang Ailing

    2012-01-01

    Negative impacts on ecological environment of thermal discharge from nuclear power plants (namely thermal pollution) take increasing concern of the society. This paper analyzes current deficiency of study on controlling thermal pollution of nuclear power plants, and recommends to make further researches on aquatic ecological impact assessment methodology, improving deficiencies in numerical simulation and physical model test for the dispersion prediction of thermal discharge, and adjusting the method of remote sensing measurement for the offshore water temperature to adapt to thermal discharge monitoring. These would support for the effective control and monitoring of thermal pollution of nuclear power plants. (authors)

  20. Smart built-in test for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Lombrozo, P.C.

    1992-03-01

    Smart built-in test (BIT) technologies are envisioned for nuclear thermal propulsion spacecraft components which undergo constant irradiation and are therefore unsafe for manual testing. Smart BIT systems of automated/remote type allow component and system tests to be conducted; failure detections are directly followed by reconfiguration of the components affected. The 'smartness' of the BIT system in question involves the reduction of sensor counts via the use of multifunction sensors, the use of components as integral sensors, and the use of system design techniques which allow the verification of system function beyond component connectivity

  1. Scaling in nuclear reactor system thermal-hydraulics

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.

    2010-01-01

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  2. Scaling in nuclear reactor system thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F., E-mail: dauria@ing.unipi.i [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Galassi, G.M. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  3. Advanced modelling and numerical strategies in nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    Staedtke, H.

    2001-01-01

    The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

  4. Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

  5. Testing for Nuclear Thermal Propulsion Systems: Identification of Technologies for Effluent Treatment in Test Facilities

    Data.gov (United States)

    National Aeronautics and Space Administration — Key steps to ensure identification of relevant effluent treatment technologies for Nuclear Thermal Propulsion (NTP) testing include the following. 1. Review of...

  6. High Thrust & High ISP Nuclear Thermal Rocket (NTR) Grooved Ring Fuel Element (GRFE)

    Data.gov (United States)

    National Aeronautics and Space Administration — Missions to Mars will benefit from propulsion systems with performance levels exceeding that of today's best chemical engines. Nuclear Thermal Rocket (NTR)...

  7. Analytical study of nozzle performance for nuclear thermal rockets

    International Nuclear Information System (INIS)

    Davidian, K.O.; Kacynski, K.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

  8. Human Exploration Mission Capabilities to the Moon, Mars, and Near Earth Asteroids Using ''Bimodal'' NTR Propulsion

    International Nuclear Information System (INIS)

    Stanley K. Borowski; Leonard A. Dudzinski; Melissa L. McGuire

    2000-01-01

    The nuclear thermal rocket (NTR) is one of the leading propulsion options for future human exploration missions because of its high specific impulse (Isp ∼ 850 to 1000 s) and attractive engine thrust-to-weight ratio (∼ 3 to 10). Because only a minuscule amount of enriched 235 U fuel is consumed in an NRT during the primary propulsion maneuvers of a typical Mars mission, engines configured both for propulsive thrust and modest power generation (referred to as 'bimodal' operation) provide the basis for a robust, power-rich stage with efficient propulsive capture capability at the moon and near-earth asteroids (NEAs), where aerobraking cannot be utilized. A family of modular bimodal NTR (BNTR) space transfer vehicles utilize a common core stage powered by three ∼15-klb f engines that produce 50 kW(electric) of total electrical power for crew life support, high data rate communications with Earth, and an active refrigeration system for long-term, zero-boiloff liquid hydrogen (LH 2 ) storage. This paper describes details of BNTR engines and designs of vehicles using them for various missions

  9. Review of Nuclear Thermal Propulsion Ground Test Options

    Science.gov (United States)

    Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

    2015-01-01

    High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

  10. Potential impact of thermal effluents from Chongqing Fuling nuclear power plant to the Three Gorges Reservoir

    International Nuclear Information System (INIS)

    Han Baohua; Li Jianguo; Ma Binghui; Zhang Yue; Sun Qunli; Hu Yuping

    2012-01-01

    This study is based on the hydrological data near Chongqing Fuling Nuclear Power Plant along the Yangtze River, the present situation of the ecological environment of the Three Gorges Reservoir and the predicted results of thermal effluents from Chongqing Fuling Nuclear Power Plant. The standards of cooling water and the thermal tolerances indexes of aquatic organisms were investigated. The effects of thermal effluents on aquatic organisms were analyzed. The potential impact of Chongqing Fuling nuclear power plant to the Three Gorges Reservoir was explained. The results show that in the most adverse working conditions, the surface temperature near the outfall area is not more than 1℃, the temperature of thermal effluents do not exceed the suitable thermal range of fish breeding, growth and other thermal tolerances indexes. Thermal effluents from nuclear power plant have no influence about fish, plankton and benthic organisms in the Three Gorges Reservoir. (authors)

  11. Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Boyd, C.H.

    1989-01-01

    This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

  12. Space nuclear thermal propulsion test facilities accommodation at INEL

    Science.gov (United States)

    Hill, Thomas J.; Reed, William C.; Welland, Henry J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway.

  13. Space nuclear thermal propulsion test facilities accommodation at INEL

    International Nuclear Information System (INIS)

    Hill, T.J.; Reed, W.C.; Welland, H.J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway

  14. Contributions to thermal and fluid dynamic problems in nuclear technology

    International Nuclear Information System (INIS)

    Mueller, U.; Krebs, L.; Rust, K.

    1984-02-01

    The majority of contributions compiled in this report deals with thermal and fluid dynamic problems in nuclear engineering. Especially problems of heat transfer and cooling are represented which may arise during and afer a loss-of-coolant accident both in light water reactors and in liquid metal cooled fast breeder reactors. Papers on the mass transfer in pressurized water, tribological problems in sodium cooled reactors, the fluid dynamics of pulsed column, and fundamental investigations of convective flows supplement these contributions on problems connected with accidents. Furthermore, a keynote paper presents the individual activities relating to the reliability of reactor components, a field recently included in our research program. Technical solutions to special problems are closely connected to the investigations based on experiments. Therefore, several contributions deal with new developments in technology and measuring techniques. (orig.) [de

  15. Nuclear thermal rockets - Key to moon-Mars exploration

    International Nuclear Information System (INIS)

    Borowski, S.K.; Clark, J.S.; Mcilwain, M.C.; Pelaccio, D.G.

    1992-01-01

    The Space Exploration Initiative (SEI) calls for lunar and Martian exploration missions for which solid-core nuclear thermal rockets (NTRs), in virtue of their single-stage, fully-reusable nature, are ideally suited. NTRs promise double the specific impulse of chemical propulsion. A lunar mission employing a reusable NTR is currently being conducted by NASA. The NTR would be assembled in LEO in such a way that it remained 'radioactively cold' during earth-to-orbit deployment by a heavy-lift chemical booster, and therefore presented no radioactive hazard. Also under consideration is a particle-bed reactor in which the hydrogen propulsive fluid directly cools coated-particle fuel spheres

  16. Integrated System Modeling for Nuclear Thermal Propulsion (NTP)

    Science.gov (United States)

    Ryan, Stephen W.; Borowski, Stanley K.

    2014-01-01

    Nuclear thermal propulsion (NTP) has long been identified as a key enabling technology for space exploration beyond LEO. From Wernher Von Braun's early concepts for crewed missions to the Moon and Mars to the current Mars Design Reference Architecture (DRA) 5.0 and recent lunar and asteroid mission studies, the high thrust and specific impulse of NTP opens up possibilities such as reusability that are just not feasible with competing approaches. Although NTP technology was proven in the Rover / NERVA projects in the early days of the space program, an integrated spacecraft using NTP has never been developed. Such a spacecraft presents a challenging multidisciplinary systems integration problem. The disciplines that must come together include not only nuclear propulsion and power, but also thermal management, power, structures, orbital dynamics, etc. Some of this integration logic was incorporated into a vehicle sizing code developed at NASA's Glenn Research Center (GRC) in the early 1990s called MOMMA, and later into an Excel-based tool called SIZER. Recently, a team at GRC has developed an open source framework for solving Multidisciplinary Design, Analysis and Optimization (MDAO) problems called OpenMDAO. A modeling approach is presented that builds on previous work in NTP vehicle sizing and mission analysis by making use of the OpenMDAO framework to enable modular and reconfigurable representations of various NTP vehicle configurations and mission scenarios. This approach is currently applied to vehicle sizing, but is extensible to optimization of vehicle and mission designs. The key features of the code will be discussed and examples of NTP transfer vehicles and candidate missions will be presented.

  17. Center of thermal-physical data for nuclear power plants

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Blokhin, A.I.; Ivashkevich, A.A.; Katan, I.B.; Peskov, O.L.; Pan'kov, V.M.; Savanin, N.K.; Sal'nikova, O.V.; Khrushcheva, E.N.; Kirova, T.S.

    1982-01-01

    The specific features of a specialized Center of thermal-physical data (CTD) are considered. The center has been created for data acquisition, storage and analysis and working out recommendations on the following NPP thermal physics sections: hydrodynamics of channel flows (monophase laminar and turbulent, and two-phase flows, hydrodynamic vibrations) heat exchange in NPP elements, thermohydraulic calculations of nuclear reactor cores, heat exchangers, steam generators and NPP cooling system elements, coolant properties (water and steam, liquid metals and gases). On the CTD data base an automated system ASKhOD, oriented to EC computer, is created. The ASKhoD software ensures data allocation on magnetic tapes or other carriers, automated renewal and data relocation, data search in compliance with a specified set of signs, data processing for the purpose of their estimation or obtaining optimized model constants. Different publications in home and foreign magazines, conference, seminar materials, organization preprints serve as the data sources used for the formation of the ASKhOD data base

  18. Bimodal immune activation in psoriasis.

    Science.gov (United States)

    Christophers, E; Metzler, G; Röcken, M

    2014-01-01

    Psoriasis is an immune-regulated skin disease with various clinical subtypes and disease activities. The majority of patients present with predominantly stable plaques. At the onset of new lesions, plaque-type psoriasis frequently demonstrates pin-sized and highly inflammatory papules sometimes with an inflammatory border. The histopathology of initial psoriasis differs from stable plaque-type psoriasis. Early lesions demonstrate innate immune cells with neutrophils, degranulating mast cells and macrophages. These are followed by interleukin (IL)-1-dependent T helper (Th)17 cells, finally resulting in the Th1-dominated immunopathology of stable plaque-type psoriasis, where mononuclear cells predominate with interspersed neutrophilic (Munro) microabscesses. These features suggest a bimodal immune pathway where alternate activation of either innate (autoinflammatory) or adaptive (autoimmune) immunity predominates. Neutrophilic infiltrations appear during early psoriasis with Munro abscesses. They are time limited and occur periodically, clinically best seen in linear nail pitting. These features strongly suggest a critical role for an IL-1-Th17-dominated autoinflammation in the initiation of psoriasis, followed by a Th1-dominated late-phase reaction. The concept of bimodal immune activation helps to explain results from therapeutic interventions that are variable and previously only partly understood. © 2013 British Association of Dermatologists.

  19. Nuclear thermal rocket propulsion application to Mars missions

    International Nuclear Information System (INIS)

    Emrich, W.J. Jr.; Young, A.C.; Mulqueen, J.A.

    1991-01-01

    Options for vehicle configurations are reviewed in which nuclear thermal rocket (NTR) propulsion is used for a reference mission to Mars. The scenario assumes an opposition-class Mars transfer trajectory, a 435-day mission, and the use of a single nuclear engine with 75,000 lbs of thrust. Engine parameters are examined by calculating mission variables for a range of specific impulses and thrust/weight ratios. The reference mission is found to have optimal values of 925 s for the specific impulse and thrust/weight ratios of 4.0 and 0.06 for the engine and total stage ratios respectively. When the engine thrust/weight ratio is at least 4/1 the most critical engine parameter is engine specific impulse for reducing overall stage weight. In the context of this trans-Mars three-burn maneuver the NTR engine with an expander engine cycle is considered a more effective alternative than chemical/aerobrake and other propulsion options

  20. High-temperature turbopump assembly for space nuclear thermal propulsion

    Science.gov (United States)

    Overholt, David M.

    1993-01-01

    The development of a practical, high-performance nuclear rocket by the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program places high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio. The operating parameters arising from these goals drive the propellant-pump design. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is effected by rapid heating of the propellant from 100 K to thousands of degrees in the particle-bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. One approach to achieve high performance is to use an uncooled carbon-carbon nozzle and duct turbine inlet. The high-temperature capability is obtained by using carbon-carbon throughout the TPA hot section. Carbon-carbon components in development include structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines plus a wide variety of other turbomachinery applications.

  1. Carbon-carbon turbopump concept for Space Nuclear Thermal Propulsion

    Science.gov (United States)

    Overholt, David M.

    1993-06-01

    The U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program is placing high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio in the development of a practical high-performance nuclear rocket. The turbopump design is driven by these goals. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is from rapid heating of the propellant from 180 R to thousands of degrees in the particle bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. A high-performance approach is to use an uncooled carbon-carbon nozzle and duct turbine inlet. Carbon-carbon components are used throughout the TPA hot section to obtain the high-temperature capability. Several carbon-carbon components are in development including structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines and many other turbomachinery applications.

  2. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Beatriz M. dos; Alvim, Antonio C.M., E-mail: bmachado@nuclear.ufrj.br, E-mail: aalvim@gmail.com [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-11-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  3. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  4. Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

    Science.gov (United States)

    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

  5. Space nuclear reactor SP-100 thermal-hydraulic simulation

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Camillo, Giannino P.; Guimaraes, Lamartine N.F.

    2009-01-01

    Since 1983 it has been under development in the USA the project SP-100 of space nuclear reactors for electric generation in a range of 100 to 1000 KWe. In this project the heat is generated at the core of a fast compact liquid lithium refrigerated reactor. Thermoelectric converters produce direct current electric energy and the primary and secondary loops flow is controlled by electromagnetic thermoelectric pumps (EMTE). In this work it is studied a system with a fast nuclear reactor, with similar characteristics to the SP-100, aiming at generating high electric power in space for a future application on the TERRA (Advanced Fast Reactor Technology) Project of IEAv (Institute for Advanced Studies). It will be presented the working principles, basic structure and operation characteristics of an electromagnetic thermoelectric pump (EMTE) for a liquid metal cooled nuclear reactor refrigeration loops flow control. In order to determine the operating point of the reactor, it is indispensable the simulation of the EMTE pump along with the other components of the system, once all the working parameters are connected. So, it has been developed a computer system, named BEMTE-3 (a FORTRAN micro-computer code), which simulates the primary and secondary refrigeration components of liquid metal cooled fast space reactor. This computer code also simulates the thermoelectric conversion, with the flow being controlled by the EMTE pump with thermoelectric converters, determining the system operation point for a given nominal operating power. The BEMTE-3 is used for the study of the SP-100 primary and secondary loops thermal-hydraulic simulation and for the calculation of the operating point of the system based on data from available projects. (author)

  6. The important role of thermal hydraulics in 50 years of nuclear power applications

    International Nuclear Information System (INIS)

    Levy, Salomon

    1994-01-01

    Thermal hydraulics have played a very important role in the safety of nuclear power plants. The purpose of this paper is to provide a summary of our knowledge of thermal hydraulics in the 1950s and of the progress made up to the early 1990s. Important ''lessons learned'' over the past 50 years, and future potential issues in nuclear thermal hydraulics, are discussed. ((orig.))

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

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

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

  8. Nuclear reactor having thermally compensated support structure for a fuel assembly

    International Nuclear Information System (INIS)

    Borst, R.

    1980-01-01

    A thermal expansion compensation system for nuclear reactor fuel assemblies is disclosed which utilizes materials with different rates of thermal expansion in appropriate components so to: retain alignment of the assembly; reduce or eliminate thermal bow; and reduce or eliminate jump movement of fuel assemblies. (orig.)

  9. BSA adsorption on bimodal PEO brushes

    NARCIS (Netherlands)

    Bosker, WTE; Iakovlev, PA; Norde, W; Stuart, Martien A. Cohen

    2005-01-01

    BSA adsorption onto bimodal PEO brushes at a solid surface was measured using optical reflectometry. Bimodal brushes consist of long (N = 770) and short (N = 48) PEO chains and were prepared on PS surfaces, applying mixtures of PS29-PEO48 and PS37-PEO770 block copolymers and using the

  10. BSA adsorption on bimodal PEO brushes

    NARCIS (Netherlands)

    Bosker, W.T.E.; Iakovlev, P.A.; Norde, W.; Cohen Stuart, M.A.

    2005-01-01

    BSA adsorption onto bimodal PEO brushes at a solid surface was measured using optical reflectometry. Bimodal brushes consist of long (N=770) and short (N=48) PEO chains and were prepared on PS surfaces, applying mixtures of PS 29-PEO48 and PS37-PEO770 block copolymers and using the Langmuir-Blodgett

  11. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J., Jr.

    2014-01-01

    Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

  12. Revisiting Nuclear Thermal Propulsion for Human Mars Exploration

    Science.gov (United States)

    Percy, Thomas K.; Rodriguez, Mitchell

    2017-01-01

    Nuclear Thermal Propulsion (NTP) has long been considered as a viable in-space transportation alternative for delivering crew and cargo to the Martian system. While technology development work in nuclear propulsion has continued over the year, general interest in NTP propulsion applications has historically been tied directly to the ebb and flow of interest in sending humans to explore Mars. As far back as the 1960’s, plans for NTP-based human Mars exploration have been proposed and periodically revisited having most recently been considered as part of NASA Design Reference Architecture (DRA) 5.0. NASA has been investigating human Mars exploration strategies tied to its current Journey to Mars for the past few years however, NTP has only recently been added into the set of alternatives under consideration for in-space propulsion under the Mars Study Capability (MSC) team, formerly the Evolvable Mars Campaign (EMC) team. The original charter of the EMC was to find viable human Mars exploration approaches that relied heavily on technology investment work already underway, specifically related to the development of large Solar Electric Propulsion (SEP) systems. The EMC team baselined several departures from traditional Mars exploration ground rules to enable these types of architectures. These ground rule changes included lower energy conjunction class trajectories with corresponding longer flight times, aggregation of mission elements in cis-Lunar space rather than Low Earth Orbit (LEO) and, in some cases, the pre-deployment of Earth return propulsion systems to Mars. As the MSC team continues to refine the in-space transportation trades, an NTP-based architecture that takes advantage of some of these ground rule departures is being introduced.

  13. Performance assessment of low pressure nuclear thermal propulsion

    Science.gov (United States)

    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.

  14. Social assessment and location of nuclear and thermal power plants

    International Nuclear Information System (INIS)

    Nemoto, Kazuyasu; Nishio, Mitsuo.

    1979-01-01

    Most of the locations of nuclear and thermal power plants in Japan are depopulated villages with remote rural character, but for the development of such districts, the policy is not yet clearly established, and the appropriate measures are not taken. The living regions of residents and the production regions of enterprises are more and more estranged. Social assessment is the scientific method to perceive the future change due to the installation of power stations. The features particular to the assessment of natural environment and social environment related to the location of power stations are considered, and the technical problems involved in the method of assessment of natural environment are solved, and the actual method of assessment of social environment is developed. Then, the possibility of establishing this method and the problems in its application are investigated. The plan of developing the surroundings of power generation facilities is criticized, and the coordination of the location plan of power companies and the regional projects of municipalities is discussed. Finally, the mechanism of consensus formation concerning the location of power stations is considered, dividing into regional consensus formation and administrative consensus formation, and the possibility of instituting social assessment is examined. (Kako, I.)

  15. Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

    International Nuclear Information System (INIS)

    Ninokata, Hisashi

    2012-01-01

    The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

  16. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    Science.gov (United States)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  17. Fabrication and Testing of Nuclear-Thermal Propulsion Ground Test Hardware, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Efficient nuclear-thermal propulsion requires heating a low molecular weight gas, typically hydrogen, to high temperature and expelling it through a nozzle. The...

  18. Fabrication and Testing of Nuclear-Thermal Propulsion Ground Test Hardware, Phase II, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Efficient nuclear-thermal propulsion (NTP) requires heating a low molecular weight gas, typically hydrogen, to high temperature and expelling it through a nozzle....

  19. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as an approach that can provide the fastest trip times to Mars and as the preferred concept for human space...

  20. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as a propulsion concept which could provide the fastest trip times to Mars and as the preferred concept for...

  1. Improved CVD Coatings for Carbide Based Nuclear Thermal Propulsion Fuel Elements, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — One of the great hurdles to further development and evaluation of nuclear thermal propulsion systems is the issue surrounding the release of radioactive material...

  2. Extreme Temperature Radiation Tolerant Instrumentation for Nuclear Thermal Propulsion Engines, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of this proposal is to develop and commercialize a high reliability, high temperature smart neutron flux sensor for NASA Nuclear Thermal Propulsion...

  3. Fabrication and Testing of Nuclear-Thermal Propulsion Ground Test Hardware Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Efficient nuclear-thermal propulsion requires heating a low molecular weight gas, typically hydrogen, to high temperature and expelling it through a nozzle. The...

  4. NEPTUNE: A new software platform for advanced nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Herard, J.M.; Peturaud, P.; Bestion, D.; Boudier, P.; Hervieu, E.; Fillion, P.; Grandotto, M.

    2007-01-01

    The NEPTUNE project constitutes the thermal-hydraulic part of the long-term Electricite de France and Commissariat a l'Energie Atomique joint research and development program for the next generation of nuclear reactor simulation tools. This program is also financially supported by the Institut de Radioprotection et Surete Nucleaire and AREVA NP. The project aims at developing a new software platform for advanced two-phase flow thermal hydraulics covering the whole range of modeling scales and allowing easy multi-scale and multidisciplinary calculations. NEPTUNE is a fully integrated project that covers the following fields: software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques, and performance of new experimental programs. The analysis of the industrial needs points out that three main simulation scales are involved. The system scale is dedicated to the overall description of the reactor. The component or subchannel scale allows three-dimensional computations of the main components of the reactors: cores, steam generators, condensers, and heat exchangers. The current generation of system and component codes has reached a very high level of maturity for industrial applications. The third scale, computational fluid dynamics (CFD) in open medium, allows one to go beyond the limits of the component scale for a finer description of the flows. This scale opens promising perspectives for industrial simulations, and the development and validation of the NEPTUNE CFD module have been a priority since the beginning of the project. It is based on advanced physical models (two-fluid or multi field model combined with interfacial area transport and two-phase turbulence) and modern numerical methods (fully unstructured finite volume solvers). For the system and component scales, prototype developments have also started, including new physical models and numerical methods. In addition to scale

  5. Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

    Science.gov (United States)

    Kirk, Daniel R.

    2005-01-01

    When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

  6. Irreducible complexity of iterated symmetric bimodal maps

    Directory of Open Access Journals (Sweden)

    J. P. Lampreia

    2005-01-01

    Full Text Available We introduce a tree structure for the iterates of symmetric bimodal maps and identify a subset which we prove to be isomorphic to the family of unimodal maps. This subset is used as a second factor for a ∗-product that we define in the space of bimodal kneading sequences. Finally, we give some properties for this product and study the ∗-product induced on the associated Markov shifts.

  7. Lattice Thermal Conductivity of Polyethylene Molecular Crystals from First-Principles Including Nuclear Quantum Effects.

    Science.gov (United States)

    Shulumba, Nina; Hellman, Olle; Minnich, Austin J

    2017-11-03

    Molecular crystals such as polyethylene are of intense interest as flexible thermal conductors, yet their intrinsic upper limits of thermal conductivity remain unknown. Here, we report a study of the vibrational properties and lattice thermal conductivity of a polyethylene molecular crystal using an ab initio approach that rigorously incorporates nuclear quantum motion and finite temperature effects. We obtain a thermal conductivity along the chain direction of around 160  W m^{-1} K^{-1} at room temperature, providing a firm upper bound for the thermal conductivity of this molecular crystal. Furthermore, we show that the inclusion of quantum nuclear effects significantly impacts the thermal conductivity by altering the phase space for three-phonon scattering. Our computational approach paves the way for ab initio studies and computational material discovery of molecular solids free of any adjustable parameters.

  8. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    Science.gov (United States)

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

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  9. Simulation of Thermal, Neutronic and Radiation Characteristics in Spent Nuclear Fuel and Radwaste Facilities

    International Nuclear Information System (INIS)

    Poskas, P.; Bartkus, G.

    1999-01-01

    The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)

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

    International Nuclear Information System (INIS)

    Clark, J.S.; Mcdaniel, P.; Howe, S.; Helms, I.; Stanley, M.

    1993-04-01

    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

  11. Computational Efficient Upscaling Methodology for Predicting Thermal Conductivity of Nuclear Waste forms

    International Nuclear Information System (INIS)

    Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

    2011-01-01

    This study evaluated different upscaling methods to predict thermal conductivity in loaded nuclear waste form, a heterogeneous material system. The efficiency and accuracy of these methods were compared. Thermal conductivity in loaded nuclear waste form is an important property specific to scientific researchers, in waste form Integrated performance and safety code (IPSC). The effective thermal conductivity obtained from microstructure information and local thermal conductivity of different components is critical in predicting the life and performance of waste form during storage. How the heat generated during storage is directly related to thermal conductivity, which in turn determining the mechanical deformation behavior, corrosion resistance and aging performance. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling models were developed and implemented. Due to the absence of experimental data, prediction results from finite element method (FEM) were used as reference to determine the accuracy of different upscaling models. Micrographs from different loading of nuclear waste were used in the prediction of thermal conductivity. Prediction results demonstrated that in term of efficiency, boundary models (Taylor and Sachs model) are better than self consistent model, statistical upscaling method and FEM. Balancing the computation resource and accuracy, statistical upscaling is a computational efficient method in predicting effective thermal conductivity for nuclear waste form.

  12. Study on relationship between aging and thermal-hydraulic behaviors of nuclear power plants

    International Nuclear Information System (INIS)

    Murata, Hiroyuki; Inasaka, Fujio; Adachi, Masaki; Sawada, Ken-ichi; Akiyama, Shigeru; Sakuma, Masaaki; Takahashi, Ichihiko; Ushijima, Michio

    2005-01-01

    The number of aged nuclear power plants will increase in the future, because operation periods of the existing nuclear power plants are being extended from thirty years of initial supposition to sixty years at the longest. Therefore, it is important to establish the methodology to guarantee integrity of the aged nuclear power plants. Among the reported damages of nuclear power plant components due to the fatigue during long terms, many cases are considered to be related to their thermal-hydraulic behaviors during operation. Thus, quantitative understanding of thermal-hydraulic behaviors in the nuclear power plants is important to estimate many kinds of aging processes accurately. The research project, 'study on relationship between nuclear power plant aging and its thermal-hydraulic behaviors', was conducted from 2001 to 2004 in order to clarify effects of thermal-hydraulic behaviors in the nuclear power plants on structural materials during aging processes. In this project, flow induced vibration of an array of circular cylinders was investigated experimentally and numerically. Rotating bending fatigue tests were also performed for the austenitic stainless steel SUS316L (JIS G4304) and Ni-Cr-Fe alloy NCF690 (JIS G4904, INCONEL alloy 690 equivalent material) in order to examine the fatigue strength in the ultra high cycle fatigue region, namely 10 7 -10 9 cycles, and the notch effects. (author)

  13. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  14. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Ahn, S.H.; Aksan, N.; Austregesilo, H.; Bestion, D.; Chung, B.D.; D’Auria, F.; Emonot, P.; Gandrille, J.L.; Hanninen, M.; Horvatović, I.; Kim, K.D.; Kovtonyuk, A.; Petruzzi, A.

    2015-01-01

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes

  15. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)

    2015-01-15

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

  16. Design of a bolted flange subjected to severe nuclear system thermal transients - A case study

    International Nuclear Information System (INIS)

    Palmer, W.J.; Tomawski, R.J.; Ezekoye, L.I.; Lacey, M.L.

    1986-01-01

    Flange design standards recognize that flanged joints may develop leakage should they be exposed to severe thermal gradients and recommend that such operating conditions be avoided. In nuclear power plants, severe thermal transients may be encountered in many plant and system operating and test conditions. In such applications, conformance with standard design practice may not ensure a leak-tight joint. This paper describes the proper consideration of thermal effects on flanged joints and how that can lead to the development of a successful leak-tight design. Similar procedures may be applied generally to evaluate and upgrade flanged joints in thermal shock applications

  17. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  18. High Temperature Resistance Claddings for Nuclear Thermal Rockets, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This program will develop a series of nano-/micro-composite coated nuclear reactor facing components using MesoCoat's CermaCladTM process. This proposed SBIR program...

  19. A coupled nuclear reactor thermal energy storage system for enhanced load following operation

    International Nuclear Information System (INIS)

    Alameri, Saeed A.; King, Jeffrey C.

    2013-01-01

    Nuclear power plants operate most economically at a constant power level, providing base load electric power. In an energy grid containing a high fraction of renewable power sources, nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling a nuclear reactor to a large thermal energy storage block will allow the reactor to better respond to variable power demands. In the system described in this paper, a Prismatic core Advanced High Temperature Reactor supplies constant power to a lithium chloride molten salt thermal energy storage block that provides thermal power as needed to a closed Brayton cycle energy conversion system. During normal operation, the thermal energy storage block stores thermal energy during the night for use in the times of peak demand during the day. In this case, the nuclear reactor stays at a constant thermal power level. After a loss of forced circulation, the reactor reaches a shut down state in less than half an hour and the average fuel, graphite and coolant temperatures remain well within the design limits over the duration of the transient, demonstrating the inherent safety of the coupled system. (author)

  20. Beta-binomial regression and bimodal utilization.

    Science.gov (United States)

    Liu, Chuan-Fen; Burgess, James F; Manning, Willard G; Maciejewski, Matthew L

    2013-10-01

    To illustrate how the analysis of bimodal U-shaped distributed utilization can be modeled with beta-binomial regression, which is rarely used in health services research. Veterans Affairs (VA) administrative data and Medicare claims in 2001-2004 for 11,123 Medicare-eligible VA primary care users in 2000. We compared means and distributions of VA reliance (the proportion of all VA/Medicare primary care visits occurring in VA) predicted from beta-binomial, binomial, and ordinary least-squares (OLS) models. Beta-binomial model fits the bimodal distribution of VA reliance better than binomial and OLS models due to the nondependence on normality and the greater flexibility in shape parameters. Increased awareness of beta-binomial regression may help analysts apply appropriate methods to outcomes with bimodal or U-shaped distributions. © Health Research and Educational Trust.

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

    Science.gov (United States)

    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.

  2. Research on technology of evaluating thermal property data of nuclear power materials

    International Nuclear Information System (INIS)

    Imai, Hidetaka; Baba, Tetsuya; Matsumoto, Tsuyoshi; Kishimoto, Isao; Taketoshi, Naoyuki; Arai, Teruo

    1997-01-01

    For the materials of first wall and diverter of nuclear fusion reactor, in order to withstand steady and unsteady high heat flux load, excellent thermal characteristics are required. It is strongly demanded to measure such thermal property values as heat conductivity, heat diffusivity, specific heat capacity, emissivity and so using small test pieces up to higher than 2000degC. As the materials of nuclear reactors are subjected to neutron irradiation, in order to secure the long term reliability of the materials, it is very important to establish the techniques for forecasting the change of the thermal property values due to irradiation effect. Also the establishment of the techniques for estimating the thermal property values of new materials like low radioactivation material is important. In National Research Laboratory of Metrology, the research on the advancement of the measuring technology for high temperature thermal properties has resulted in the considerably successful development of such technologies. In this research, the rapid measurement of thermal property values up to superhigh temperature with highest accuracy, the making of thermal property data set of high level, the analysis and evaluation of the correlation of material characters and thermal property values, and the development of the basic techniques for estimating the thermal property values of solid materials are aimed at and advanced. These are explained. (K.I.)

  3. Language choice in bimodal bilingual development

    Directory of Open Access Journals (Sweden)

    Diane eLillo-Martin

    2014-10-01

    Full Text Available Bilingual children develop sensitivity to the language used by their interlocutors at an early age, reflected in differential use of each language by the child depending on their interlocutor. Factors such as discourse context and relative language dominance in the community may mediate the degree of language differentiation in preschool age children.Bimodal bilingual children, acquiring both a sign language and a spoken language, have an even more complex situation. Their Deaf parents vary considerably in access to the spoken language. Furthermore, in addition to code-mixing and code-switching, they use code-blending – expressions in both speech and sign simultaneously – an option uniquely available to bimodal bilinguals. Code-blending is analogous to code-switching sociolinguistically, but is also a way to communicate without suppressing one language. For adult bimodal bilinguals, complete suppression of the non-selected language is cognitively demanding. We expect that bimodal bilingual children also find suppression difficult, and use blending rather than suppression in some contexts. We also expect relative community language dominance to be a factor in children’s language choices.This study analyzes longitudinal spontaneous production data from four bimodal bilingual children and their Deaf and hearing interlocutors. Even at the earliest observations, the children produced more signed utterances with Deaf interlocutors and more speech with hearing interlocutors. However, while three of the four children produced >75% speech alone in speech target sessions, they produced <25% sign alone in sign target sessions. All four produced bimodal utterances in both, but more frequently in the sign sessions, potentially because they find suppression of the dominant language more difficult.Our results indicate that these children are sensitive to the language used by their interlocutors, while showing considerable influence from the dominant

  4. Bimodal condensation silicone elastomers as dielectric elastomers

    DEFF Research Database (Denmark)

    Yu, Liyun; Madsen, Frederikke Bahrt; Skov, Anne Ladegaard

    unimodal refers to that there is one polymer only in the system. As an alternative to unimodal networks there are the bimodal networks where two polymers with significantly different molecular weights are mixed with one crosslinker. [2]Silicone rubber can be divided into condensation type and addition type...... according to the curing reaction. The advantages of condensation silicones compared to addition are the relatively low cost, the curing rate largely being independent of temperature, the excellent adhesion, and the catalyst being nontoxic. [3]In this work, a series of bimodal condensation silicone...

  5. Origin of low thermal conductivity in nuclear fuels.

    Science.gov (United States)

    Yin, Quan; Savrasov, Sergey Y

    2008-06-06

    Using a novel many-body approach, we report lattice dynamical properties of UO2 and PuO2 and uncover various contributions to their thermal conductivities. Via calculated Grüneisen constants, we show that only longitudinal acoustic modes having large phonon group velocities are efficient heat carriers. Despite the fact that some optical modes also show their velocities which are extremely large, they do not participate in the heat transfer due to their unusual anharmonicity. Ways to improve thermal conductivity in these materials are discussed.

  6. A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

    Science.gov (United States)

    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.

  7. Thermal-CFD Analysis of Combined Solar-Nuclear Cycle Systems.

    Energy Technology Data Exchange (ETDEWEB)

    Fathi, Nima [Univ. of New Mexico, Albuquerque, NM (United States); McDaniel, Patrick [Univ. of New Mexico, Albuquerque, NM (United States); Vorobieff, Peter [Univ. of New Mexico, Albuquerque, NM (United States); de Oliveira, Cassiano [Univ. of New Mexico, Albuquerque, NM (United States); Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Aleyasin, Seyed Sobhan [Univ. of Manitoba (Canada)

    2015-09-01

    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 typical 1000 MW nuclear power plant with a 0.33 thermal efficiency, we can increase it to 0.39.

  8. A bimodal power and propulsion system based on cermet fuel and heat pipe energy transport

    International Nuclear Information System (INIS)

    Polansky, G.F.; Gunther, N.A.; Rochow, R.F.; Bixler, C.H.

    1995-01-01

    Bimodal space reactor systems provide both thermal propulsion for the spacecraft orbital transfer and electrical power to the spacecraft bus once it is on station. These systems have the potential to increase both the available payload in high energy orbits and the available power to that payload. These increased mass and power capabilities can be used to either reduce mission cost by permitting the use of smaller launch vehicles or to provide increased mission performance from the current launch vehicle. A major barrier to the deployment of these bimodal systems has been the cost associated with their development. This paper describes a bimodal reactor system with performance potential to permit more than 70% of the instrumented payload of the Titan IV/Centaur to be launched from the Atlas IIAS. The development cost is minimized by basing the design on existing component technologies

  9. A study on the ocean circulation and thermal diffusion near a nuclear power plant

    International Nuclear Information System (INIS)

    Shu, Kyung Suk; Han, Moon Hee; Kim, Eun Han; Hwang, Won Tae

    1994-08-01

    The thermal discharge used with cooling water at nuclear power plant is released to a neighbour sea and it is influenced on marine environment. The thermal discharge released from power plant is mainly transported and diffused by ocean circulation of neighbour sea. So the evaluation for characteristics of ocean circulation around neighbour sea is firstly performed. The purpose of this research is primarily analyzed the thermal diffusion in sea around Yongkwang nuclear power plant. For this viewpoint, fundamental oceanographic data sets are collected and analyzed in Yellow sea, west sea of Korea, sea around Yongkwang. The ocean circulation and the effects of temperature increase by thermal discharge are evaluated using these data. The characteristics of tide is interpreted by the analysis of observed tidal elevation and tidal currents. The characteristics of temperature and salinity is investigated by the long-term observation of Korea Fisheries Research and Development Agency and the short-term observation around Yongkwang. (Author)

  10. Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

    The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

  11. Microscopic simulation of nuclear reactor thermal-hydraulics

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Aoki, Takayuki; Muramatsu, Toshiharu

    2000-01-01

    On three-dimensional behavior, multi-phase and multi-component thermal-hydraulics co-existing complex phase change and chemical reaction seen at severe accident, and so on under complex systems such as liquid film and drop behavior in a fuel assembly, inside of reactor pressure vessel, and so on, on aiming at further precise analysis on thermal-hydraulics phenomena in a reactor, researches on simulation focusing at microscopic mechanism of the phenomena have been carried out. And, simulations on microscopic thermal-hydraulics phenomena using numerical analysis such as super fine mesh method based on continuous body dynamic procedure, particle method, lattice Boltzmann method, Automaton method, Monte Carlo method at a view point of molecular dynamics, and so on have also become to be found out. This report was summary of previous survey actions under the research special committee on the 'microscopic simulation on reactor thermal-hydraulics' of the Japanese Society of Atomic Energy started in 1996. Here were elucidated a relationship between previous applicable technologies such as macroscopic simulation focusing a motion at local mean field and microscopic simulation one as well as survey and systematization of present condition of microscopic understanding on various physical phenomena, to scope directivity and subject on future research. (G.K.)

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

    Science.gov (United States)

    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.

  13. Thermal-work strain in law enforcement personnel during chemical, biological, radiological, and nuclear (CBRN) training

    OpenAIRE

    Yokota, M; Karis, A J; Tharion, W J

    2014-01-01

    Background: Thermal safety standards for the use of chemical, biological, radiological, and nuclear (CBRN) ensembles have been established for various US occupations, but not for law enforcement personnel. Objectives: We examined thermal strain levels of 30 male US law enforcement personnel who participated in CBRN field training in Arizona, Florida, and Massachusetts. Methods: Physiological responses were examined using unobtrusive heart rate (HR) monitors and a simple thermoregulatory model...

  14. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  15. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications

  16. Refining Bimodal Microstructure of Materials with MSTRUCT

    Czech Academy of Sciences Publication Activity Database

    Matěj, Z.; Kadlecová, A.; Janeček, M.; Matějová, Lenka; Dopita, M.; Kužel, R.

    2014-01-01

    Roč. 29, S2 (2014), S35-S41 ISSN 0885-7156 R&D Projects: GA ČR GA14-23274S Grant - others:UK(CZ) UNCE 204023/2012 Institutional support: RVO:67985858 Keywords : XRD * bimodal * crystallite size Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.636, year: 2014

  17. Deaf Children's Bimodal Bilingualism and Education

    Science.gov (United States)

    Swanwick, Ruth

    2016-01-01

    This paper provides an overview of the research into deaf children's bilingualism and bilingual education through a synthesis of studies published over the last 15 years. This review brings together the linguistic and pedagogical work on bimodal bilingualism to inform educational practice. The first section of the review provides a synthesis of…

  18. Nonlinear dynamics of the bimodal optical computer

    Science.gov (United States)

    Caulfield, H. John

    1999-03-01

    In the bimodal optical computer, linear and nonlinear acts occur in rapid succession generating solutions to Ax equals b. Both chaos and stochastic resonance can appear in some cases. This is the first observation of such complexity effects in optical processors.

  19. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  20. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

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

    Science.gov (United States)

    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

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

    Science.gov (United States)

    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.

  3. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  4. Nuclear Thermal Propulsion: Past, Present, and a Look Ahead

    Science.gov (United States)

    Borowski, Stanley K.

    2014-01-01

    NTR: High thrust high specific impulse (2 x LOXLH2 chemical) engine uses high power density fission reactor with enriched uranium fuel as thermal power source. Reactor heat is removed using H2 propellant which is then exhausted to produce thrust. Conventional chemical engine LH2 tanks, turbo pumps, regenerative nozzles and radiation-cooled shirt extensions used -- NTR is next evolutionary step in high performance liquid rocket engines.

  5. Thermal hydraulic characteristics of a double-walled tube advanced nuclear steam generator

    International Nuclear Information System (INIS)

    Cho, S.M.; Seltzer, A.H.

    1989-01-01

    In this paper the thermal hydraulic characteristics of double-walled tube steam generator designed for sodium-cooled nuclear reactors are presented. The double-walled tube construction, along with double-barrier welds for tube-to-tubesheet joints, virtually eliminates the probability of heat transfer tube failure. Considerations are given to the use of the internal core tube, helical vane swirl generator, external protector tube, and variably perforated flow baffles to improve thermal and hydraulic performance of the steam generator. These thermal hydraulic design features with a particular reference to a 432 MW PRISM steam generator are discussed

  6. Arc-Heater Facility for Hot Hydrogen Exposure of Nuclear Thermal Rocket Materials

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Wang,Ten-See; Hickman, Robert; Panda, Binayak; Dobson, Chris; Osborne, Robin; Clifton, Scooter

    2006-01-01

    A hyper-thermal environment simulator is described for hot hydrogen exposure of nuclear thermal rocket material specimens and component development. This newly established testing capability uses a high-power, multi-gas, segmented arc-heater to produce high-temperature pressurized hydrogen flows representative of practical reactor core environments and is intended to serve. as a low cost test facility for the purpose of investigating and characterizing candidate fueUstructura1 materials and improving associated processing/fabrication techniques. Design and development efforts are thoroughly summarized, including thermal hydraulics analysis and simulation results, and facility operating characteristics are reported, as determined from a series of baseline performance mapping tests.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler

    2012-01-01

    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 selected 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

  8. Thermal stratification in a scaled-down suppression pool of the Fukushima Daiichi nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Byeongnam, E-mail: jo@vis.t.u-tokyo.ac.jp [Nuclear Professional School, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan); Erkan, Nejdet [Nuclear Professional School, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan); Takahashi, Shinji [Department of Nuclear Engineering and Management, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Song, Daehun [Nuclear Professional School, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan); Hyundai and Kia Corporate R& D Division, Hyundai Motors, 772-1, Jangduk-dong, Hwaseong-Si, Gyeonggi-Do 445-706 (Korea, Republic of); Sagawa, Wataru; Okamoto, Koji [Nuclear Professional School, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan)

    2016-08-15

    Highlights: • Thermal stratification was reproduced in a scaled-down suppression pool of the Fukushima Daiichi nuclear power plants. • Horizontal temperature profiles were uniform in the toroidal suppression pool. • Subcooling-steam flow rate map of thermal stratification was obtained. • Steam bubble-induced flow model in suppression pool was suggested. • Bubble frequency strongly depends on the steam flow rate. - Abstract: Thermal stratification in the suppression pool of the Fukushima Daiichi nuclear power plants was experimentally investigated in sub-atmospheric pressure conditions using a 1/20 scale torus shaped setup. The thermal stratification was reproduced in the scaled-down suppression pool and the effect of the steam flow rate on different thermal stratification behaviors was examined for a wide range of steam flow rates. A sparger-type steam injection pipe that emulated Fukushima Daiichi Unit 3 (F1U3) was used. The steam was injected horizontally through 132 holes. The development (formation and disappearance) of thermal stratification was significantly affected by the steam flow rate. Interestingly, the thermal stratification in the suppression pool vanished when subcooling became lower than approximately 5 °C. This occurred because steam bubbles are not well condensed at low subcooling temperatures; therefore, those bubbles generate significant upward momentum, leading to mixing of the water in the suppression pool.

  9. RECENT ACTIVITIES AT THE CENTER FOR SPACE NUCLEAR RESEARCH FOR DEVELOPING NUCLEAR THERMAL ROCKETS

    Energy Technology Data Exchange (ETDEWEB)

    Robert C. O' Brien

    2001-09-01

    Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent investigations into alternatives in design and materials for future systems will be discussed.

  10. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-01-01

    The age of nuclear waste - the length of time between its removal from the reactor cores and its emplacement in a repository - is a significant factor in determining the thermal loading of a repository. The surface cooling period as well as the density and sequence of waste emplacement affects both the near-field repository structure and the far-field geologic environment. To investigate these issues, a comprehensive review was made of the available literature pertaining to thermal effects and thermal properties of mined geologic repositories. This included a careful evaluation of the effects of different surface cooling periods of the wastes, which is important for understanding the optimal thermal loading of a repository. The results led to a clearer understanding of the importance of surface cooling in evaluating the overall thermal effects of a radioactive waste repository. The principal findings from these investigations are summarized in this paper

  11. Preparation of processed nuclear data libraries for thermal, fast and fusion research and power reactor applications

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

    A Consultants Meeting on ''Preparation of Processed Nuclear Data Libraries for Thermal, Fast and Fusion Research and Power Reactor Applications'' was convened by the International Atomic Energy Agency and held during December 13-16, 1993 December 8-10, 1993 at the IAEA Headquarters, Vienna. The detailed agenda, the complete list of participants and the recommendations are presented in this report. (author)

  12. Design data and safety features of commercial nuclear power plants in the United States: thermal hydraulic

    International Nuclear Information System (INIS)

    Kuriyama, Minoru; Morishima, Atsuyoshi

    1975-02-01

    The thermal-hydraulic data of commercial nuclear power plants in the United States are presented for 85 PWRs and 52 BWRs. Covered are the temperature, pressure, heat flux, flow rate, etc. of coolant and/or fuel cladding; the diffinitions are also given for some of these items. (auth.)

  13. Trend analysis of troubles caused by thermal-hydraulic phenomena at nuclear power plants

    International Nuclear Information System (INIS)

    Komatsu, Teruo

    2010-01-01

    The Institute of Nuclear Safety System (INSS) is promoting researches to improve the safety and reliability of nuclear power plants. In the present study, our attention was focused on troubles attributed to thermal-hydraulic phenomena in particular, trend analysis were carried out to learn lessons from these troubles and to prevent their recurrence. Through our survey, we found the following two points. First, many thermal-hydraulics related troubles can be attributed to design faults, since we found some events in foreign countries took place after inadequate facility renovation. To ensure appropriate design verification, it is important to take account of state-of-the-art science and technology and at the same time to pay attention to the compatibility with the initial design concept. Second point, thermal-hydraulic related troubles are common and recurrent to nuclear power plants worldwide. Japanese utilities are planning to introduce some of overseas experiences to their plants, such as power uprate and renovations of aged facilities. It is important to learn lessons from experiences paying close attention continuously to overseas trouble events, including thermal-hydraulics related events, and to use them to improve safety and reliability of nuclear power plants. (author)

  14. Survey of thermal-hydraulic models of commercial nuclear power plants

    International Nuclear Information System (INIS)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described

  15. Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

    International Nuclear Information System (INIS)

    Xie Hui; Zhou Jie; He Yingchao

    1993-01-01

    Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

  16. Nanotechnology for advanced nuclear thermal-hydraulics and safety: boiling and condensation

    International Nuclear Information System (INIS)

    Bang, In Cheol; Jeong, Ji Hwan

    2011-01-01

    A variety of Generation III/III+ water-cooled reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world in efforts to solve the future energy supply shortfall. Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. Phase change by boiling and condensation in the reverse process is a highly efficient heat transport mechanism that accommodates large heat fluxes with relatively small driving temperature differences. This mode of heat transfer is encountered in a wide spectrum of nuclear systems,and thus it is necessary to determine the thermal limit of water-cooled nuclear energy conversion in terms of economic and safety. Such applications are being advanced with the introduction of new technologies such as nanotechnology. Here, we investigated newly-introduced nanotechnologies relevant to boiling and condensation in general engineering applications. We also evaluated the potential linkage between such new advancements and nuclear applications in terms of advanced nuclear thermal-hydraulics

  17. Thermal Bremsstrahlung probing nuclear multifragmentation in nucleus-nucleus collisions around the Fermi energy

    International Nuclear Information System (INIS)

    D'Enterria, D.G.

    2000-05-01

    The thermodynamical properties of nuclear matter at moderate temperatures and densities, in the vicinity of the predicted nuclear liquid-gas phase transition, are studied using as experimental probe the hard-photons (E γ > 30 MeV) emitted in nucleus-nucleus collisions. Photon and charged-particle production in four different heavy-ion reactions (Ar 36 + Au 197 , Ag 107 , Ni 58 , C 12 at 60 A*MeV) is measured exclusively and inclusively coupling the TAPS photon spectrometer with two charged-particle and intermediate-mass-fragment detectors covering nearly 4π. We confirm that Bremsstrahlung emission in first-chance (off-equilibrium) proton-neutron collisions (pnγ) is the dominant origin of hard photons. We also firmly establish the existence of a thermal radiation component emitted in second-chance proton-neutron collisions. This thermal Bremsstrahlung emission takes place in semi-central and central nucleus-nucleus reactions involving heavy targets. We exploit this observation i) to demonstrate that thermal equilibrium is reached during the reaction, ii) to establish a new thermometer of nuclear matter based on Bremsstrahlung photons, iii) to derive the thermodynamical properties of the excited nuclear sources and, in particular, to establish a 'caloric curve' (temperature versus excitation energy), and iv) to assess the time-scales of the nuclear break-up process. (author)

  18. Nuclear energy generation and waste transmutation using an accelerator-driven intense thermal neutron source

    International Nuclear Information System (INIS)

    Bowman, C.D.; Arthur, E.D.; Lisowski, P.W.; Lawrence, G.P.; Jensen, R.J.; Anderson, J.L.; Blind, B.; Cappiello, M.; Davidson, J.W.; England, T.R.; Engel, L.N.; Haight, R.C.; Hughes, H.G. III; Ireland, J.R.; Krakowski, R.A.; LaBauve, R.J.; Letellier, B.C.; Perry, R.T.; Russell, G.J.; Staudhammer, K.P.; Versamis, G.; Wilson, W.B.

    1992-01-01

    We describe a new approach for commercial nuclear energy production without a long-term high-level waste stream and for transmutation of both fission product and higher actinide commercial nuclear waste using a thermal flux of accelerator-produced neutrons in the 10 16 n/cm 2 s range. Continuous neutron fluxes at this intensity, which is approximately 100 times larger than is typically available in a large scale thermal reactor, appear practical, owing to recent advances in proton linear accelerator technology and to the spallation target-moderator design presented here. This large flux of thermal neutrons makes possible a waste inventory in the transmutation system which is smaller by about a factor of 100 than competing concepts. The accelerator allows the system to operate well below criticality so that the possibility for a criticality accident is eliminated. No control rods are required. The successful implementation of this new method for energy generation and waste transmutation would eliminate the need for nuclear waste storage on a geologic time scale. The production of nuclear energy from 232 Th or 238 U is used to illustrate the general principles of commercial nuclear energy, production without long-term high-level waste. There appears to be sufficient thorium to meet the world's energy needs for many millenia. (orig.)

  19. Recent advances in nuclear fuels technology for thermal reactors

    International Nuclear Information System (INIS)

    Sutharshan, Balendra

    2011-01-01

    In today's competitive electrical generation, many nuclear power generators are lowering operating and fuel cycle costs by extending burnups, utilizing longer cycles, reducing outage duration, increasing peaking factors for more efficient fuel management; and by up rating to maximize energy output from the reactors. To better equip nuclear operators to meet these competitive challenges, Westinghouse has strategically aligned its goals to ensure that customer needs are met and that fuel supplied operates flawlessly. Westinghouse's fuel performance program implements design features and manufacturing processes to maximize margins to failure, specify bounds of reactor operation, and monitor critical operating parameters using BEACON software as well as specify and implement a robust Post Irradiation Examination (PIE) to obtain early feedback on fuel performance. Westinghouse's unwavering commitment to achieve flawless fuel performance and to innovate resulted in exceptional pressurized water reactor (PWR), boiling water reactor (BWR), and VVER fuel performance worldwide. This paper covers decades of continuous innovation in fuel design and manufacturing process which supports our outstanding fuel performance in all LWR fuel types. This paper also includes information about Westinghouse's state-of-the-art tools and methodologies utilized to improve fuel performance as well as recent developments in fuel cladding material. (author)

  20. Innovative nuclear thermal propulsion technology evaluation: Results of the NASA/DOE Task Team study

    International Nuclear Information System (INIS)

    Howe, S.; Borowski, S.; Helms, I.; Diaz, N.; Anghaie, S.; Latham, T.

    1991-01-01

    In response to findings from two NASA/DOE nuclear propulsion workshops held in the summer of 1990, six task teams were formed to continue evaluation of various nuclear propulsion concepts. The Task Team on Nuclear Thermal Propulsion (NTP) created the Innovative Concepts Subpanel to evaluate thermal propulsion concepts which did not utilize solid fuel. The Subpanel endeavored to evaluate each of the concepts on a ''level technological playing field,'' and to identify critical technologies, issues, and early proof-of-concept experiments. The concepts included the liquid core fission, the gas core fission, the fission foil reactors, explosively driven systems, fusion, and antimatter. The results of the studies by the panel will be provided. 13 refs., 6 figs., 2 tabs

  1. Thermal hydraulic stability in a pressure tube nuclear reactor

    International Nuclear Information System (INIS)

    Villani, A.; Ravetta, R.; Mansani, L.

    1986-01-01

    The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)

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

    Science.gov (United States)

    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.

  3. Application of thermal analysis in quality assurance of nuclear materials

    International Nuclear Information System (INIS)

    Raje, N.; Ghonge, D.K.; Reddy, A.V.R.

    2010-01-01

    Uranium finds its application as nuclear fuel and has stringent specifications regarding the presence of certain impurities. Magnesium diuranate (yellow cake) is the intermediate product in the hydrometallurgy of uranium. It is prepared by leaching of uranium ore using dilute sulphuric acid and subsequent ion-exchange separation of trace impurities from the leach solution followed by magnesium diuranate precipitation using magnesia. Determination of U, Mg, Na, K, Fe, Mn, Si, Al, Cd, Cr, Cu,Co, V, Mo, B, Cd, REEs, Th, chloride, sulphate and phosphate is required in yellow cake to examine whether their levels are in permissible limits. Currently, due to the processing of low-grade ores, higher levels of impurities are being encountered in the leach solution that affect the properties of magnesium diuranate (MDU). In order to meet the fuel specifications, quality assurance of MDU is essential

  4. The Efficiency of the Bimodal System Transportation

    Directory of Open Access Journals (Sweden)

    Nada Štrumberger

    2012-10-01

    Full Text Available The development of fast railway results in an increased applicationof Trailer Train bimodal system transportation. Thetraffic costs are multiply reduced, particularly the variablecosts. On the other hand the environmental pollution from exhaustgases is also reduced. Therefore, by the year 2010 cargotransport should be preponderant~v used which would be characterisedby fast electric trains producing less noise, at lowercosts and with clean environment.

  5. A bimodal flexible distribution for lifetime data

    OpenAIRE

    Ramires, Thiago G.; Ortega, Edwin M. M.; Cordeiro, Gauss M.; Hens, Niel

    2016-01-01

    A four-parameter extended bimodal lifetime model called the exponentiated log-sinh Cauchy distribution is proposed. It extends the log-sinh Cauchy and folded Cauchy distributions. We derive some of its mathematical properties including explicit expressions for the ordinary moments and generating and quantile functions. The method of maximum likelihood is used to estimate the model parameters. We implement the fit of the model in the GAMLSS package and provide the codes. The flexibility of the...

  6. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  7. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  8. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    International Nuclear Information System (INIS)

    Saha, P.; Aksan, N.; Andersen, J.; Yan, J.; Simoneau, J.P.; Leung, L.; Bertrand, F.; Aoto, K.; Kamide, H.

    2013-01-01

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs

  9. Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

    International Nuclear Information System (INIS)

    Zhou Tao; Sun Canhui; Li Zhenyang; Wang Zenghui

    2011-01-01

    Human factor errors in probabilistic safety assessment (PSA) of a nuclear power plant (NPP) can be prevented using thermal comfort analysis. In this paper, the THERP + HCR model is modified by using PMV (Predicted Mean Vote) and PPD (Predicted Percentage Dissatisfied) index system, so as to obtain the operator cognitive reliability,and to reflect and analyze human perception, thermal comfort status,and cognitive ability in a specific NPP environment. The mechanism of human factors in the PSA is analyzed by operators of skill, rule and knowledge types. The THERP + HCR model modified by thermal comfort theory can reflect the conditions in actual environment, and optimize reliability analysis of human factors. Improving human thermal comfort for different types of operators reduces adverse factors due to human errors, and provides a safe and optimum decision-making for NPPs. (authors)

  10. Gamma irradiation effects on the thermal properties of Bayfol nuclear track detector

    International Nuclear Information System (INIS)

    Marie, H.K.; Said, A.F.; Hegazy, T.M.

    2007-01-01

    Non isothermal studies were carried out using thermogravimetry (T G) and differential thermogravimetry (DTG) to obtain the thermal activation energy of decomposition for Bayfol solid state nuclear track detector (SSNTD) before and after exposure to gamma doses at levels between 5.0 and 100 KGy. Thermo-gravimetric analysis (TGA) indicated that the Bayfol detector decomposes in one main weight loss stage. Also, the thermal activation energy of decomposition was found to be dependent on the gamma dose. In addition, the variation of transition temperatures with the gamma dose has been determined using differential scanning calorimetry (DSC). The results indicate that the exposure to gamma doses at levels between 10 and 60 kGy further enhances the thermal stability of the polymer samples due to the crosslinking phenomenon. This suggests that gamma irradiation could be a suitable technique for producing a plastic material with enhanced thermal properties that can be suitable candidate for high temperature applications

  11. Thermal fatigue crack growth in mixing tees nuclear piping - An analytical approach

    International Nuclear Information System (INIS)

    Radu, V.

    2009-01-01

    The assessment of fatigue crack growth due to cyclic thermal loads arising from turbulent mixing presents significant challenges, principally due to the difficulty of establishing the actual loading spectrum. So-called sinusoidal methods represent a simplified approach in which the entire spectrum is replaced by a sine-wave variation of the temperature at the inner pipe surface. The need for multiple calculations in this process has lead to the development of analytical solutions for thermal stresses in a pipe subject to sinusoidal thermal loading, described in previous work performed at JRC IE Petten, The Netherlands, during the author's stage as seconded national expert. Based on these stress distributions solutions, the paper presents a methodology for assessment of thermal fatigue crack growth life in mixing tees nuclear piping. (author)

  12. Turbo-generator foundations for thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Gomi, Masanori

    1981-01-01

    The foundations of rotary machines are classified into rigid foundations and elastic foundations from the viewpoint of vibration technology, or into high harmonic order and low harmonic order foundations. As the foundations for turbogenerators, those of high harmonic order type, made of reinforced concrete, have been constructed in USA and Japan. While in Europe, low harmonic order type has been used, and there are many steel frame foundations besides those of reinforced concrete. In Japan, by the import of the know-how from Kraftwerk Union AG in 1963, the low harmonic order type foundations of reinforced concrete have been constructed. The external forms of the foundations of both types are compared, and the meaning of the names of foundations regarding the vibration characteristics is explained. The actual natural frequency of upper foundations, the actual vibration of total system due to machine vibration, and the strength to endure vibration are discussed. The Kraftwerk Union AG has developed and put in practical use the foundations made of reinforced concrete and equipped with steel springs as the standard structure for large turbogenerators for nuclear power generation. With this structure, the degree of harmonic order can be selected effectively low. The precast construction method for foundations using steel box type joints is introduced. (Kako, I.)

  13. Air quality assessment in the vicinity of nuclear and thermal power stations

    International Nuclear Information System (INIS)

    Sivaramasundaram, K.; Vijay Bhaskar, B.; Muthusubramanian, P.; Rajan, M.P.; Hegde, A.G.

    2007-01-01

    The status and ranking of any country, in the context of globalisation, is decided by its economic progress, which is directly linked into power generation. The power is generated by many routes and the nuclear and thermal routes are noteworthy among them. As the power production and its associated activities may cause qualitative deterioration, it is essential to study the impact of power production on atmospheric environment. In this connection, a comparative study has been carried out to assess the air quality with special reference to criteria pollutants in the vicinity of nuclear and thermal power stations. In the present investigation, the air samples are collected on weekly basis and the pollutants such as sulphur dioxide (SO 2 ), nitrogen oxides (NOx), carbon monoxide (CO), suspended particulate matter (SPM) and respirable particulate matter (RPM) are estimated by adopting standard procedures set by United States-Environmental Protection Agency (US-EPA) and Central Pollution Control Board (CPCB). As the micro meteorological parameters influence on the status of air quality, simultaneous measurements of these parameters are also carried, out during sampling. It is studied that estimated concentrations of all criteria pollutants in the vicinity of these power stations are within the permissible limits set by CPCB. On the basis of the generated database pertaining to the concentrations of criteria air pollutants in the vicinity of nuclear and thermal power stations, it is concluded that nuclear power production may be considered as a viable option in terms of environmental protection in our country. (author)

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

    Science.gov (United States)

    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.

  15. Thermal and nuclear power generation cost estimates using corporate financial statements

    International Nuclear Information System (INIS)

    Matsuo, Yuhji; Nagatomi, Yu; Murakami, Tomoko

    2012-01-01

    There are two generally accepted methods for estimating power generation costs: so-called 'model plant' method and the method using corporate financial statements. The method using corporate financial statements, though under some constraints, can provide useful information for comparing thermal and nuclear power generation costs. This study used this method for estimating thermal and nuclear power generation costs in Japan for the past five years, finding that the nuclear power generation cost remained stable at around 7 yen per kilowatt-hour (kWh) while the thermal power generation cost moved within a wide range of 9 to 12 yen/kWh in line with wild fluctuations in primary energy prices. The cost of nuclear power generation is expected to increase due to the enhancement of safety measures and accident damage compensation in the future, while there are reactor decommissioning, backend and many other costs that the financial statement-using approach cannot accurately estimate. In the future, efforts should be continued to comprehensively and accurately estimate total costs. (author)

  16. Failure at Zainsk thermal power station: lesson for thermal and nuclear power stations

    International Nuclear Information System (INIS)

    Derkach, A.L.; Klyuchnikov, A.A.; Fedorenko, G.M.; Kuz'min, V.V.

    2007-01-01

    An account of system failure at Zainsk Thermal PS on January 1-st, 1979 is given. The cause of failure - sudden unauthorized energizing of block transformer which led to a direct asynchronous start of 200 MW turbine generator from grid. The failure resulted in the explosion and fire in generator, shaft destruction, and the damage of the machine hall's roof. The core roots of the failure have been scrutinised

  17. Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics

    International Nuclear Information System (INIS)

    Bernard, J.P.; Haapalehto, T.

    1996-01-01

    The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)

  18. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  19. Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors

    International Nuclear Information System (INIS)

    Pavelescu, M.; Pioaru, C.; Ursu, I.

    1988-01-01

    Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)

  20. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

    2015-01-01

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

  1. Thermal behaviour of a nuclear power plant concrete during a serious accident

    International Nuclear Information System (INIS)

    Journeau, Ch.; Malaval, S.

    2006-01-01

    Concrete is a multi-scale heterogenous material (from cement particulates to gravel). Its behaviour with respect to high temperatures is studied in the framework of fire hazards inside tunnels and in the framework of an hypothetical serious accident at a nuclear reactor. Several reactions take place inside concrete between 100 deg. C (free water evaporation) and 900 deg. C (end of limestone de-carbonation). The thermal analysis of concrete samples, typical of a nuclear reactor containment building, has been carried out in order to foresee their behaviour in contact with a molten corium bath during a test with prototype materials. (J.S.)

  2. Nuclear future: thinking for building. Proceedings of the 12. Brazilian national meeting on reactor physics and thermal hydraulics; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12 ENFIR), 8. General congress on nuclear energy (8. CGEN), and 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  3. Nuclear Thermal Rocket (Ntr) Propulsion: A Proven Game-Changing Technology for Future Human Exploration Missions

    Science.gov (United States)

    Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

    2012-01-01

    The NTR represents the next evolutionary step in high performance rocket propulsion. It generates high thrust and has a specific impulse (Isp) of approx.900 seconds (s) or more V twice that of today s best chemical rockets. The technology is also proven. During the previous Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) nuclear rocket programs, 20 rocket reactors were designed, built and ground tested. These tests demonstrated: (1) a wide range of thrust; (2) high temperature carbide-based nuclear fuel; (3) sustained engine operation; (4) accumulated lifetime; and (5) restart capability V all the requirements needed for a human mission to Mars. Ceramic metal cermet fuel was also pursued, as a backup option. The NTR also has significant growth and evolution potential. Configured as a bimodal system, it can generate electrical power for the spacecraft. Adding an oxygen afterburner nozzle introduces a variable thrust and Isp capability and allows bipropellant operation. In NASA s recent Mars Design Reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, simple assembly and mission operations. In contrast to other advanced propulsion options, NTP requires no large technology scale-ups. In fact, the smallest engine tested during the Rover program V the 25,000 lbf (25 klbf) Pewee engine is sufficient for human Mars missions when used in a clustered engine arrangement. The Copernicus crewed spacecraft design developed in DRA 5.0 has significant capability and a human exploration strategy is outlined here that uses Copernicus and its key components for precursor near Earth asteroid (NEA) and Mars orbital missions prior to a Mars landing mission. Initially, the basic Copernicus vehicle can enable reusable 1-year round trip human missions to candidate NEAs like 1991 JW and Apophis in the late 2020 s to check out vehicle systems. Afterwards, the

  4. Image processing techniques for thermal, x-rays and nuclear radiations

    International Nuclear Information System (INIS)

    Chadda, V.K.

    1998-01-01

    The paper describes image acquisition techniques for the non-visible range of electromagnetic spectrum especially thermal, x-rays and nuclear radiations. Thermal imaging systems are valuable tools used for applications ranging from PCB inspection, hot spot studies, fire identification, satellite imaging to defense applications. Penetrating radiations like x-rays and gamma rays are used in NDT, baggage inspection, CAT scan, cardiology, radiography, nuclear medicine etc. Neutron radiography compliments conventional x-rays and gamma radiography. For these applications, image processing and computed tomography are employed for 2-D and 3-D image interpretation respectively. The paper also covers main features of image processing systems for quantitative evaluation of gray level and binary images. (author)

  5. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    Directory of Open Access Journals (Sweden)

    Ten-See Wang

    Full Text Available A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze the entire process on a three-dimensional domain. The computed flammability at the exit of the heat exchanger was less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process. Keywords: Hydrogen decomposition reactions, Hydrogen recombination reactions, Hydrogen containment process, Nuclear thermal propulsion, Ground testing

  6. Diffusion, Thermal Properties and Chemical Compatibilities of Select MAX Phases with Materials For Advanced Nuclear Systems

    Energy Technology Data Exchange (ETDEWEB)

    Barsoum, Michel [Drexel Univ., Philadelphia, PA (United States); Bentzel, Grady [Drexel Univ., Philadelphia, PA (United States); Tallman, Darin J. [Drexel Univ., Philadelphia, PA (United States); Sindelar, Robert [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, Brenda [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hoffman, Elizabeth [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-04

    The demands of Gen IV nuclear power plants for long service life under neutron irradiation at high temperature are severe. Advanced materials that would withstand high temperatures (up to 1000+ ºC) to high doses in a neutron field would be ideal for reactor internal structures and would add to the long service life and reliability of the reactors. The objective of this work is to investigate the chemical compatibility of select MAX with potential materials that are important for nuclear energy, as well as to measure the thermal transport properties as a function of neutron irradiation. The chemical counterparts chosen for this work are: pyrolytic carbon, SiC, U, Pd, FLiBe, Pb-Bi and Na, the latter 3 in the molten state. The thermal conductivities and heat capacities of non-irradiated MAX phases will be measured.

  7. 11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

    Energy Technology Data Exchange (ETDEWEB)

    Lemonnier, H. (ed.)

    2005-07-01

    The main topics covered by the NURETH 11 meeting are the thermal-hydraulics of existing and future nuclear power plants as foreseen by the Generation IV worldwide initiative. Normal operation and accidental situations are also relevant topics of the Conference. The topics cover modeling, experiments, instrumentation and numerical simulations related to flow and heat transfer in nuclear reactors with a special emphasis on the advances of multiphase CFD methods. The first part of this Book of Abstracts enumerates the Organizing Scientific Societies, the Sponsors of the Conference, the Conference Chairs, and the members of the Steering Committee and of the Technical Program Committee. The second part of this Book of Abstracts contains the list of the titles of the contributed papers. Each item includes the log number of the paper, the abstract of which can therefore be easily located in the next section of this book. The titles of the papers have been sorted out by topics to provide a synthetic view of the contributions in a selected domain. The last section of this Book includes an index of authors and co-authors with a reference to the log number(s) of their contributed paper(s). Finally, the CD-Rom of the Conference Proceedings containing the full-length papers is inserted at the inside back cover. Sessions content: A - two-phase flow and heat transfer fundamentals: computational and mathematical techniques (numerical schemes, LBM, BEM, mesh-less, etc.); contact angle and wettability phenomena; experiments and data bases for the assessment and the verification of 3D models; flow regime identification and modelling; heat transfer near critical pressure and supercritical water reactors; interfacial area (data base, modeling, measurement techniques); instrumentation techniques; micro-scale basic phenomena, fluid flow and heat transfer; scaling methods; counter current flow; B - code developments: containment analysis; core thermal-hydraulics and subchannel analysis

  8. Nuclear thermal propulsion engine based on particle bed reactor using light water steam as a propellant

    Science.gov (United States)

    Powell, James R.; Ludewig, Hans; Maise, George

    1993-01-01

    In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

  9. Thermal modeling of nuclear waste package designs for disposal in tuff

    International Nuclear Information System (INIS)

    Hockman, J.N.; O'Neal, W.C.

    1983-09-01

    Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 3 tables

  10. Nuclear thermal rocket clustering: 1, A summary of previous work and relevant issues

    International Nuclear Information System (INIS)

    Buksa, J.J.; Houts, M.G.

    1991-01-01

    A general review of the technical merits of nuclear thermal rocket clustering is presented. A summary of previous analyses performed during the Rover program is presented and used to assess clustering in the context of projected Space Exploration Initiative missions. A number of technical issues are discussed including cluster reliability, engine-out operation, neutronic coupling, shutdown core power generation, shutdown reactivity requirements, reactor kinetics, and radiation shielding. 7 refs., 3 figs., 2 tabs

  11. Thermal characterization of tubular SiC/SiC composite structures for nuclear applications

    International Nuclear Information System (INIS)

    Duquesne, Loys

    2015-01-01

    Researches on the development on SiCf/SiC refractory composites for generation IV nuclear fuel cladding led the CEA to focus on the thermal behavior of these materials. In particular, knowledge of the thermal properties is essential for designing the components. Regarding the development of the 'sandwich' cladding concept, for which the complexity and the geometry differ from the conventionally used flat tubes, usual measurement methods are unsuitable. This study reports on the characterization and modeling of the thermal behavior of these structures. The first part deals with the identification of the global thermal parameters for the different layers of a 'sandwich' cladding. For this purpose, a flash method is used and an experimental device suitable for tubular geometries was developed. A new estimation method based on the combination of both collected signals in front and rear faces allows the identification of the thermal diffusivity of tubular composites using infrared thermography. The second part focuses on a virtual material approach, established to describe the thermal behavior of a 'sandwich' cladding, starting from the measured properties of the elementary components (fibers and matrix). They are then used as input data for the heat transfer modeling. Confrontations between experimental measurements and numerical results finally allow us to understand the importance of the various key parameters governing the heat transfer. (author) [fr

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

    Science.gov (United States)

    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

  13. New bimodal pore catalysts for Fischer-Tropsch synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Shinoda, Misao; Zhang, Yi; Yoneyama, Yoshiharu; Hasegawa, Kiyoshi; Tsubaki, Noritatsu [Department of Material System and Life Science, School of Engineering, Toyama University, Gofuku 3190, Toyama 930-8555 (Japan)

    2004-11-15

    A simple preparation method of bimodal pore supports was developed by introducing SiO{sub 2} or ZrO{sub 2} sols into large pores of SiO{sub 2} gel pellets directly. The pores of the obtained bimodal pore supports distributed distinctly as two kinds of main pores. On the other hand, the increased BET surface area and decreased pore volume, compared to those of original silica gel, indicated that the obtained bimodal pore supports formed according to the designed route. The obtained bimodal pore supports were applied in liquid-phase Fischer-Tropsch synthesis (FTS) where cobalt was supported. The bimodal pore catalysts presented the best reaction performance in liquid-phase Fischer-Tropsch synthesis (FTS) as higher reaction rate and lower methane selectivities, because the spatial promotional effect of bimodal pore structure and chemical effect of the porous zirconia behaved inside the large pores of original silica gel.

  14. Non-Contact Measurement of Thermal Diffusivity in Ion-Implanted Nuclear Materials

    Science.gov (United States)

    Hofmann, F.; Mason, D. R.; Eliason, J. K.; Maznev, A. A.; Nelson, K. A.; Dudarev, S. L.

    2015-11-01

    Knowledge of mechanical and physical property evolution due to irradiation damage is essential for the development of future fission and fusion reactors. Ion-irradiation provides an excellent proxy for studying irradiation damage, allowing high damage doses without sample activation. Limited ion-penetration-depth means that only few-micron-thick damaged layers are produced. Substantial effort has been devoted to probing the mechanical properties of these thin implanted layers. Yet, whilst key to reactor design, their thermal transport properties remain largely unexplored due to a lack of suitable measurement techniques. Here we demonstrate non-contact thermal diffusivity measurements in ion-implanted tungsten for nuclear fusion armour. Alloying with transmutation elements and the interaction of retained gas with implantation-induced defects both lead to dramatic reductions in thermal diffusivity. These changes are well captured by our modelling approaches. Our observations have important implications for the design of future fusion power plants.

  15. Experiments on one-phase thermally stratified flows in nuclear reactor pipe lines

    International Nuclear Information System (INIS)

    Rezende, Hugo Cesar; Navarro, Moyses Alberto; Jordao, Elizabete; Santos, Andre Augusto Campagnole dos

    2009-01-01

    The phenomenon of thermal stratified flows occurs when two different layers of the same liquid at different temperatures flow separately in horizontal pipes without appreciable mixing. This phenomenon was not considered in the design stage of most of the operating nuclear power plants, but in last two decades it has become apparent due to the temperature monitoring of piping systems. The occurrence of temperature differences of about 200 deg C have been found in a narrow band around the hot and cold water interface in components under stratified flows. Loadings due to thermal stratification affected the integrity of safety related piping systems. This paper presents the results of a range of experiments performed to simulate one phase thermally stratified flows in geometry and flow condition representing a nuclear reactor steam generator injection nozzle. They have the objective of studying the flow configurations and understanding the evolution of the thermal stratification process. The driving parameter considered to characterize flow under stratified regime due to difference in specific masses is the Froude number. Different Froude numbers, from 0.018 to 0.22, were obtained in different testes by setting injection cold water flow rates and hot water initial temperatures as planned in the test matrix. Results are presented showing the influence of Froude number on the hot and cold water interface position, temperature gradients and striping phenomenon. (author)

  16. Reactive Sintering of Bimodal WC-Co Hardmetals

    OpenAIRE

    Marek Tarraste; Kristjan Juhani; Jüri Pirso; Mart Viljus

    2015-01-01

    Bimodal WC-Co hardmetals were produced using novel technology - reactive sintering. Milled and activated tungsten and graphite powders were mixed with commercial coarse grained WC-Co powder and then sintered. The microstructure of produced materials was free of defects and consisted of evenly distributed coarse and fine tungsten carbide grains in cobalt binder. The microstructure, hardness and fracture toughness of reactive sintered bimodal WC-Co hardmetals is exhibited. Developed bimodal har...

  17. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Abbott, L.

    1985-09-01

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures

  18. Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L [ed.

    1985-09-01

    An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures.

  19. Certain aspects of the environmental impact of nuclear power engineering and thermal power engineering

    International Nuclear Information System (INIS)

    Malenchenko, A.F.

    1979-01-01

    A review is made of the both environmental impact and hazard to man resulting from nuclear power engineering comparing with those of thermal power engineering. At present, in addition to such criteria, as physical-chemical characteristic of energy sources, their efficiency and accessibility for exploitation, new requirements were substantiated in relation to safety of their utilization for environment. So, one of essential problems of nuclear power engineering development consists in assessment and prediction of radioecological consequence. The analysis and operating experience of more than 1000 reactor/years with no accidents and harm for pupulation show, that in respect to impact on environment and man nuclear power engineering is much more safe in comparison with energy sources using tradidional fossile fuel

  20. Analyses of thermal plume of Cernavoda nuclear power plant by satellite remote sensing data

    Science.gov (United States)

    Zoran, M. A.; Nicolae, D. N.; Talianu, C. L.; Ciobanu, M.; Ciuciu, J. G.

    2005-10-01

    The synergistic use of multi-temporal and multi-spectral remote sensing data offers the possibility of monitoring of environment quality in the vicinity of nuclear power plants (NPP). Advanced digital processing techniques applied to several LANDSAT, MODIS and ASTER data are used to assess the extent and magnitude of radiation and non-radiation effects on the water, near field soil, vegetation and air for NPP Cernavoda , Romania . Cernavoda Unit 1 power plant, using CANDU technology, having 706.5 MW power, is successfully in operation since 1996. Cernavoda Unit 2 which is currently under construction will be operational in 2007. Thermal discharge from nuclear reactor cooling is dissipated as waste heat in Danube-Black -Sea Canal and Danube river. Water temperature distributions captured in thermal IR imagery are correlated with meteorological parameters. Additional information regarding flooding events and earthquake risks is considered . During the winter, the thermal plume is localized to an area within a few km of the power plant, and the temperature difference between the plume and non-plume areas is about 1.5 oC. During the summer and fall, there is a larger thermal plume extending 5-6 km far along Danube Black Sea Canal, and the temperature change is about 1.0 oC. Variation of surface water temperature in the thermal plume is analyzed. The strong seasonal difference in the thermal plume is related to vertical mixing of the water column in winter and to stratification in summer. Hydrodynamic simulation leads to better understanding of the mechanisms by which waste heat from NPP Cernavoda is dissipated in the environment.

  1. Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition

    International Nuclear Information System (INIS)

    Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B.

    2016-01-01

    The UO 2 fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO 2 and UO 2 -BeO were obtained from a homogenized mixture of powders of UO 2 and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H 2 atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO 2 pellets. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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 [National Research University “Moscow State University of Civil Engineering,” (Russian Federation)

    2017-03-15

    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.

  3. Establishment of International Cooperative Network and Cooperative Research Strategy Between Korea and USA on Nuclear Thermal Hydraulics

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, Chul Hwa; Jeong, Jae Jun; Choi, Ki Yong; Kang, Kyoung Ho

    2004-07-01

    1. Scope and Objectives of the Project - Successful holding of the NURETH-10 - Analysis of the international trends in technology development and applications for nuclear thermal-hydraulics - Establishment of the international cooperative network and cooperative research strategy between Korea and USA on nuclear thermal-hydraulics 2. Research Results - Successful holding of the NURETH-10 - Analysis of the international trends in technology development and applications for nuclear thermal-hydraulics: - Establishment of international cooperative network and cooperative research strategy focused between Korea and USA on nuclear thermal-hydraulics: 3. Application Plan of the Research Results - Utilization as the basic data/information in establishing the domestic R and D directions and the international cooperative research strategy, - Application of the relevant experiences and data bases of NURETH-10 for holding future international conferences, - Promote more effective and productive research cooperation between Korea and USA

  4. Penetration in bimodal, polydisperse granular material

    KAUST Repository

    Kouraytem, N.

    2016-11-07

    We investigate the impact penetration of spheres into granular media which are compositions of two discrete size ranges, thus creating a polydisperse bimodal material. We examine the penetration depth as a function of the composition (volume fractions of the respective sizes) and impact speed. Penetration depths were found to vary between delta = 0.5D(0) and delta = 7D(0), which, for mono-modal media only, could be correlated in terms of the total drop height, H = h + delta, as in previous studies, by incorporating correction factors for the packing fraction. Bimodal data can only be collapsed by deriving a critical packing fraction for each mass fraction. The data for the mixed grains exhibit a surprising lubricating effect, which was most significant when the finest grains [d(s) similar to O(30) mu m] were added to the larger particles [d(l) similar to O(200 - 500) mu m], with a size ratio, epsilon = d(l)/d(s), larger than 3 and mass fractions over 25%, despite the increased packing fraction. We postulate that the small grains get between the large grains and reduce their intergrain friction, only when their mass fraction is sufficiently large to prevent them from simply rattling in the voids between the large particles. This is supported by our experimental observations of the largest lubrication effect produced by adding small glass beads to a bed of large sand particles with rough surfaces.

  5. Uncertainty-driven nuclear data evaluation including thermal (n,α) applied to 59Ni

    Science.gov (United States)

    Helgesson, P.; Sjöstrand, H.; Rochman, D.

    2017-11-01

    This paper presents a novel approach to the evaluation of nuclear data (ND), combining experimental data for thermal cross sections with resonance parameters and nuclear reaction modeling. The method involves sampling of various uncertain parameters, in particular uncertain components in experimental setups, and provides extensive covariance information, including consistent cross-channel correlations over the whole energy spectrum. The method is developed for, and applied to, 59Ni, but may be used as a whole, or in part, for other nuclides. 59Ni is particularly interesting since a substantial amount of 59Ni is produced in thermal nuclear reactors by neutron capture in 58Ni and since it has a non-threshold (n,α) cross section. Therefore, 59Ni gives a very important contribution to the helium production in stainless steel in a thermal reactor. However, current evaluated ND libraries contain old information for 59Ni, without any uncertainty information. The work includes a study of thermal cross section experiments and a novel combination of this experimental information, giving the full multivariate distribution of the thermal cross sections. In particular, the thermal (n,α) cross section is found to be 12.7 ± . 7 b. This is consistent with, but yet different from, current established values. Further, the distribution of thermal cross sections is combined with reported resonance parameters, and with TENDL-2015 data, to provide full random ENDF files; all of this is done in a novel way, keeping uncertainties and correlations in mind. The random files are also condensed into one single ENDF file with covariance information, which is now part of a beta version of JEFF 3.3. Finally, the random ENDF files have been processed and used in an MCNP model to study the helium production in stainless steel. The increase in the (n,α) rate due to 59Ni compared to fresh stainless steel is found to be a factor of 5.2 at a certain time in the reactor vessel, with a relative

  6. Bimodal distribution of damage morphology generated by ion implantation

    International Nuclear Information System (INIS)

    Mok, K.R.C.; Jaraiz, M.; Martin-Bragado, I.; Rubio, J.E.; Castrillo, P.; Pinacho, R.; Srinivasan, M.P.; Benistant, F.

    2005-01-01

    A nucleation and evolution model of damage based on amorphous pockets (APs) has recently been developed and implemented in an atomistic kinetic Monte Carlo simulator. In the model, APs are disordered structures (I n V m ), which are agglomerates of interstitials (I) and vacancies (V). This model has been used to study the composition and size distribution of APs during different ion implantations. Depending strongly on the dose rate, ion mass and implant temperature, the APs can evolve to a defect population where the agglomerates have a similar number of I and V (n ∼ m), or to a defect population with pure I (m ∼ 0) and pure V (n ∼ 0) clusters, or a mixture of APs and clusters. This behaviour corresponds to a bimodal (APs/clusters) distribution of damage. As the AP have different thermal stability compared to the I and V clusters, the same damage concentration obtained through different implant conditions has a different damage morphology and, consequently, exhibit a different resistance to subsequent thermal treatments

  7. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y.

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core

  8. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  9. Thermal treatment for radioactive HEPA filter media generated from nuclear facilities

    International Nuclear Information System (INIS)

    Yoon, In Ho; Choi, Wang Kyu; Lee, Suk Chol; Min, Byung Youn; Yang, Hee Chul; Lee, Kun Woo; Moon, Jei Kwon

    2012-01-01

    Many radioactive HEPA filter wastes are generated from the high radioactive facilities in operation, improvement and repair, and under decommissioning. Spent filter wastes of about 1,500 drums have been stored in the waste storage facility of the Korea Atomic Energy Research Institute (KAERI) since its operation. In the future, a lot of HEPA filters in high radioactivity will be occurred from pyroprocessing which is treatment facility for used nuclear fuel. Therefore, the technology development for the radioactive HEPA filter treatment is necessary for effective management and safe disposal for HEPA filter wastes. The thermal treatment has been known as one of the most effective technologies for volume reduction and recycling of metallic radioactive wastes. In this study, the thermal treatment for radioactive HEPA filter media was conducted for the volume reduction. The volatility and leachability for heavy metals and radionuclides in radioactive HEPA filter media were analyzed to investigate the volatilization during thermal treatment and stability after thermal treatment for safe disposal, respectively. The knowledge gained from this study will aid in the development of thermal treatment for HEPA filter media

  10. Status and Trends of Thermal-Hydraulic System Codes for Nuclear Power Plants With Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Liu Zhitao; Wang Binghua; Qin Benke; Xie Heng

    2009-01-01

    Research and development of thermal-hydraulic system codes for nuclear power plants with pressurized water reactors were analyzed on their history, status and application ranges. The important roles of best-estimate methodology, codes coupling and codes qualification were pointed out. The development models of thermal-hydraulic system codes around the world provide references to China's self-innovation. (authors)

  11. Anharmonic thermal vibrations of be metal found in the MEM nuclear density map

    International Nuclear Information System (INIS)

    Takata, Masaki; Sakata, Makoto; Larsen, F.K.; Kumazawa, Shintaro; Iversen, B.B.

    1993-01-01

    A direct observation of the thermal vibrations of Be metal was performed by the Maximum Entropy Method (MEM) using neutron single crystal data. In the previous study, the existence of the small but significant cubic anharmonicity of Be has been found by the conventional least squares refinement of the observed structure factors [Larsen, Lehmann and Merisalo (1980) Acta Cryst. A36, 159-163]. In the present study, the same data were used for the MEM analysis which are comprised of 48 reflections up to sinθ/λ = 1.41A -1 in order to obtain the high resolution nuclear density of Be without using any thermal vibrational model. It was directly visible in the MEM map that not only the cubic terms but also quartic anharmonicities exist in the thermal vibrations of Be nuclei. In order to evaluate thermal parameters of Be including anharmonic terms quantitatively, the least squares refinement of the effective one-particle potential (OPP) parameters up to quartic term was carried out by using the MEM nuclear densities around atomic sites as the data set to be fitted. It was found that the present treatment has a great advantage to decide the most appropriate model of OPP by visually comparing the model with MEM density map. As a result of the least squares refinement, the anharmonic thermal parameters are obtained as α 33 = -0.340(5)[eV/A 3 ], α 40 = 0, β 20 = 9.89(1)[eV/A 4 ] and γ 00 = 0. No other anharmonic term was significant. (author)

  12. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Denise A., E-mail: deniseadorno@hotmail.com [Escola Politécnica, Universidade de São Paulo, Av. Prof. Mello Moraes, 2463, 05508-030 São Paulo, SP (Brazil); Zimmermann, Angelo J.O.; Silva, Selma L. [Centro Tecnológico da Marinha de São Paulo – CTMSP – CEA, Estrada Sorocaba Iperó, Km 12,5, 18560-000 Iperó, SP (Brazil); Piqueira, J.R.C. [Escola Politécnica, Universidade de São Paulo, Av. Prof. Mello Moraes, 2463, 05508-030 São Paulo, SP (Brazil)

    2016-05-15

    Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 °C at a rate of 1 °C/min (∼125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.

  13. Filtration of analog data measured on-line in automated control system of thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Nazarov, V.I.; Krupnov, V.P.

    1991-01-01

    Two approaches to additional filtration of analog data in automated control systems of thermal and nuclear power plants are considered. The algorithms studied were used for the development of data reliability control subsystem in automated control system of nuclear power plants

  14. Thermal Analysis of a Nuclear Waste Repository in Argillite Host Rock

    Science.gov (United States)

    Hadgu, T.; Gomez, S. P.; Matteo, E. N.

    2017-12-01

    Disposal of high-level nuclear waste in a geological repository requires analysis of heat distribution as a result of decay heat. Such an analysis supports design of repository layout to define repository footprint as well as provide information of importance to overall design. The analysis is also used in the study of potential migration of radionuclides to the accessible environment. In this study, thermal analysis for high-level waste and spent nuclear fuel in a generic repository in argillite host rock is presented. The thermal analysis utilized both semi-analytical and numerical modeling in the near field of a repository. The semi-analytical method looks at heat transport by conduction in the repository and surroundings. The results of the simulation method are temperature histories at selected radial distances from the waste package. A 3-D thermal-hydrologic numerical model was also conducted to study fluid and heat distribution in the near field. The thermal analysis assumed a generic geological repository at 500 m depth. For the semi-analytical method, a backfilled closed repository was assumed with basic design and material properties. For the thermal-hydrologic numerical method, a repository layout with disposal in horizontal boreholes was assumed. The 3-D modeling domain covers a limited portion of the repository footprint to enable a detailed thermal analysis. A highly refined unstructured mesh was used with increased discretization near heat sources and at intersections of different materials. All simulations considered different parameter values for properties of components of the engineered barrier system (i.e. buffer, disturbed rock zone and the host rock), and different surface storage times. Results of the different modeling cases are presented and include temperature and fluid flow profiles in the near field at different simulation times. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and

  15. Effects of Magnetite Aggregate and Steel Powder on Thermal Conductivity and Porosity in Concrete for Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Han-Seung Lee

    2016-01-01

    Full Text Available Among many engineering advantages in concrete, low thermal conductivity is an attractive property. Concrete has been widely used for nuclear vessels and plant facilities for its excellent radiation shielding. The heat isolation through low thermal conductivity is actually positive for nuclear power plant concrete; however the property may cause adverse effect when fires and melt-down occur in nuclear vessel since cooling down from outer surface is almost impossible due to very low thermal conductivity. If concrete containing atomic reactor has higher thermal conductivity, the explosion risk of conductive may be partially reduced. This paper presents high thermally conductive concrete development. For the work, magnetite with varying replacements of normal aggregates and steel powder of 1.5% of volume are considered, and the equivalent thermal conductivity is evaluated. Only when the replacement ratio goes up to 30%, thermal conductivity increases rapidly to 2.5 times. Addition of steel powder is evaluated to be effective by 1.08~1.15 times. In order to evaluate the improvement of thermal conductivity, several models like ACI, DEMM, and MEM are studied, and their results are compared with test results. In the present work, the effects of steel powder and magnetite aggregate are studied not only for strength development but also for thermal behavior based on porosity.

  16. Speech Recognition and Cognitive Skills in Bimodal Cochlear Implant Users

    Science.gov (United States)

    Hua, Håkan; Johansson, Björn; Magnusson, Lennart; Lyxell, Björn; Ellis, Rachel J.

    2017-01-01

    Purpose: To examine the relation between speech recognition and cognitive skills in bimodal cochlear implant (CI) and hearing aid users. Method: Seventeen bimodal CI users (28-74 years) were recruited to the study. Speech recognition tests were carried out in quiet and in noise. The cognitive tests employed included the Reading Span Test and the…

  17. Formation of bimodal porous silica-titania monoliths by sol-gel route

    Energy Technology Data Exchange (ETDEWEB)

    Ruzimuradov, O N, E-mail: ruzimuradov@rambler.ru [Department of General Chemistry, Faculty of Chemistry, National University of Uzbekistan, 15, Vuzgorodok, Tashkent, 100174 (Uzbekistan)

    2011-10-29

    Silica-titania monoliths with micrometer-scale macroporous and nanometer-scale mesoporous structure and high titania contents are prepared by sol-gel process and phase separation. Titanium alkoxide precursor was not effective in the preparation of high titania content composites because of strong decrease in phase separation tendency. Bimodal porous gels with high titania content were obtained by using inorganic salt precursors such as titanium sulfate and titanium chloride. Various characterization techniques, including SEM, XRD, Hg porosimetry and N{sub 2} adsorption have been carried out to investigate the formation process and physical-chemical properties of silica-titania monoliths. The characterization results show that the silica-titania monoliths possess a bimodal porous structure with well-dispersed titania inside silica network. The addition of titania in silica improves the thermal stability of both macroporous and mesoporous structures.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    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.

  19. Bimodal condensation silicone elastomers as dielectric elastomers

    DEFF Research Database (Denmark)

    Yu, Liyun; Madsen, Frederikke Bahrt; Skov, Anne Ladegaard

    as well as high electrical and mechanical breakdown strengths. [1] Most model elastomers are prepared by an end-linking process using a crosslinker with a certain functionality ƒ and a linear polymer with functional groups in both ends, and the resulting networks are so-called unimodal networks where...... unimodal refers to that there is one polymer only in the system. As an alternative to unimodal networks there are the bimodal networks where two polymers with significantly different molecular weights are mixed with one crosslinker. [2]Silicone rubber can be divided into condensation type and addition type...... elastomers were prepared by mixing different mass ratios (9:1, 8:2, 7:3, 6:4, 5:5, 4:6) between long polydimethylsiloxane (PDMS) chains and short PDMS chains. The resulting elastomers were investigated with respect to their rheology, dielectric properties, tensile strength, electrical breakdown, as well...

  20. A numerical simulation package for analysis of neutronics and thermal fluids of space nuclear power and propulsion systems

    International Nuclear Information System (INIS)

    Anghaie, S.; Feller, G.J.; Peery, S.D.; Parsley, R.C.

    1993-01-01

    A system of computer codes for engineering simulation and in-depth analysis of nuclear and thermal fluid design of nuclear thermal rockets is developed. The computational system includes a neutronic solver package, a thermal fluid solver package and a propellant and materials property package. The Rocket Engine Transient Simulation (ROCETS) system code is incorporated with computational modules specific to nuclear powered engines. ROCETS features a component based performance architecture that interfaces component modules into the user designed configuration, interprets user commands, creates an executable FORTRAN computer program, and executes the program to provide output to the user. Basic design features of the Pratt ampersand Whitney XNR2000 nuclear rocket concept and its operational performance are analyzed and simulated

  1. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  2. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  3. MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

    Science.gov (United States)

    Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

    1993-01-01

    Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

  4. MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

    International Nuclear Information System (INIS)

    Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

    1993-01-01

    Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

  5. The Choice of thermal reactor systems. A report by the National Nuclear Corporation Limited

    International Nuclear Information System (INIS)

    1977-01-01

    The report to the Secretary of State in Great Britain by the National Nuclear Corporation following their assessment of the three thermal reactor systems, the AGR, PWR and SGHWR type reactors, which was performed in order to assist in the decision on the choice of thermal reactors for the U.K., is in three parts. Part I is an assessment of the three systems. It comprises: a description of the general method of assessment; a commentary in which are summarised discussions on the most important issues influencing reactor choice, i.e. safety, component failure, operational characteristics, development programme, construction programme; implications for the U.K. industry; costs; and reference design of each system. Part II consists of related questions and answers accompanied by commentaries on public acceptability and views from industry. Part III contains some conclusions including an analysis on the implications of the choices open and a summary of the main features of the assessment. (U.K.)

  6. Thermal mixing in multiple-pulse nuclear quadrupole resonance spin-locking

    International Nuclear Information System (INIS)

    Beltjukov, P A; Kibrik, G E; Furman, G B; Goren, S D

    2007-01-01

    We report on an experimental and theoretical nuclear quadrupole resonance (NQR) multiple-pulse spin-locking study of the thermal mixing process in solids containing nuclei of two different sorts, I>1/2 and S = 1/2, coupled by dipole-dipole interactions and influenced by an external magnetic field. Two coupled equations for the inverse spin temperatures of both the spin systems describing the mutual spin-lattice relaxation and the thermal mixing were obtained using the method of the nonequilibrium state operator. It is shown that the relaxation process is realized with non-exponential time dependence described by a sum of two exponents. The calculated relaxation time versus the multiple-pulse field parameters agrees well with the obtained experimental data in 1,4-dichloro-2-nitrobenzene. The calculated magnetization relaxation time versus the strength of the applied magnetic field agrees well with the obtained experimental data

  7. Development of general-purpose software to analyze the static thermal characteristic of nuclear power plant

    International Nuclear Information System (INIS)

    Nakao, Yoshinobu; Koda, Eiichi; Takahashi, Toru

    2009-01-01

    We have developed the general-purpose software by which static thermal characteristic of the power generation system is analyzed easily. This software has the notable features as follows. It has the new algorithm to solve non-linear simultaneous equations to analyze the static thermal characteristics such as heat and mass balance, efficiencies, etc. of various power generation systems. It has the flexibility for setting calculation conditions. It is able to be executed on the personal computer easily and quickly. We ensured that it is able to construct heat and mass balance diagrams of main steam system of nuclear power plant and calculate the power output and efficiencies of the system. Furthermore, we evaluated various heat recovery measures of steam generator blowdown water and found that this software could be a useful operation aid for planning effective changes in support of power stretch. (author)

  8. Geração hidrelétrica, termelétrica e nuclear Hydroelectric, thermal and nuclear generation

    Directory of Open Access Journals (Sweden)

    Luiz Pinguelli Rosa

    2007-04-01

    Full Text Available O artigo apresenta a situação da produção de energia elétrica no Brasil e expõe os problemas para a implementação de um novo modelo no setor energético e para a inclusão de termelétricas em um grande sistema hidrelétrico. Questões ambientais são consideradas, particularmente as emissões de gás de efeito estufa. Atenta ainda para a possível construção de novos reatores nucleares no Brasil e destaca a importância da conservação energética e do uso de fontes de energia renovável.The situation of electric energy generation in Brazil is presented here, showing the problems in the implementation of the new model for the Power Sector, as well as in the inclusion of thermal plants in a very big hydroelectric system. Environment issues are considered, in particular the greenhouse gas emissions. The article pays attention to the possible construction of new nuclear reactors in Brazil. It is pointed out the importance of energy conservation and of using renewable energy sources.

  9. Improvement of thermal conductivity of ceramic matrix composites for 4. generation nuclear reactors

    International Nuclear Information System (INIS)

    Cabrero, J.

    2009-11-01

    This study deals with thermal conductivity improvement of SiCf/SiC ceramic matrix composites materials to be used as cladding material in 4. generation nuclear reactor. The purpose of the study is to develop a composite for which both the temperature and irradiation effect is less pronounced on thermal conductivity of material than for SiC. This material will be used as matrix in CMC with SiC fibers. Some TiC-SiC composites with different SiC volume contents were prepared by spark plasma sintering (SPS). The sintering process enables to fabricate specimens very fast, with a very fine microstructure and without any sintering aids. Neutron irradiation has been simulated using heavy ions, at room temperature and at 500 C. Evolution of the thermal properties of irradiated materials is measured using modulated photothermal IR radiometry experiment and was related to structural evolution as function of dose and temperature. It appears that such approach is reliable to evaluate TiC potentiality as matrix in CMC. Finally, CMC with TiC matrix and SiC fibers were fabricated and both mechanical and thermal properties were measured and compare to SiCf/SiC CMC. (author)

  10. Effective thermal conductivity and diffusivity of containment wall for nuclear power plant OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun; Park, Hyun Sun [Div. of Advanced Nuclear Engineering (DANE), Pohang University of Science and Technology (POSTECH), Pohang (Korea, Republic of); Lee, Jong Hwi; Kang, Hie Chan [Mechanical Engineering Div., Kunsan National University (KNU), Gunsan (Korea, Republic of)

    2017-04-15

    The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

  11. The development of monitoring techniques for thermal stratification in nuclear plant piping

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Joo, Young Sang; Yoon, Kwang Sik; Park, Chi Seung; Choi, Ha Lim; Moon, Jae Wha; Bae, Sang Ho.

    1996-12-01

    The conventional nondestructive testing has been performed in those area which are susceptible to thermal stress in according to NRC 88-08,11. In addition to that, it is necessary to set up a monitoring system to prevent severe thermal stress to pipes in early stages and to develop the non-intrusive techniques to diagnose the check valve because the thermal stratification has been caused by the malfunction of the check valve in ECCS pipe. Thermal stratification monitoring system has been designed and installed at ECCS line permanently and surge line temporally in YG nuclear power plant. The data is acceptable in according to TASCS guide line. Also, the data originated from ISMS is useful for the arrangement of a special UT program and stress analysis. Applying a togetherness of acoustics and magnetics signal, it is possible to determine the parameters of the function of the check valve internals without disassembling it. This series of tests show that the accelerometers can be use d to measure and to differentiate the three types of impacts; metal to metal impacts mechanical rubs, and worn internal parts. The magnet sensors can be used to detect the opening/closing of stainless check and fluttering of disk. (author). 50 refs., 5 tabs., 28 figs

  12. Mathematical modelling of thermal-plume interaction at Waterford Nuclear Power Station

    International Nuclear Information System (INIS)

    Tsai, S.Y.H.

    1981-01-01

    The Waldrop plume model was used to analyze the mixing and interaction of thermal effluents in the Mississippi River resulting from heated-water discharges from the Waterford Nuclear Power Station Unit 3 and from two nearby fossil-fueled power stations. The computer program of the model was modified and expanded to accommodate the multiple intake and discharge boundary conditions at the Waterford site. Numerical results of thermal-plume temperatures for individual and combined operation of the three power stations were obtained for typical low river flow (200,000 cfs) and maximum station operating conditions. The predicted temperature distributions indicated that the surface jet discharge from Waterford Unit 3 would interact with the thermal plumes produced by the two fossil-fueled stations. The results also showed that heat recirculation between the discharge of an upstream fossil-fueled plant and the intake of Waterford Unit 3 is to be expected. However, the resulting combined temperature distributions were found to be well within the thermal standards established by the state of Louisiana

  13. Thermal Lens Spectroscopy as a 'new' analytical tool for actinide determination in nuclear reprocessing processes

    International Nuclear Information System (INIS)

    Canto, Fabrice; Couston, Laurent; Magnaldo, Alastair; Broquin, Jean-Emmanuel; Signoret, Philippe

    2008-01-01

    Thermal Lens Spectroscopy (TLS) consists of measuring the effects induced by the relaxation of molecules excited by photons. Twenty years ago, the Cea already worked on TLS. Technologic reasons impeded. But, needs in sensitive analytical methods coupled with very low sample volumes (for example, traces of Np in the COEX TM process) and also the reduction of the nuclear wastes encourage us to revisit this method thanks to the improvement of optoelectronic technologies. We can also imagine coupling TLS with micro-fluidic technologies, decreasing significantly the experiments cost. Generally two laser beams are used for TLS: one for the selective excitation by molecular absorption (inducing the thermal lens) and one for probing the thermal lens. They can be coupled with different geometries, collinear or perpendicular, depending on the application and on the laser mode. Also, many possibilities of measurement have been studied to detect the thermal lens signal: interferometry, direct intensities variations, deflection etc... In this paper, one geometrical configuration and two measurements have been theoretically evaluated. For a single photodiode detection (z-scan) the limit of detection is calculated to be near 5*10 -6 mol*L -1 for Np(IV) in dodecane. (authors)

  14. Recent advances in modeling and validation of nuclear thermal-hydraulics applications with NEPTUNE CFD - 15471

    International Nuclear Information System (INIS)

    Guingo, M.; Baudry, C.; Hassanaly, M.; Lavieville, J.; Mechitouna, N.; Merigoux, N.; Mimouni, S.; Bestion, D.; Coste, P.; Morel, C.

    2015-01-01

    NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)

  15. A model selection support system for numerical simulations of nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    Gofuku, Akio; Shimizu, Kenji; Sugano, Keiji; Yoshikawa, Hidekazu; Wakabayashi, Jiro

    1990-01-01

    In order to execute efficiently a dynamic simulation of a large-scaled engineering system such as a nuclear power plant, it is necessary to develop intelligent simulation support system for all phases of the simulation. This study is concerned with the intelligent support for the program development phase and is engaged in the adequate model selection support method by applying AI (Artificial Intelligence) techniques to execute a simulation consistent with its purpose and conditions. A proto-type expert system to support the model selection for numerical simulations of nuclear thermal-hydraulics in the case of cold leg small break loss-of-coolant accident of PWR plant is now under development on a personal computer. The steps to support the selection of both fluid model and constitutive equations for the drift flux model have been developed. Several cases of model selection were carried out and reasonable model selection results were obtained. (author)

  16. Properties of gallium arsenide alloyed with Ge and Se by irradiation in nuclear reactor thermal column

    International Nuclear Information System (INIS)

    Kolin, N.G.; Osvenskij, V.B.; Tokarevskij, V.V.; Kharchenko, V.A.; Ievlev, S.M.

    1985-01-01

    Dependences of electrophysical properties as well as lattice unit spacing and density of nuclear-alloyed gallium arsenide on the fluence of reactor neutrons and heat treatment are investigated. Neutron radiation of gallium arsenide with different energy spectra is shown to differently affect material properties. Fast neutrons make the main contribution to defect formation. Concentration of compensating acceptor defects formed under GaAs radiation in a thermal column practically equals concentration of introduced donor impurities. Radiation defects of acceptor type are not annealed in the material completely even at 900-1000 deg C

  17. UMo/Al nuclear fuel plate behavior under thermal treatment (425-550 C)

    International Nuclear Information System (INIS)

    Palancher, H.; Champion, G.; Bonnin, A.; Colin, C.V.; Nassif, V.

    2013-01-01

    Nuclear fuel plates based on a γU-Mo/Al mixture are proposed for research reactors. In this work their thermal behavior in the [425; 550 C] temperature range has been studied mainly by neutron and high energy X-ray diffraction. Even if complementary studies will be necessary, the kinetics of first the growth of the interaction layer between γU-Mo and Al and second of the γU-Mo destabilization have been accurately measured. This basic work should be helpful for defining manufacturing conditions for fuel plates with optimized composition. (authors)

  18. Microstructural and Fractographic Characterization of a Thermally Embrittled Nuclear Grade Steel: Part II - Quenching and Tempering

    Directory of Open Access Journals (Sweden)

    José R. Tarpani

    2002-09-01

    Full Text Available A nuclear reactor pressure vessel steel was submitted to different quenching and tempering heat treatments aimed at simulating neutron irradiation damage. The obtained microstructures were mechanically tested and submitted to metallographic and fractographic survey. The relevant microstructural and fractographic aspects were employed in the interpretation of the mechanical performance of the thermally embrittled microstructures. A well defined correlation was determined between the elastic-plastic fracture toughness parameter J-integral and the Charpy impact energy, which was achieved for some of the Q&T microstructures.

  19. Open thermal-hydraulic research issues in western nuclear power reactors

    International Nuclear Information System (INIS)

    Hicken, E.F.

    2004-01-01

    It is common practice to access the system or component behavior of Nuclear Power Reactors on the basis of code calculations, operating experience and common engineering practice. It is, therefore, necessary to base the design and the approval by Licensing Bodies on a confidence in thermal-hydraulic and neutronic knowledge. It has be assessed if relevant phenomena may exist which have not yet been identified. This question is assessed by studying if the three basic safety requirements - the control of the reactivity - the cooling of the core - the confinement of fission products are met; if these requirements are met, the plant is 'safe'. Because these basic safety requirements are generally met - with the possible exception of containment integrity when exposed to loads from Severe Accidents - there are no additional needs for research to validate the safety level of existing reactors. However, research for new designs with a higher safety level or new phenomena might be mandatory. Also some additional research will be necessary to keep the level of safety (e.g. due to aging) or might be beneficial to reduce operating costs. Besides keeping a high safety level industry is aiming for lifetime extension, power increase, higher burn-ups and higher availability. Therefore, it is understandable that industry decreased its support for thermal-hydraulics for operating reactors. With regard to operation and the safety of existing reactors the author does not know of any request by licensing authorities for major R and D in thermal-hydraulics. Only in the area of Severe Accidents with core melt sequences some validation is requested. There is one area of common interest for utilities as well as for licensing bodies, namely the full scope, real-time simulators. Industry, licensing bodies and the European Union are responsible to keep the competence in nuclear safety - consequently also in thermal-hydraulics. Because advanced computer codes require much more information than

  20. Safety analysis on CANDU-6 nuclear power plant: changes in thermal hydraulic operational conditions concerning regional over power trip setpoints

    International Nuclear Information System (INIS)

    Lee, Jae Yong; Kim, Yong Bae; Kim, Jong Hyun; Son, Hyung Min

    2009-01-01

    A CANDU-6 nuclear power plant has the variable of regional overpower trip (ROPT) to prevent regional overpower within the reactor core. ROPT setpoints are calculated on the basis of channel power where dryout starts to take place in each nuclear fuel channel (i.e. critical channel power; CCP), which is determined based on various core-physical configurations and thermal hydraulic boundary conditions that may be generated throughout the entire life of a nuclear reactor. Variables included in the thermal hydraulic boundary condition (i.e. temperature of the inlet header, pressure on the outlet header, and differential pressure between inlet and outlet headers) change gradually as the number of operational years increases. As for these three operational variables, their operational constraints in consideration of reactor safety are suggested in the operational technical specifications for nuclear power plants. This paper first uses NUCIRC, a code for analyzing thermal hydraulic power at the core of heavy water nuclear reactor, to examine the impacts of changes in these thermal hydraulic boundary condition variables on CCP. To analyze the impacts of changes in the variables for thermal hydraulic boundary conditions on the safety of nuclear reactors, safety analysis is then performed on three representative types of design basis accidents in heavy water reactors-small break loss of coolant accident (SBLOCA), loss of regulations (LOR), and loss of forced circulation-using CATHENA, a thermal hydraulic safety analysis code. By performing two types of thermal hydraulic analysis, the following additional operational margins are ensured against the current operating limits: +2.1 .deg. C for the temperature of the reactor inlet header; -60kPa for differential pressure between inlet and outlet headers; and -40kPa for pressure on the reactor outlet header. By revising the operating limits on this basis, it will be possible to prevent possible reactor power cutbacks caused by

  1. Nuclear thermal source transfer unit, post-blast soil sample drying system

    Energy Technology Data Exchange (ETDEWEB)

    Wiser, Ralph S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Valencia, Matthew J [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-03

    Los Alamos National Laboratory states that its mission is “To solve national security challenges through scientific excellence.” The Science Undergraduate Laboratory Internship (SULI) programs exists to engage undergraduate students in STEM work by providing opportunity to work at DOE facilities. As an undergraduate mechanical engineering intern under the SULI program at Los Alamos during the fall semester of 2016, I had the opportunity to contribute to the mission of the Laboratory while developing skills in a STEM discipline. I worked with Technology Applications, an engineering group that supports non-proliferation, counter terrorism, and emergency response missions. This group specializes in tool design, weapons engineering, rapid prototyping, and mission training. I assisted with two major projects during my appointment Los Alamos. The first was a thermal source transportation unit, intended to safely contain a nuclear thermal source during transit. The second was a soil drying unit for use in nuclear postblast field sample collection. These projects have given me invaluable experience working alongside a team of professional engineers. Skills developed include modeling, simulation, group design, product and system design, and product testing.

  2. Validation experiments of nuclear characteristics of the fast-thermal system HERBE

    International Nuclear Information System (INIS)

    Pesic, M.; Zavaljevski, N.; Marinkovic, P.; Stefanovis, D.; Nikolic, D.; Avdic, S.

    1992-01-01

    In 1988/90 a coupled fast-thermal system HERBE at RB reactor, based on similar facilities, is designed and realized. Fast core of HERBE is built of natural U fuel in RB reactor center surrounded by the neutron filter and neutron converter located in an independent Al tank. Fast zone is surrounded by thermal neutron core driver. Designed nuclear characteristics of HERBE core are validated in the experiments described in the paper. HERBE cell parameters were calculated with developed computer codes: VESNA and DENEB. HERBE system criticality calculation are performed with 4G 2D RZ computer codes GALER and TWENTY GRAND, 1D multi-group AVERY code and 3D XYZ few-group TRITON computer code. The experiments for determination of critical level, dρ/dH, and reactivity of safety rods are accomplished in order to validate calculation results. Specific safety experiment is performed in aim to determine reactivity of flooded fast zone in possible accident. A very good agreements with calculation results are obtained and the validation procedures are presented. It is expected that HERBE will offer qualitative new opportunities for work with fast neutrons at RB reactor including nuclear data determination. (author)

  3. Gas core nuclear thermal rocket engine research and development in the former USSR

    International Nuclear Information System (INIS)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

  4. Thermal and stress analysis of hot isostatically pressed, alumina ceramic, nuclear waste containers

    International Nuclear Information System (INIS)

    Chang, Yun; Hoenig, C.L.

    1990-01-01

    The Yucca Mountain Project is studying design and fabrication options for a safe durable container in which to store nuclear waste underground at Yucca Mountain, Nevada. The ceramic container discussed here is an alternative to using a metal container. This ceramic alternative would be selected if site conditions prove too corrosive to use metals for nuclear waste storage. Some of the engineering problems addressed in this study were: the stress generated in the alumina container by compressive loads when 4000 to 40,000 psi of external pressure is applied; the thermal stress in the container during the heating and cooling processes; the temperature histories of the container in various production scenarios and the power required for typical heaters; the fastest possible turnaround time to heat, seal, and cool the container commensurate with preserving the structural integrity of the ceramic and the closure; the testing of some commercial heating elements to determine the maximum available heat output; and the trade-offs between the minimization in thermal stress and cycle time for closure. 2 refs., 23 figs., 2 tabs

  5. Overview of nuclear thermal-hydraulic activities performed by the European Commission

    International Nuclear Information System (INIS)

    Bucalossi, Andrea; Bieth, Michel; Santi, Giovanni De

    2009-01-01

    The European Community has been actively involved in Nuclear Safety since its creation in 1957 as its EURATOM treaty specifically promotes research and ensures the dissemination of technical information. Thermal-Hydraulic (T-H) related projects have also been executed to enhance nuclear safety of European Union (EU) candidate countries and countries from the former Soviet Union. These have been financed in the past via the PHARE and TACIS nuclear safety programmes which are now coming to an end and are now substituted by new EU financial instruments such as IPA and INSC. All the information is being currently gathered in specific databases. The EURATOM research framework programmes have financed in the past many multi-partner projects in the T-H field for both experimental and for code validation purposes which have been performed either directly by the Joint Research Centre (JRC) of the European Commission (EC) or indirectly by multi-partner projects. The current direct research activity in the field of T-H is mainly performed by two units of the JRC: the Safety of Present Nuclear Reactors (SPNR) and Safety of Future Nuclear Reactors (SFNR) units, both located in the JRC Institute for Energy (IE) based in the Netherlands. The main activities are performed within the European Union's Seventh Research Framework Programme in specific actions such as POS, NUSAC, CAPTURE, AMA, FANGS which involve several tasks such as: performing T-H calculations for Generation IV (GEN-IV) reactors, drafting EUR reports on specific topics, delivering training, performing dissemination of results, hosting and co-sponsoring conferences, seminars, etc. In the past, the JRC located in Ispra, Italy has performed extensive experimental research on the T-H facilities, now dismantled. Today most of the experiments data can be found in a database. This paper presents an overview of all the activities in the field of T-H performed in the European Union and its future trends. (author)

  6. Computational Fluid Dynamics Thermal Simulations of a Nuclear Fuel Canister During Drying

    Science.gov (United States)

    Trujillo, Corey

    Drying of nuclear fuel canisters is essential to ensure the long-term integrity and safety of nuclear fuel. Vacuum drying, which is one of the drying processes applied to nuclear fuel canisters, consists of lowering the gas pressure in the canister. This introduces a temperature-jump thermal resistance at the gas-solid interface that can cause the cladding temperature to rise beyond the regulated limits. In this thesis, the details of a numerical model of a TN24 PWR nuclear fuel canister filled with Westinghouse 15x15 assemblies is discussed. The model was constructed in ANSYS/Fluent to assess the peak cladding temperatures during vacuum drying and is geometrically-accurate and three-dimensional with distinct regions for the fuel, cladding, backfill, steel basket, and aluminum support. Considerations have been made for conduction in the solid and fluid regions as well as radiation in the fluid regions. A uniform temperature boundary condition of 101.7°C is used at the outside of the canister to conservatively model canister immersed in boiling water. Symmetry boundary conditions were employed such that only one-eighth of the canister was modeled. Steady-state simulations are performed for different fuel heat generation rates and helium pressures, ranging from atmospheric pressure to 100 Pa. Constant thermal accommodation coefficients, which characterize the effect of the temperature-jump thermal resistance at the gas-surface interface are employed. The peak cladding temperature and its radial and axial locations are reported. The maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit (TRH = 400°C) is also reported. The results of the three-dimensional model simulations are compared to two-dimensional simulations for the same heat generation rate and pressures. The results show that the rarefaction condition causes the temperature of the rods to significantly increase compared to the continuum condition, which

  7. Roles of factorial noise in inducing bimodal gene expression

    Science.gov (United States)

    Liu, Peijiang; Yuan, Zhanjiang; Huang, Lifang; Zhou, Tianshou

    2015-06-01

    Some gene regulatory systems can exhibit bimodal distributions of mRNA or protein although the deterministic counterparts are monostable. This noise-induced bimodality is an interesting phenomenon and has important biological implications, but it is unclear how different sources of expression noise (each source creates so-called factorial noise that is defined as a component of the total noise) contribute separately to this stochastic bimodality. Here we consider a minimal model of gene regulation, which is monostable in the deterministic case. Although simple, this system contains factorial noise of two main kinds: promoter noise due to switching between gene states and transcriptional (or translational) noise due to synthesis and degradation of mRNA (or protein). To better trace the roles of factorial noise in inducing bimodality, we also analyze two limit models, continuous and adiabatic approximations, apart from the exact model. We show that in the case of slow gene switching, the continuous model where only promoter noise is considered can exhibit bimodality; in the case of fast switching, the adiabatic model where only transcriptional or translational noise is considered can also exhibit bimodality but the exact model cannot; and in other cases, both promoter noise and transcriptional or translational noise can cooperatively induce bimodality. Since slow gene switching and large protein copy numbers are characteristics of eukaryotic cells, whereas fast gene switching and small protein copy numbers are characteristics of prokaryotic cells, we infer that eukaryotic stochastic bimodality is induced mainly by promoter noise, whereas prokaryotic stochastic bimodality is induced primarily by transcriptional or translational noise.

  8. Nuclear future: thinking for building. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8. General congress on nuclear energy; 12. Brazilian national meeting on reactor physics and thermal hydraulics

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12. ENFIR), the 8. General congress on nuclear energy (8. CGEN), and the 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  9. Thermal oxidation of nuclear graphite: A large scale waste treatment option.

    Directory of Open Access Journals (Sweden)

    Alex Theodosiou

    Full Text Available This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF. Particulate samples of Magnox Reactor Pile Grade-A (PGA graphite, were oxidised in both air and 60% O2, over the temperature range 400-1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700-800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000-1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.

  10. Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods

    International Nuclear Information System (INIS)

    Jiang Yijie; Cui Yi; Huo Yongzhong; Ding Shurong

    2011-01-01

    Highlights: → We establish three-dimensional finite element models for nuclear fuel rods. → The thermal-mechanical behaviors at the initial stage of burnup are obtained. → Several parameters on the in-pile performances are investigated. → The parameters have remarkable effects on the in-pile behaviors. → This study lays a foundation for optimal design and irradiation safety. - Abstract: In order to implement numerical simulation of the thermal-mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal-mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal-mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm 3 , the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm 2 K to 0.01 W/mm 2 K, while increases up to 54.7% when h decreases from 0.01 W/mm 2 K to 0.005 W/mm 2 K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal-mechanical behaviors in the fuel rod; when the

  11. FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

  12. Thermophillic and thermotolerant fungi isolated from the thermal effluent of nuclear power generating reactors

    International Nuclear Information System (INIS)

    Rippon, J.W.; Gerhold, R.; Heath, M.

    1980-01-01

    Over a period of a year, samples of water, foam, microbial mat, soil and air were obtained from areas associated with the cooling canal of a nuclear power station. The seventeen sample sites included water in the cooling canal that was thermally enriched and soil and water adjacent to, up-stream, downstream and at a distance from the generator. Air samples were taken at the plant and at various disstances from the plant. Fifty-two species of thermotolerant and thermophilic fungi were isolated. Of these, eleven species are grouped as opportunistic Mucorales or opportunistic Aspergillus sp. One veterinary pathogen was also isolated (Dactylaria gallopara). The opportunistic/pathogenic fungi were found primarily in the intake bay, the discharge bay and the cooling canal. Smaller numbers were obtained at both upstream and downstream locations. Soil samples near the cooling canal reflected an enrichment of thermophilous organisms, the previously mentioned opportunistic Mucorales and Aspergillus spp. Their numbers were found to be greater than that usually encountered in a mesophilic environment. However, air and soil samples taken at various distances from the power station indicated no greater abundance of these thermophilous fungi than would be expected from a thermal enriched environment. Our results indicate that there was no significant dissemination of thermophilous fungi from the thermal enriched effluents to the adjacent environment. These findings are consistent with the results of other investigators. (orig.)

  13. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  14. Ground Testing a Nuclear Thermal Rocket: Design of a sub-scale demonstration experiment

    Energy Technology Data Exchange (ETDEWEB)

    David Bedsun; Debra Lee; Margaret Townsend; Clay A. Cooper; Jennifer Chapman; Ronald Samborsky; Mel Bulman; Daniel Brasuell; Stanley K. Borowski

    2012-07-01

    In 2008, the NASA Mars Architecture Team found that the Nuclear Thermal Rocket (NTR) was the preferred propulsion system out of all the combinations of chemical propulsion, solar electric, nuclear electric, aerobrake, and NTR studied. Recently, the National Research Council committee reviewing the NASA Technology Roadmaps recommended the NTR as one of the top 16 technologies that should be pursued by NASA. One of the main issues with developing a NTR for future missions is the ability to economically test the full system on the ground. In the late 1990s, the Sub-surface Active Filtering of Exhaust (SAFE) concept was first proposed by Howe as a method to test NTRs at full power and full duration. The concept relied on firing the NTR into one of the test holes at the Nevada Test Site which had been constructed to test nuclear weapons. In 2011, the cost of testing a NTR and the cost of performing a proof of concept experiment were evaluated.

  15. Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

  16. Comparison of the Thermal Response of Two Calorimetric Cells Dedicated to Nuclear Heating Measurements during Calibration

    International Nuclear Information System (INIS)

    Brun, J.; Reynard, C.; De-Vita, C.; Carette, M.; Muraglia, M.; Lyoussi, A.; Fourmentel, D.; Guimbal, P.; Villard, J-F.

    2013-06-01

    Nuclear heating is a key parameter which contributes to the thermal design and the quality of in-pile experiments performed in Material Testing Reactors (MTRs) for the study of nuclear materials and fuels under irradiation. Nuclear heating is typically measured in MTRs by radiometric calorimeters. However this kind of sensor has to be suited and improved in perspective of the new experimental conditions inside the channels of Jules Horowitz Reactor (JHR). In this paper, we study the responses of two non adiabatic differential calorimeter cells having the same geometric design, but different dimensions. These experimental works are carried out during a preliminary out-of-pile calibration operating procedure of these sensors which consists in simulating the sample heating by Joule effect. The influence of the imposed electrical power and of the forced cooling flow is determined on the sensor calibration curves. A more sensitive sensor leads to a quadratic calibration curve. This behavior difference of the two calorimetric configurations is explained by means of temperature and heat flux measurements performed with a new instrumented jacket. (authors)

  17. Application of remote sensing techniques for monitoring the thermal pollution of cooling-water discharge from nuclear power plant.

    Science.gov (United States)

    Chen, Chuqun; Shi, Ping; Mao, Qingwen

    2003-08-01

    This article introduces a practical method to investigate thermal pollution in coastal water from satellite data. The intensity and distribution areas of thermal pollution by the heated effluent discharge from the nuclear power plant on Daya Bay, southern China were investigated by using Landsat-5 Thematic Mapper (TM) thermal band data from 1994 to 2001. A local algorithm was developed, based on sea-truth data of water surface temperature measured when the satellite passed over the study area. The local algorithm was then applied to estimate water temperature from TM data. It shows that the remote sensing technique provides an effective means to quantitatively monitor the intensity of thermal pollution and to retrieve a very detailed distribution pattern of thermal pollution in coastal waters. The remotely-sensed results of the thermal pollution can be used for environmental management of coastal waters.

  18. Fragmentation versus stability in bimodal coalitions

    Science.gov (United States)

    Galam, Serge

    1996-02-01

    Competing bimodal coalitions among a group of actors are discussed. First, a model from political sciences is revisited. Most of the model statements are found not to be contained in the model. Second, a new coalition model is built. It accounts for local versus global alignment with respect to the joining of a coalition. The existence of two competing world coalitions is found to yield one unique stable distribution of actors. On the opposite a unique world leadership allows the emergence of unstable relationships. In parallel to regular actors which have a clear coalition choice, “neutral”, “frustrated” and “risky” actors are produced. The cold war organisation after world war II is shown to be rather stable. The emergence of a fragmentation process from eastern group disappearance is explained as well as continuing western group stability. Some hints are obtained about possible policies to stabilize world nation relationships. European construction is analyzed with respect to European stability. Chinese stability is also discussed.

  19. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  20. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    Science.gov (United States)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  1. Experimental and computational studies of thermal mixing in next generation nuclear reactors

    Science.gov (United States)

    Landfried, Douglas Tyler

    The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

  2. Contribution to the study of thermal-hydraulic problems in nuclear reactors

    International Nuclear Information System (INIS)

    Cognet, G.

    1998-01-01

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in 'in-situ' thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  3. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  4. SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

    Science.gov (United States)

    Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

    2005-01-01

    Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

  5. Bimodal metal micro-nanopowders for powder injection molding

    Science.gov (United States)

    Pervikov, Aleksandr; Rodkevich, Nikolay; Glazkova, Elena; Lerner, Marat

    2017-12-01

    The paper studies a bimodal metal powder composition designed to prepare feedstock for powder injection molding, as well as microstructure and porosity of sintered pats. Two kinds of metal powder compositions are used, in particular, a mixture of micro- and nanopowders and a bimodal powder prepared with dispersion of steel wire. The feedstock is prepared by mixing a bimodal metal powder composition with acetylacetone and paraffin wax. The microstructure of the debound parts is observed by scanning electron microscopy. The sintered parts are characterized by density measurements and metallographic analysis. The technique of the metal powder composition proves to affect the characteristics of sintered parts. Nanoparticles are shown in the interstitial spaces among the microparticles upon mixing micro- and nanopowders, but the regular distribution of nanoparticles on the surface of microparticles is observed in the bimodal powder providing the reduction of the porosity of sintered parts and increasing the density to the proper density of steel.

  6. Functionalized bimodal mesoporous silicas as carriers for controlled aspirin delivery

    Science.gov (United States)

    Gao, Lin; Sun, Jihong; Li, Yuzhen

    2011-08-01

    The bimodal mesoporous silica modified with 3-aminopropyltriethoxysilane was performed as the aspirin carrier. The samples' structure, drug loading and release profiles were characterized with X-ray diffraction, scanning electron microscopy, N 2 adsorption and desorption, Fourier transform infrared spectroscopy, TG analysis, elemental analysis and UV-spectrophotometer. For further exploring the effects of the bimodal mesopores on the drug delivery behavior, the unimodal mesoporous material MCM-41 was also modified as the aspirin carrier. Meantime, Korsmeyer-Peppas equation ft= ktn was employed to analyze the dissolution data in details. It is indicated that the bimodal mesopores are beneficial for unrestricted drug molecules diffusing and therefore lead to a higher loading and faster releasing than that of MCM-41. The results show that the aspirin delivery properties are influenced considerably by the mesoporous matrix, whereas the large pore of bimodal mesoporous silica is the key point for the improved controlled-release properties.

  7. Reactive Sintering of Bimodal WC-Co Hardmetals

    Directory of Open Access Journals (Sweden)

    Marek Tarraste

    2015-09-01

    Full Text Available Bimodal WC-Co hardmetals were produced using novel technology - reactive sintering. Milled and activated tungsten and graphite powders were mixed with commercial coarse grained WC-Co powder and then sintered. The microstructure of produced materials was free of defects and consisted of evenly distributed coarse and fine tungsten carbide grains in cobalt binder. The microstructure, hardness and fracture toughness of reactive sintered bimodal WC-Co hardmetals is exhibited. Developed bimodal hardmetal has perspective for demanding wear applications for its increased combined hardness and toughness. Compared to coarse material there is only slight decrease in fracture toughness (K1c is 14.7 for coarse grained and 14.4 for bimodal, hardness is increased from 1290 to 1350 HV units.DOI: http://dx.doi.org/10.5755/j01.ms.21.3.7511

  8. Proceedings of fifth international topical meeting on nuclear thermal hydraulics, operations and safety

    International Nuclear Information System (INIS)

    1997-01-01

    The fifth international topical meeting on nuclear thermohydraulics, operations and safety was convened in Beijing in April 14-18, 1997. The topical meeting was sponsored by the Chinese Nuclear Society and cosponsored by American Nuclear Society, Atomic Energy Society of Japan, American Society of Mechanical Engineers, Canada Nuclear Society, Korean Nuclear Society, Mexican Nuclear Society, Nuclear Society of Slovenia and Spanish Nuclear Society. There were 262 articles were published in the meeting. They are related nuclear power thermohydraulics, operations and safety

  9. Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y. Gutknecht; Guy L. Fredrickson; Vivek Utgikar

    2012-04-01

    This project is a fundamental study to measure thermal properties (liquidus, solidus, phase transformation, and enthalpy) of molten salt systems of interest to electrorefining operations, which are used in both the fuel cycle research & development mission and the spent fuel treatment mission of the Department of Energy. During electrorefining operations the electrolyte accumulates elements more active than uranium (transuranics, fission products and bond sodium). The accumulation needs to be closely monitored because the thermal properties of the electrolyte will change as the concentration of the impurities increases. During electrorefining (processing techniques used at the Idaho National Laboratory to separate uranium from spent nuclear fuel) it is important for the electrolyte to remain in a homogeneous liquid phase for operational safeguard and criticality reasons. The phase stability of molten salts in an electrorefiner may be adversely affected by the buildup of fission products in the electrolyte. Potential situations that need to be avoided are: (i) build up of fissile elements in the salt approaching the criticality limits specified for the vessel (ii) freezing of the salts due to change in the liquidus temperature and (iii) phase separation (non-homogenous solution) of elements. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This work describes the experimental results of typical salts compositions, consisting of chlorides of strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium (as a surrogate for both uranium and plutonium), used in the processing of used nuclear fuels. Differential scanning calorimetry was used to analyze numerous salt samples providing results on the thermal properties. The property of most interest to pyroprocessing is the liquidus temperature. It was

  10. Application of neural network technology to nuclear plant thermal efficiency improvement

    International Nuclear Information System (INIS)

    Doremus, Rick; Allen Ho, S.; Bailey, James V.; Roman, Harry

    2004-01-01

    Due to the tremendous cost of building new nuclear power plants, it has become increasingly attractive to increase the power output from the existing operating power plants. There are two options that may be available to accomplish this goal. One option is to uprate the plant through licensing modification for a comfortably achievable goal of 4% to 6%. However, the licensing efforts required are no small task, vary from plant to plant, and may take years to accomplish. Some nuclear power plants may not have this option because of design, environmental, political, or geographical limitations. A second option exists that is simpler and more immediate. It focuses on improving the plant operating conditions using adaptive software that could increase the total plant output by approximately one-half percent by adjusting certain key operating parameters. No design basis analyses, hardware modifications, or licensing changes are required. In fact, this technique can be used on a plant that has already obtained licensing modification to obtain an additional one-half percent on top of the 4% to 6% increase. Public Service Electric and Gas and ARD Corporation are jointly investigating the creation of a Plant Optimization System, called POSITIVE. POSITIVE is an adaptive software tool that enables a user to analyze current plant data to identify potential problem areas and to obtain recommendations for increasing the plant's electric output. POSITIVE uses a combination of expert systems and adaptive software to analyze the thermal performance of a nuclear power plant. Historical data, obtained while the plant was above 93% power, is used to train neural networks to determine the current electric output of the plant. Once sufficiently trained, new data can be processed through the neural network. The neural network first determines the electric output associated with the current data. If the actual power matches the power predicted by the network, the neural network can be used

  11. Development of thermal scanning probe microscopy for the determination of thin films thermal conductivity: application to ceramic materials for nuclear industry

    International Nuclear Information System (INIS)

    David, L.

    2006-10-01

    Since the 1980's, various thermal metrologies have been developed to understand and characterize the phenomena of transport of thermal energy at microscopic and submicroscopic scales. Thermal Scanning Probe Microscopy (SThM) is promising. Based on the analysis of the thermal interaction between an heated probe and a sample, it permits to probe the matter at the level of micrometric size in volumes. Performed in the framework of the development of this technique, this work more particularly relates to the study of thin films thermal conductivity. We propose a new modelling of the prediction of measurement with SThM. This model allows not only the calibration of the method for the measurement of bulk material thermal conductivity but also to specify and to better describe the probe - sample thermal coupling and to estimate, from its inversion, thin films thermal conductivity. This new approach of measurement has allowed the determination of the thermal conductivity of micrometric and sub-micrometric thicknesses of meso-porous silicon thin film in particular. Our estimates for the micrometric thicknesses are in agreement with those obtained by the use of Raman spectrometry. For the lower thicknesses of film, we give new data. Our model has, moreover, allowed a better definition of the in-depth resolution of the apparatus. This one is strongly linked to the sensitivity of SThM and strongly depends on the probe-sample thermal coupling area and on the geometry of the probe used. We also developed the technique by the vacuum setting of SThM. Our first results under this environment of measurement are encouraging and validate the description of the coupling used in our model. Our method was applied to the study of ceramics (SiC, TiN, TiC and ZrC) under consideration in the composition of future nuclear fuels. Because of the limitations of SThM in terms of sensitivity to thermal conductivity and in-depth resolution, measurements were also undertaken with a modulated thermo

  12. Doubly localized surface plasmon resonance in bimodally distributed silver nanoparticles.

    Science.gov (United States)

    Ranjan, M

    2012-06-01

    Growth of bimodally distributed silver nanoparticles using sequential physical vapour deposition (PVD) is reported. Growth conditions of nanoparticles are defined in the following three steps: In the first step, nanoparticles are grown at a heated substrate and then exposed to atmosphere, in the second step, nanoparticles are vacuum annealed and finally re-deposition of silver is performed in the third step. This special way of deposition leads to the formation of bimodally distributed nanoparticles. It has been investigated that by changing the deposition time, different sets of bimodally distributed nanoparticles can be grown. Localized surface plasmon resonance (LSPR) of such bimodally distributed nanoparticles generates double plasmon resonance peaks with overlapped absorption spectra. Double plasmon resonance peaks provide a quick indication of the existence of two sets of nanoparticles. LSPR spectra of such bimodally distributed nanoparticles could be modeled with double Lorentz oscillator model. Inclusion of double Lorentz oscillator model indicates that there exist two sets of non-interacting nanoparticles resonating at different plasma frequencies. It is also reported that silver nanoparticles grown at a heated substrate, again attain the new shape while being exposed to atmosphere, followed by vacuum annealing at the same temperature. This is because of physisorption of oxygen at the silver surface and change in surface free energy. The re-shaping due to the adsorbed oxygen on the surface is responsible for bimodal size distribution of nanoparticles.

  13. Bimodal Programming: A Survey of Current Clinical Practice.

    Science.gov (United States)

    Siburt, Hannah W; Holmes, Alice E

    2015-06-01

    The purpose of this study was to determine the current clinical practice in approaches to bimodal programming in the United States. To be specific, if clinicians are recommending bimodal stimulation, who programs the hearing aid in the bimodal condition, and what method is used for programming the hearing aid? An 11-question online survey was created and sent via email to a comprehensive list of cochlear implant programming centers in the United States. The survey was sent to 360 recipients. Respondents in this study represented a diverse group of clinical settings (response rate: 26%). Results indicate little agreement about who programs the hearing aids, when they are programmed, and how they are programmed in the bimodal condition. Analysis of small versus large implant centers indicated small centers are less likely to add a device to the contralateral ear. Although a growing number of cochlear implant recipients choose to wear a hearing aid on the contralateral ear, there is inconsistency in the current clinical approach to bimodal programming. These survey results provide evidence of large variability in the current bimodal programming practices and indicate a need for more structured clinical recommendations and programming approaches.

  14. Laser-induced time-resolved spectrofluorometry and thermal lensing: applications in the nuclear industry

    International Nuclear Information System (INIS)

    Decambox, P.; Delorme, N.; Mauchien, P.; Moulin, C.

    1989-01-01

    Sensitive spectroscopic methods for the determination of actinides and lanthanides in various media are required in the nuclear industry. Laser-Induced Time-Resolved Spectrofluorometry (LITRS) for several actinides and lanthanides at very low levels and thermal lensing (TL) for oxidation state characterization allow these determinations. The set-up of LITRS is presented. Spectra, limit of detections and lifetimes obtained for U, Cm, Am, Eu, Gd, Tb, Dy, Ce, Sm, Tm are shown. Detection limit as low as 5.10 -12 M can be achieved. Examples of matrices encountered for the determination of uranium are given as well as comparison with mass spectrometry and alpha counting. The set-up of TL and performances obtained on plutonium as well as future developments are presented

  15. An integrity evaluation method of the pressure vessel of nuclear reactors under pressurized thermal shock

    International Nuclear Information System (INIS)

    Matsubara, Masaaki; Okamura, Hiroyuki.

    1987-01-01

    Present paper proposes a new algorithm of the integrity evaluation of the pressure vessel of nuclear reactors under pressurized thermal shock, PTS. This method enables us to do an effective evaluation by superimposing proposed ''PTS state-transient curves'' and ''toughness transient curves'', and is superior to a conventional one in the following points; (1) easy to get an overall view of the result of PTS event for the variations of several parameters, (2) possible to evaluate a safety margin for irradiation embrittlement, and (3) enable to construct an Expert-friendly evaluation system. In addition, the paper shows that we can execute a safety assurance test by using a flat plate model with the same thickness as that of real plant. (author)

  16. A unique nuclear thermal rocket engine using a particle bed reactor

    Science.gov (United States)

    Culver, Donald W.; Dahl, Wayne B.; McIlwain, Melvin C.

    1992-01-01

    Aerojet Propulsion Division (APD) studied 75-klb thrust Nuclear Thermal Rocket Engines (NTRE) with particle bed reactors (PBR) for application to NASA's manned Mars mission and prepared a conceptual design description of a unique engine that best satisfied mission-defined propulsion requirements and customer criteria. This paper describes the selection of a sprint-type Mars transfer mission and its impact on propulsion system design and operation. It shows how our NTRE concept was developed from this information. The resulting, unusual engine design is short, lightweight, and capable of high specific impulse operation, all factors that decrease Earth to orbit launch costs. Many unusual features of the NTRE are discussed, including nozzle area ratio variation and nozzle closure for closed loop after cooling. Mission performance calculations reveal that other well known engine options do not support this mission.

  17. Kinetic—a system code for analyzing nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

    1993-01-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  18. KINETIC: A system code for analyzing Nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-07-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of control elements (drums or rods). The worth of the control clement and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  19. Reactor physics analysis of the pin-cell Doppler effect in a thermal nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kruijf, W.J.M. de

    1995-01-01

    This report has also been published as a PhD thesis. It deals with the Doppler effect in thermal nuclear reactors. Especially the behaviour of the reactor in transient conditions is an important issue. During such a transient the radial temperature profile in a fuel pin changes. In this PhD research effective fuel temperatures have been calculated for arbitrary temperature profiles in the fuel pin with the improved slowing-down code ROLAIDS-CPM. A general expression for the effective fuel temperature in a specific fuel pin is found by defining this effective fuel temperature as a weighted sum of the temperatures in different radial fuel zones. Also, the radial power profile in a fuel pin has been calculated by performing detailed burnup calculations, which agree very well with experimental data. (orig.).

  20. An assessment of testing requirement impacts on nuclear thermal propulsion ground test facility design

    International Nuclear Information System (INIS)

    Shipers, L.R.; Ottinger, C.A.; Sanchez, L.C.

    1993-01-01

    Programs to develop solid core nuclear thermal propulsion (NTP) systems have been under way at the Department of Defense (DoD), the National Aeronautics and Space Administration (NASA), and the Department of Energy (DOE). These programs have recognized the need for a new ground test facility to support development of NTP systems. However, the different military and civilian applications have led to different ground test facility requirements. The Department of Energy (DOE) in its role as landlord and operator of the proposed research reactor test facilities has initiated an effort to explore opportunities for a common ground test facility to meet both DoD and NASA needs. The baseline design and operating limits of the proposed DoD NTP ground test facility are described. The NASA ground test facility requirements are reviewed and their potential impact on the DoD facility baseline is discussed

  1. Thermal Analysis for Environmental Qualification of Kori Nuclear power plant unit 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kwi Hyun [ENERGEO Inc., Sungnam (Korea, Republic of); Byun, Choong Sup; Song, Dong Soo [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    This paper shows the temperature profiles of safety related electrical equipment exposed to MSLB inside containment. It must be demonstrated that the LOCA qualification conditions exceed or are equivalent to the maximum calculated MSLB conditions. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPT-LT/028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. Power uprating code that is, COPATTA benchmarking study performed in six equipment at saturation temperature and surface temperature. Specially, thermal analysis carefully investigate that view point environmental qualification and NUREG- 0588 be mentioned in regard to safety-related heat sink it boundary condition or geometry information.

  2. Thermal Analysis for Environmental Qualification of Kori Nuclear power plant unit 3 and 4

    International Nuclear Information System (INIS)

    Seo, Kwi Hyun; Byun, Choong Sup; Song, Dong Soo

    2006-01-01

    This paper shows the temperature profiles of safety related electrical equipment exposed to MSLB inside containment. It must be demonstrated that the LOCA qualification conditions exceed or are equivalent to the maximum calculated MSLB conditions. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPT-LT/028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. Power uprating code that is, COPATTA benchmarking study performed in six equipment at saturation temperature and surface temperature. Specially, thermal analysis carefully investigate that view point environmental qualification and NUREG- 0588 be mentioned in regard to safety-related heat sink it boundary condition or geometry information

  3. Multi-scale analysis of nuclear reactor thermal-hydraulics-first applications using the NEPTUNE platform

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Mimouni, S.; Bestion, D.; Boudier, P.

    2005-01-01

    The NEPTUNE project aims at building a new two-phase flow thermal-hydraulics platform for nuclear reactor simulation. EDF (Electricite de France) and CEA (Commissariat a l'Energie Atomique) with the co-sponsorship of IRSN (Institut de Radioprotection et Surete Nucleaire) and FRAMATOME-ANP, are jointly developing the NEPTUNE multi-scale platform that includes new physical models and numerical methods for each of the computing scales. One usually distinguishes three different scales for industrial simulations: the 'system' scale, the 'component' scale (subchannel analysis) and CFD (Computational Fluid Dynamics). In addition DNS (Direct Numerical Simulation) can provide information at a smaller scale that can be useful for the development of the averaged scales. The NEPTUNE project also includes work on software architecture and research on new numerical methods for coupling codes since both are required to improve industrial calculations. All these R and D challenges have been defined in order to meet industrial needs and the underlying stakes (mainly the competitiveness and the safety of Nuclear Power Plants). This paper focuses on three high priority needs: DNB (Departure from Nucleate Boiling) prediction, directly linked to fuel performance; PTS (Pressurized Thermal Shock), a key issue when studying the lifespan of critical components and LBLOCA (Large Break Loss of Coolant Accident), a reference accident for safety studies. For each of these industrial applications, we provide a review of the last developments within the NEPTUNE platform and we present the first results. A particular attention is also given to physical validation and the needs for further experimental data. (authors)

  4. Thermal and physicochemical properties important for the long term behavior of nuclear waste glasses

    Science.gov (United States)

    Matzke, Hj.; Vernaz, E.

    High level nuclear waste from reprocessing of spent nuclear fuel has to be solidified in a stable matrix for safe long-time storage. Vitrification in borosilicate glasses is the technique accepted worldwide as the best combination of engineering constraints from fabrication and physicochemical properties of the matrix. A number of different glasses was developed in different national programs. The criteria and the reasons for selecting the final compositions are described briefly. Emphasis is placed on the French product R7T7 and on thermal and physicochemical properties though glasses developed in other national projects (e.g., the German product GP 98/12, etc.) are also treated. The basic physical and mechanical properties and the chemical durability of the glass in contact with water are described. The basic mechanisms of aqueous corrosion are discussed and the evolving modelling of the leaching process is dealt with, as well as effects of container material, backfill, etc. The thermal behavior has also been studied and extensive data exist on diffusion of glass constituents (Na) and of interesting elements of the waste such as the alkalis Rb and Cs or the actinides U and Pu, as well as on crystallization processes in the glass during storage at elevated temperatures. Emphasis is placed on the radiation stability of the glasses, based on extensive studies using short-lived actinides (e.g., 244Cm) or ion implantation to produce the damage expected during long storage at an accelerated rate. The radiation stability is shown to be very good, if realistic damage conditions are used. The knowledge accumulated in the past years is used to evaluate and predict the long-term evolution of the glass under storage conditions.

  5. Thermal modelling of in situ experiments supporting the nuclear waste repository program

    International Nuclear Information System (INIS)

    Filippi, M.; Mugler, C.; Caron-Charles, M.; Montarnal, P.; Martinez, J.M.; Wileveau, Y.

    2005-01-01

    Full text of publication follows: The Mont Terri underground laboratory supports a part of the research program on nuclear waste repository in deep geological formation. Among the in-situ experiments, two of them (HE-C and HED) take part to the thermal hydraulic mechanical program. In fact, the storage of high level radioactive waste induces a coupling process between the thermal field, the water content and the velocity field, then the mechanical stress of the host rock. These experiments are designed to provide data base for Opalinus clay characterization. In HE-C experiment, a central borehole is drilled at a 2 m depth, fitted with a temperature controlled heating source. The temperature field among the source is continuously monitored using thermocouples, still reaching a steady state. In a first phase, thermal conductivity is still to be accurately defined. The solution presented here, uses numerical inverse method. This requires first a direct modelling, in 3 dimensional geometry, since in fact, the rock bedding plane, as well as the thermal conductivity anisotropy greatly affect the temperature field. Variable boundary conditions, meaning to the atmospheric variation in the gallery all over the monitoring, stand as an other arising difficulty. In the examples, the numerical resolution uses a Finite Element model implemented in Cast3m, applied to a 60 000-element meshing. Inverse method proceeds then by minimizing the cost function, expressed as the quadratic difference between calculated and measured temperatures. The KALIF code developed in CEA defines different parameter sets for the direct modelling. As direct calculations needs 3 hours CPU time, the neural network method provides a good approximation of the solution for short CPU time, allowing then more than 500 000 calculations. Then the best set of parameters means to the minimization of the cost function. The following example deals with the numerical estimation of the longitudinal and transverse

  6. The trade-off between heat tolerance and metabolic cost drives the bimodal life strategy at the air-water interface

    KAUST Repository

    Fusi, Marco

    2016-01-13

    The principle of oxygen and capacity limitation of thermal tolerance in ectotherms suggests that the long-term upper limits of an organism\\'s thermal niche are equivalent to the upper limits of the organism\\'s functional capacity for oxygen provision to tissues. Air-breathing ectotherms show wider thermal tolerances, since they can take advantage of the higher availability of oxygen in air than in water. Bimodal species move from aquatic to aerial media and switch between habitats in response to environmental variations such as cyclical or anomalous temperature fluctuations. Here we tested the prediction that bimodal species cope better with thermal stress than truly aquatic species using the crab Pachygrapsus marmoratus as a model species. When in water, oxygen consumption rates of P. marmoratus acutely rise during warming. Beyond a temperature threshold of 23 °C the crab\\'s aerobic metabolism in air remains lower than in water. In parallel, the haemolymph oxygen partial pressure of submerged animals progressive decreases during warming, while it remains low but constant during emersion. Our results demonstrate the ability of a bimodal breathing ectotherm to extend its thermal tolerance during air-breathing, suggesting that there are temperature-related physiological benefits during the evolution of the bimodal life style.

  7. Thermal and hydrothermal stability of selected polymers in a nuclear reactor environment

    Science.gov (United States)

    Kim, Jinho

    The focus of this study is the development and understanding of polymer based burnable poison rod assemblies (BPRAs) in pressurized water reactors (PWRs). This material substitution reduces the water displacement penalty at the end of cycle (EOC) currently found with the B4C/Al 2O3 BPRAs that displace moderator water in PWRs. This gives rise to a longer fuel cycle due to the extra moderation from hydrogen in polymer structures. Finding synthetic polymers that endure a severe nuclear reactor circumstance is a challenge. Aside from the proper thermal stability at the range of 350--600°C in the core for a single cycle, the hydrothermal stability at near-critical water condition (350°C, 20.7MPa) is required to maintain the safe and controlled nuclear reaction because a danger comes if water might possibly penetrate inside the burnable poison rod by the failure of zircaloy cladding. There are two approaches to obtain a boron source (burnable position material) in hydrogen containing polymers. One is to utilize the boron source directly by synthesizing boron-containing polymers. A second approach is to find commercial polymers that have an appropriate thermal, hydrothermal, radiational stability and high hydrogen content; and then add an inorganic boron source such as B4C to form a composite material. Poly (diacetylene-siloxane-carborane)s and other silicon based precursor polymers were introduced to observe their thermal and hydrothermal stability. However, we found that the degradation of Si-O-Si, which was presented in the polymer, was an unfavorable disadvantage under near-critical water (350°C, 20.7MPa) even though they formed dense network structures. In addition, the Si-O bond is quite sensitive to variety of reagents, including base and acid. Therefore, the degradation rate might be accelerated by high H+ and OH- ion concentrations at the near-critical water condition. For the second approach, a number of candidate matrix polymers were screened for new

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

    International Nuclear Information System (INIS)

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

    1987-01-01

    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 requirements 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

  9. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  10. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  11. Thermal power evaluation of the TRIGA nuclear reactor at CDTN in operations of long duration

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias; Ferreira, Andrea Vidal; Rezende, Hugo Cesar, E-mail: amir@cdtn.b, E-mail: avf@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2010-07-01

    The standard operations of nuclear research reactor IPR-R1 TRIGA located at CDTN (Belo Horizonte) usually have duration of not more than 8h. However in 2009 two operations for samples irradiations lasted about 12 hours each at a power of 100 kW. These long lasting operations started in the evening and most of them were carried out at night, when there are only small fluctuations in atmosphere temperature. Therefore the conditions were ideal for evaluating the thermal balance of the power dissipated by the reactor core through the forced cooling system. Heat balance is the standard methodology for power calibration of the IPR-R1 reactor. As in any reactor operation, the main operating parameters were monitored and stored by the Data Acquisition System developed for the reactor. These data have been used for the analysis and calculation of the evolution of several neutronic and thermalhydraulic parameters involved in the reactor operation. This paper analyzes the two long lasting operations of the IPR-R1 TRIGA and compares the recorded results for the power dissipated through the primary cooling loop with the results of the power calibration conducted in March 2009. The results corresponded to those of the thermal power calibration within the uncertainty of this methodology, indicating system stability over a period of six months. (author)

  12. The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators

    International Nuclear Information System (INIS)

    Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.

    2004-01-01

    It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur

  13. Limitations to the use of two-dimensional thermal modeling of a nuclear waste repository

    International Nuclear Information System (INIS)

    Davis, B.W.

    1979-01-01

    Thermal modeling of a nuclear waste repository is basic to most waste management predictive models. It is important that the modeling techniques accurately determine the time-dependent temperature distribution of the waste emplacement media. Recent modeling studies show that the time-dependent temperature distribution can be accurately modeled in the far-field using a 2-dimensional (2-D) planar numerical model; however, the near-field cannot be modeled accurately enough by either 2-D axisymmetric or 2-D planar numerical models for repositories in salt. The accuracy limits of 2-D modeling were defined by comparing results from 3-dimensional (3-D) TRUMP modeling with results from both 2-D axisymmetric and 2-D planar. Both TRUMP and ADINAT were employed as modeling tools. Two-dimensional results from the finite element code, ADINAT were compared with 2-D results from the finite difference code, TRUMP; they showed almost perfect correspondence in the far-field. This result adds substantially to confidence in future use of ADINAT and its companion stress code ADINA for thermal stress analysis. ADINAT was found to be somewhat sensitive to time step and mesh aspect ratio. 13 figures, 4 tables

  14. The rationale/benefits of nuclear thermal rocket propulsion for NASA's lunar space transportation system

    Science.gov (United States)

    Borowski, Stanley K.

    1994-09-01

    The solid core nuclear thermal rocket (NTR) represents the next major evolutionary step in propulsion technology. With its attractive operating characteristics, which include high specific impulse (approximately 850-1000 s) and engine thrust-to-weight (approximately 4-20), the NTR can form the basis for an efficient lunar space transportation system (LTS) capable of supporting both piloted and cargo missions. Studies conducted at the NASA Lewis Research Center indicate that an NTR-based LTS could transport a fully-fueled, cargo-laden, lunar excursion vehicle to the Moon, and return it to low Earth orbit (LEO) after mission completion, for less initial mass in LEO than an aerobraked chemical system of the type studied by NASA during its '90-Day Study.' The all-propulsive NTR-powered LTS would also be 'fully reusable' and would have a 'return payload' mass fraction of approximately 23 percent--twice that of the 'partially reusable' aerobraked chemical system. Two NTR technology options are examined--one derived from the graphite-moderated reactor concept developed by NASA and the AEC under the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) programs, and a second concept, the Particle Bed Reactor (PBR). The paper also summarizes NASA's lunar outpost scenario, compares relative performance provided by different LTS concepts, and discusses important operational issues (e.g., reusability, engine 'end-of life' disposal, etc.) associated with using this important propulsion technology.

  15. The engineering of a nuclear thermal landing and ascent vehicle utilizing indigenous Martian propellant

    International Nuclear Information System (INIS)

    Zubrin, R.M.

    1991-01-01

    The following paper reports on a design study of a novel space transportation concept known as a ''NIMF'' (Nuclear rocket using Indigenous Martian Fuel.) The NIMF is a ballistic vehicle which obtains its propellant out of the Martian air by compression and liquefaction of atmospheric CO 2 . This propellant is subsequently used to generate rocket thrust at a specific impulse of 264 s by being heated to high temperature (2800 K) gas in the NIMFs' nuclear thermal rocket engines. The vehicle is designed to provide surface to orbit and surface to surface transportation, as well as housing, for a crew of three astronauts. It is capable of refueling itself for a flight to its maximum orbit in less than 50 days. The ballistic NIMF has a mass of 44.7 tonnes and, with the assumed 2800 K propellant temperature, is capable of attaining highly energetic (250 km by 34000 km elliptical) orbits. This allows it to rendezvous with interplanetary transfer vehicles which are only very loosely bound into orbit around Mars. If a propellant temperature of 2000 K is assumed, then low Mars orbit can be attained; while if 3100 K is assumed, then the ballistic NIMF is capable of injecting itself onto a minimum energy transfer orbit to Earth in a direct ascent from the Martian surface

  16. Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

    1981-01-01

    The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

  17. Thermal tests of large recirculation cooling installations for nuclear power plants

    Science.gov (United States)

    Balunov, B. F.; Lychakov, V. D.; Il'in, V. A.; Shcheglov, A. A.; Maslov, O. P.; Rasskazova, N. A.; Rakhimov, R. Z.; Boyarov, R. A.

    2017-11-01

    The article presents the results from thermal tests of some recirculation installations for cooling air in nuclear power plant premises, including the volume under the containment. The cooling effect in such installations is produced by pumping water through their heat-transfer tubes. Air from the cooled room is blown by a fan through a bundle of transversely finned tubes and is removed to the same room after having been cooled. The finning of tubes used in the tested installations was made of Grade 08Kh18N10T and Grade 08Kh18N10 stainless steels or Grade AD1 aluminum. Steel fins were attached to the tube over their entire length by means of high-frequency welding. Aluminum fins were extruded on a lathe from the external tube sheath into which a steel tube had preliminarily been placed. Although the fin extrusion operation was accompanied by pressing the sheath inner part to the steel tube, tight contact between them over the entire surface was not fully achieved. In view of this, the air gap's thermal resistance coefficient was introduced in calculating the heat transfer between the heat-transferring media. The air gap average thickness was determined from the test results taking into account the gap variation with temperature due to different linear expansion coefficients of steel and aluminum. These tests, which are part of the acceptance tests of the considered installations, were carried out at the NPO TsKTI test facility and were mainly aimed at checking if the obtained thermal characteristics were consistent with the values calculated according to the standard recommendations with introduction, if necessary, of modifications to those recommendations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)

    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 deal 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.

  19. Thermal energy storage in rock chambers - a complement to nuclear power

    International Nuclear Information System (INIS)

    Margen, P.H.

    1971-01-01

    Within about a decade from now, the nuclear capacity on several generation systems will have become larger than the night load, thus increasing the incentive to exploit cheap night energy for daily storage schemes. In Sweden, energy storage schemes using rock cavities have been studied for a number of years. These include pumped storage schemes with lower magazines well below ground surface and gas turbine schemes with compressed air magazines. Recently preliminary studies have been made of a third form - that of storing hot high pressure water in rock cavities with a simple thermal insulation. One method of utilizing this water is as feed water for a nuclear power station, the water in the store being heated from about 73 ° C to 21 7°C at night, and the stored hot water being fed directly to the Nuclear Steam Supply System (NSSS) during the day. An increase in turbine output by about 25% can then be achieved at peak periods due to the elimination of the h.p. steam bleeding for unchanged reactor power. About 35 kWh of electricity can be recovered per m 3 of storage volume, i.e. 30 times as much as if one m 3 of cold water had been allowed to descend 450 m under gravity to the lower magazine of a pumped storage plant. This illustrates how much more effective hot water storage utilizes the space of a rock cavity than does cold water storage for a pumped storage plant even at very great depths. The paper describes the circuit proposed and the design of the accumulator to meet the requirements concerning thermal insulation (to avoid exposing the rock walls to daily temperature cycles), avoidance of risk of leakage of slightly active feed water to the surrounding ground water even under severe accident conditions such as pipe and tank ruptures, and water chemistry to avoid water containing impurities or dissolved gases from reaching the feed water circuit. A preliminary cost analysis is given which shows that the proposal allows the generation of additional blocks of

  20. Auxiliary System Load Schemes in Large Thermal and Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kuzle, I.; Bosnjak, D.; Pandzic, H.

    2010-01-01

    Uninterrupted auxiliary system power supply in large power plants is a key factor for normal operation, transient states, start-ups and shutdowns and particularly during fault conditions. Therefore, there are many challenges in designing the main electrical system as well as the auxiliary systems power supply. Depending upon the type of fuel used and the environmental control system required, a thermal power plant may consume as much as 10% of its total generation for auxiliary power, while a nuclear power plant may require only 4 - 6% auxiliaries. In general, the larger the power generating plant, the higher the voltage selected for the AC auxiliary electric system. Most stations in the 75 to 500 MW range utilize 4,2 kV as the base auxiliary system voltage. Large generating stations 500 - 1000 MW and more use voltage levels of 6,9 kV and more. Some single dedicated loads such as electric driven boiler feed pumps are supplied ba a 13,8 kV bus. While designing the auxiliary electric system, the following areas must be considered: motor starting requirements, voltage regulation requirements, short-circuit duty requirements, economic considerations, reliability and alternate sources. Auxiliary power supply can't be completely generalized and each situation should be studied on its own merits to determine the optimal solution. Naturally, nuclear power plants have more reliability requirements and safety design criteria. Main coolant-pump power supply and continuity of service to other vital loads deserve special attention. This paper presents an overview of some up-to-date power plant auxiliary load system concepts. The main types of auxiliary loads are described and the electric diagrams of the modern auxiliary system supply concepts are given. Various alternative sources of auxiliary electrical supply are considered, the advantages and disadvantages of these are compared and proposals are made for high voltage distribution systems around the thermal and nuclear plant

  1. Localization ability with bimodal hearing aids and bilateral cochlear implants

    Science.gov (United States)

    Seeber, Bernhard U.; Baumann, Uwe; Fastl, Hugo

    2004-09-01

    After successful cochlear implantation in one ear, some patients continue to use a hearing aid at the contralateral ear. They report an improved reception of speech, especially in noise, as well as a better perception of music when the hearing aid and cochlear implant are used in this bimodal combination. Some individuals in this bimodal patient group also report the impression of an improved localization ability. Similar experiences are reported by the group of bilateral cochlear implantees. In this study, a survey of 11 bimodally and 4 bilaterally equipped cochlear implant users was carried out to assess localization ability. Individuals in the bimodal implant group were all provided with the same type of hearing aid in the opposite ear, and subjects in the bilateral implant group used cochlear implants of the same manufacturer on each ear. Subjects adjusted the spot of a computer-controlled laser-pointer to the perceived direction of sound incidence in the frontal horizontal plane by rotating a trackball. Two subjects of the bimodal group who had substantial residual hearing showed localization ability in the bimodal configuration, whereas using each single device only the subject with better residual hearing was able to discriminate the side of sound origin. Five other subjects with more pronounced hearing loss displayed an ability for side discrimination through the use of bimodal aids, while four of them were already able to discriminate the side with a single device. Of the bilateral cochlear implant group one subject showed localization accuracy close to that of normal hearing subjects. This subject was also able to discriminate the side of sound origin using the first implanted device alone. The other three bilaterally equipped subjects showed limited localization ability using both devices. Among them one subject demonstrated a side-discrimination ability using only the first implanted device.

  2. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    Science.gov (United States)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  3. Merging history of three bimodal clusters

    Science.gov (United States)

    Maurogordato, S.; Sauvageot, J. L.; Bourdin, H.; Cappi, A.; Benoist, C.; Ferrari, C.; Mars, G.; Houairi, K.

    2011-01-01

    We present a combined X-ray and optical analysis of three bimodal galaxy clusters selected as merging candidates at z ~ 0.1. These targets are part of MUSIC (MUlti-Wavelength Sample of Interacting Clusters), which is a general project designed to study the physics of merging clusters by means of multi-wavelength observations. Observations include spectro-imaging with XMM-Newton EPIC camera, multi-object spectroscopy (260 new redshifts), and wide-field imaging at the ESO 3.6 m and 2.2 m telescopes. We build a global picture of these clusters using X-ray luminosity and temperature maps together with galaxy density and velocity distributions. Idealized numerical simulations were used to constrain the merging scenario for each system. We show that A2933 is very likely an equal-mass advanced pre-merger ~200 Myr before the core collapse, while A2440 and A2384 are post-merger systems (~450 Myr and ~1.5 Gyr after core collapse, respectively). In the case of A2384, we detect a spectacular filament of galaxies and gas spreading over more than 1 h-1 Mpc, which we infer to have been stripped during the previous collision. The analysis of the MUSIC sample allows us to outline some general properties of merging clusters: a strong luminosity segregation of galaxies in recent post-mergers; the existence of preferential axes - corresponding to the merging directions - along which the BCGs and structures on various scales are aligned; the concomitance, in most major merger cases, of secondary merging or accretion events, with groups infalling onto the main cluster, and in some cases the evidence of previous merging episodes in one of the main components. These results are in good agreement with the hierarchical scenario of structure formation, in which clusters are expected to form by successive merging events, and matter is accreted along large-scale filaments. Based on data obtained with the European Southern Observatory, Chile (programs 072.A-0595, 075.A-0264, and 079.A-0425

  4. Thermal Parameters of Ultra-High-Temperature Reactors for Closed-Cycle Nuclear MHD Power Plants

    International Nuclear Information System (INIS)

    Gilli, P.V.

    1968-01-01

    Thermal parameters of ultra-high temperature reactors for closed-cycle MHD power plants are investigated. The 'pressure dilemma' (core and heat exchangers requiring highworking pressures for high performance; the MHD duct requiring low working pressures for strong non-equilibrium ionization and high power density) turns out to be less serious than is often assumed because: the core power density of present-day high temperature reactors is determined by nuclear considerations rather than thermal ones; higher plant efficiency reduces thermal power; larger temperature rise and higher efficiency lead to reduced mass flow rate; for the same ratio of pumping power to net output, core pressure loss ratio may be higher than in a conventional HTR; the cost of the prestressed concrete pressure vessel is considerably reduced at lower gas pressure. Dimensional and dimensionless equations are deduced that show the dependence of core dimensions and coolant flow data on thermodynamic parameters, and on thermal properties and state of the coolant. Helium, neon, and argon are compared, as are mixtures of them, showing advantages. Film and fuel temperature differences are small in comparison with gas temperature rise, particularly for the combined MHD/steam cycle. Maximum fuel element temperature - the crucial parameter - therefore occurs at the hot end of the channel. A worked example demonstrates that power densities only slightly lower than in conventional HTR's should be achievable with helium at a pressure of 5 to 10 bar and moderate pressure losses, even if the difference between maximum fuel temperature and mixed mean gas outlet temperature is only 150 to 250°C. There may be some penalty for fuel manufacturing and fuel inventory cost. Fuel cycle cost will, however, benefit from increased plant efficiency. Radial power and temperature flattening is of paramount importance; it requires a low age factor and thus a fuel management scheme with fuel reprocessing, relatively low

  5. Nuclear Thermal Rocket/Vehicle Design Options for Future NASA Missions to the Moon and Mars

    Science.gov (United States)

    Borowski, Stanley K.; Corban, Robert R.; Mcguire, Melissa L.; Beke, Erik G.

    1995-01-01

    The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a

  6. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  7. Functionalized bimodal mesoporous silicas as carriers for controlled aspirin delivery

    International Nuclear Information System (INIS)

    Gao Lin; Sun Jihong; Li Yuzhen

    2011-01-01

    The bimodal mesoporous silica modified with 3-aminopropyltriethoxysilane was performed as the aspirin carrier. The samples' structure, drug loading and release profiles were characterized with X-ray diffraction, scanning electron microscopy, N 2 adsorption and desorption, Fourier transform infrared spectroscopy, TG analysis, elemental analysis and UV-spectrophotometer. For further exploring the effects of the bimodal mesopores on the drug delivery behavior, the unimodal mesoporous material MCM-41 was also modified as the aspirin carrier. Meantime, Korsmeyer-Peppas equation f t =kt n was employed to analyze the dissolution data in details. It is indicated that the bimodal mesopores are beneficial for unrestricted drug molecules diffusing and therefore lead to a higher loading and faster releasing than that of MCM-41. The results show that the aspirin delivery properties are influenced considerably by the mesoporous matrix, whereas the large pore of bimodal mesoporous silica is the key point for the improved controlled-release properties. - Graphical abstract: Loading (A) and release profiles (B) of aspirin in N-BMMs and N-MCM-41 indicated that BMMs have more drug loading capacity and faster release rate than that MCM-41. Highlights: → Bimodal mesoporous silicas (BMMs) and MCM-41 modified with amino group via post-treatment procedure. → Loading and release profiles of aspirin in modified BMMs and MCM-41. → Modified BMMs have more drug loading capacity and faster release rate than that modified MCM-41.

  8. Immobilization of high level nuclear wastes in sintered glasses. Devitrification evaluation produced with different thermal treatments

    International Nuclear Information System (INIS)

    Messi de Bernasconi, N.B.; Russo, D.O.; Bevilacqua, M.E.; Sterba, M.E.; Heredia, A.D.; Audero, M.A.

    1990-01-01

    This work describes immobilization of high level nuclear wastes in sintered glass, as alternative way to melting glass. Different chemical compositions of borosilicate glass with simulate waste were utilized and satisfactory results were obtained at laboratory scale. As another contribution to the materials studies by X ray powder diffraction analysis, the devitrification produced with different thermal treatments, was evaluated. The effect of the thermal history on the behaviour of fission products containing glasses has been studied by several working groups in the field of high level waste fixation. When the glass is cooled through the temperature range from 800 deg C down to less than 400 deg C (these temperatures are approximates) nucleation and crystal growth can take place. The rate of crystallization will be maximum near the transformation point but through this rate may be low at lower temperatures, devitrification can still occur over long periods of time, depending on the glass composition. It was verified that there can be an appreciable increase in leaching in some waste glass compositions owing to the presence of crystalline phases. On the other hand, other compositions show very little change in leachability and the devitrified product is often preferable as there is less tendency to cracking, particularly in massive blocks of glass. A borosilicate glass, named SG7, which was developed specially in the KfK for the hot pressing of HLW with glass frit was studied. It presents a much enhanced chemical durability than borosolicate glass developed for the melting process. The crystallization behaviour of SG7 glass products was investigated in our own experiments by annealing sintered samples up to 3000 h at temperatures between 675 and 825 deg C. The samples had contained simulated waste with noble metals, since these might act as foreign nuclei for crystallization. Results on the extent of devitrification and time- temperature- transformation curves are

  9. 900 MW CP1 nuclear steam turbine retrofit thermal effects on low pressure diaphragms

    International Nuclear Information System (INIS)

    Buguin, A.; Gruau, P.; Lamarque, F.; Huggett, J.

    2015-01-01

    The steam turbines of the Koeberg units 1 and 2 operated by ESKOM in South Africa have been retrofitted in order to mitigate the generic problems of stress corrosion cracking of the original shrunk-on disk rotor design. As already done in Belgium and France, the implementation of welded rotors improves the turbine reliability and availability. Moreover, the new technology implemented associated with a new steam path allows a significant performance improvement. With a wealth of experience in CP1 retrofit, ALSTOM has put in place new technical features in the steam path in order to further improve the heat rate. Among them, steam balance holes drilled in the rotor disks have exacerbated the thermal sensitivity of the LP diaphragms. During the commissioning of the Unit 1 LP turbines following the retrofit, the load increase led to unacceptable vibrations. An investigation program was launched to determine the root causes of the problem. This paper presents the findings following the turbine inspection, as well as the recommendations and modifications to allow a smooth return to service of the unit. In addition, the results of the root cause analysis of the vibration incident are explained. Based on finite element calculations and site measurements, ALSTOM has established that the diaphragm thermal behavior, intensified by the steam balance holes, has led to radial rubbing. It was also established that the phenomena had no effect on the diaphragms mechanical integrity. Design changes have been proposed to ensure a safe and reliable long term operation of the units. These modifications have been successfully implemented onto the Koeberg Unit 2 Nuclear Steam Turbine commissioned in November 2012. (authors)

  10. Decontamination of nuclear graphite by thermal processing; Dekontamination von Nukleargraphit durch thermische Behandlung

    Energy Technology Data Exchange (ETDEWEB)

    Florjan, Monika W.

    2010-04-15

    The main problem in view of the direct disposal of the nuclear graphite is its large volume. This waste contains long-lived and short-lived radionuclides which determine the waste strategy. The irradiated graphite possess high amount of the {sup 14}C isotope. The main object of the present work was the selective separation of {sup 14}C isotope from the isotope {sup 12}C by thermal treatment (pyrolysis, partial oxidation). A successful separation could reduce the radiotoxicity and offer a different disposal strategy. Three different graphite types were investigated. The samples originate from the reflector and from the flaking of spherical fuel elements of the high-temperature reactor (AVR) Juelich. The samples from the thermal column of the research reactor (Merlin, Juelich) were also investigated. The maximum tritium releases were obtained both in inert gas atmosphere (N{sub 2}) and under water vapour-oxidizing conditions at 1280 C and 900 C. Furthermore it could be shown that 28% of {sup 14}C could be released under inert gas conditions at a 1280 C. By additive of oxidizing agent such as water vapour and oxygen the {sup 14}C release could be increased. Under water vapour-oxidizing conditions at a temperature of 1280 C up to 93% of the {sup 14}C was separated from the graphite. The matrix corrosion of 5.4% was obtained. The selective separation of the {sup 14}C is possible, because a substantial part of the radiocarbon is bound near the grain boundary surfaces. (orig.)

  11. Stability analysis of BWR nuclear-coupled thermal-hyraulics using a simple model

    Energy Technology Data Exchange (ETDEWEB)

    Karve, A.A.; Rizwan-uddin; Dorning, J.J. [Univ. of Virginia, Charlottesville, VA (United States)

    1995-09-01

    A simple mathematical model is developed to describe the dynamics of the nuclear-coupled thermal-hydraulics in a boiling water reactor (BWR) core. The model, which incorporates the essential features of neutron kinetics, and single-phase and two-phase thermal-hydraulics, leads to simple dynamical system comprised of a set of nonlinear ordinary differential equations (ODEs). The stability boundary is determined and plotted in the inlet-subcooling-number (enthalpy)/external-reactivity operating parameter plane. The eigenvalues of the Jacobian matrix of the dynamical system also are calculated at various steady-states (fixed points); the results are consistent with those of the direct stability analysis and indicate that a Hopf bifurcation occurs as the stability boundary in the operating parameter plane is crossed. Numerical simulations of the time-dependent, nonlinear ODEs are carried out for selected points in the operating parameter plane to obtain the actual damped and growing oscillations in the neutron number density, the channel inlet flow velocity, and the other phase variables. These indicate that the Hopf bifurcation is subcritical, hence, density wave oscillations with growing amplitude could result from a finite perturbation of the system even where the steady-state is stable. The power-flow map, frequently used by reactor operators during start-up and shut-down operation of a BWR, is mapped to the inlet-subcooling-number/neutron-density (operating-parameter/phase-variable) plane, and then related to the stability boundaries for different fixed inlet velocities corresponding to selected points on the flow-control line. The stability boundaries for different fixed inlet subcooling numbers corresponding to those selected points, are plotted in the neutron-density/inlet-velocity phase variable plane and then the points on the flow-control line are related to their respective stability boundaries in this plane.

  12. Reynolds stress turbulence model applied to two-phase pressurized thermal shocks in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mérigoux, Nicolas, E-mail: nicolas.merigoux@edf.fr; Laviéville, Jérôme; Mimouni, Stéphane; Guingo, Mathieu; Baudry, Cyril

    2016-04-01

    Highlights: • NEPTUNE-CFD is used to model two-phase PTS. • k-ε model did produce some satisfactory results but also highlights some weaknesses. • A more advanced turbulence model has been developed, validated and applied for PTS. • Coupled with LIM, the first results confirmed the increased accuracy of the approach. - Abstract: Nuclear power plants are subjected to a variety of ageing mechanisms and, at the same time, exposed to potential pressurized thermal shock (PTS) – characterized by a rapid cooling of the internal Reactor Pressure Vessel (RPV) surface. In this context, NEPTUNE-CFD is used to model two-phase PTS and give an assessment on the structural integrity of the RPV. The first available choice was to use standard first order turbulence model (k-ε) to model high-Reynolds number flows encountered in Pressurized Water Reactor (PWR) primary circuits. In a first attempt, the use of k-ε model did produce some satisfactory results in terms of condensation rate and temperature field distribution on integral experiments, but also highlights some weaknesses in the way to model highly anisotropic turbulence. One way to improve the turbulence prediction – and consequently the temperature field distribution – is to opt for more advanced Reynolds Stress turbulence Model. After various verification and validation steps on separated effects cases – co-current air/steam-water stratified flows in rectangular channels, water jet impingements on water pool free surfaces – this Reynolds Stress turbulence Model (R{sub ij}-ε SSG) has been applied for the first time to thermal free surface flows under industrial conditions on COSI and TOPFLOW-PTS experiments. Coupled with the Large Interface Model, the first results confirmed the adequacy and increased accuracy of the approach in an industrial context.

  13. Design of a compact thermal ionization mass spectrometer for isotopic ratio measurement of nuclear material

    International Nuclear Information System (INIS)

    Bhatia, R.K.; Yadav, V.K.; Ravisankar, E.; Nataraju, V.; Gadkari, S.C.

    2017-01-01

    High precision isotope ratio analysis of materials of interest in nuclear and geological applications is carried out by thermal ionization mass spectrometry (TIMS) technique. One of the important mandates of Bhabha Atomic Research Centre (BARC) has been developing these instruments and several TIMS instruments have been developed and deployed at user sites covering a wide range material of interest relevant to various stages of the nuclear power cycle. The instrument designs for above applications are based on two geometries of magnetic sector ie., 15 cm sector radius and 30 cm sector radius with resolutions as 200 and 400 respectively. There has been a conscious effort to improve the the sensitivity and precision of these models by modifying the designs of the sub-systems. In the recent past, a new ion optical element viz., variable dispersion zoom optics (VDZO) was introduced in the collector system of the standard model with 30cm radius magnet, to increase the dispersion of the ion beams which enabled to fix the locations of the Faraday cups (upto 6 nos.) instead of the conventional movable ones. After establishing the usefulness of VDZO, an attempt is being made to design and develop a 20 cm magnet based TIMS which will have a much smaller foot print compared to the standard 30 cm model and also covers the usual range of elements (viz. Li - U). The ion optical design was optimized using computer simulations with SIMION 7.0 software and subsequently the mechanical design was carried out using Autocad computer software. Some of the details of this new design are presented in this abstract

  14. Thermal hydrodynamic modeling and simulation of hot-gas duct for next-generation nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Injun [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Hong, Sungdeok; Kim, Chansoo [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Bai, Cheolho; Hong, Sungyull [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Shim, Jaesool, E-mail: jshim@ynu.ac.kr [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of)

    2016-12-15

    Highlights: • Thermal hydrodynamic nonlinear model is presented to examine a hot gas duct (HGD) used in a fourth-generation nuclear power reactor. • Experiments and simulation were compared to validate the nonlinear porous model. • Natural convection and radiation are considered to study the effect on the surface temperature of the HGD. • Local Nusselt number is obtained for the optimum design of a possible next-generation HGD. - Abstract: A very high-temperature gas-cooled reactor (VHTR) is a fourth-generation nuclear power reactor that requires an intermediate loop that consists of a hot-gas duct (HGD), an intermediate heat exchanger (IHX), and a process heat exchanger for massive hydrogen production. In this study, a mathematical model and simulation were developed for the HGD in a small-scale nitrogen gas loop that was designed and manufactured by the Korea Atomic Energy Research Institute. These were used to investigate the effect of various important factors on the surface of the HGD. In the modeling, a porous model was considered for a Kaowool insulator inside the HGD. The natural convection and radiation are included in the model. For validation, the modeled external surface temperatures are compared with experimental results obtained while changing the inlet temperatures of the nitrogen working fluid. The simulation results show very good agreement with the experiments. The external surface temperatures of the HGD are obtained with respect to the porosity of insulator, emissivity of radiation, and pressure of the working fluid. The local Nusselt number is also obtained for the optimum design of a possible next-generation HGD.

  15. How to effectively compute the reliability of a thermal-hydraulic nuclear passive system

    International Nuclear Information System (INIS)

    Zio, E.; Pedroni, N.

    2011-01-01

    Research highlights: → Optimized LS is the preferred choice for failure probability estimation. → Two alternative options are suggested for uncertainty and sensitivity analyses. → SS for simulation codes requiring seconds or minutes to run. → Regression models (e.g., ANNs) for simulation codes requiring hours or days to run. - Abstract: The computation of the reliability of a thermal-hydraulic (T-H) passive system of a nuclear power plant can be obtained by (i) Monte Carlo (MC) sampling the uncertainties of the system model and parameters, (ii) computing, for each sample, the system response by a mechanistic T-H code and (iii) comparing the system response with pre-established safety thresholds, which define the success or failure of the safety function. The computational effort involved can be prohibitive because of the large number of (typically long) T-H code simulations that must be performed (one for each sample) for the statistical estimation of the probability of success or failure. The objective of this work is to provide operative guidelines to effectively handle the computation of the reliability of a nuclear passive system. Two directions of computation efficiency are considered: from one side, efficient Monte Carlo Simulation (MCS) techniques are indicated as a means to performing robust estimations with a limited number of samples: in particular, the Subset Simulation (SS) and Line Sampling (LS) methods are identified as most valuable; from the other side, fast-running, surrogate regression models (also called response surfaces or meta-models) are indicated as a valid replacement of the long-running T-H model codes: in particular, the use of bootstrapped Artificial Neural Networks (ANNs) is shown to have interesting potentials, including for uncertainty propagation. The recommendations drawn are supported by the results obtained in an illustrative application of literature.

  16. Aggressive Bimodal Communication in Domestic Dogs, Canis familiaris.

    Science.gov (United States)

    Déaux, Éloïse C; Clarke, Jennifer A; Charrier, Isabelle

    2015-01-01

    Evidence of animal multimodal signalling is widespread and compelling. Dogs' aggressive vocalisations (growls and barks) have been extensively studied, but without any consideration of the simultaneously produced visual displays. In this study we aimed to categorize dogs' bimodal aggressive signals according to the redundant/non-redundant classification framework. We presented dogs with unimodal (audio or visual) or bimodal (audio-visual) stimuli and measured their gazing and motor behaviours. Responses did not qualitatively differ between the bimodal and two unimodal contexts, indicating that acoustic and visual signals provide redundant information. We could not further classify the signal as 'equivalent' or 'enhancing' as we found evidence for both subcategories. We discuss our findings in relation to the complex signal framework, and propose several hypotheses for this signal's function.

  17. Aggressive Bimodal Communication in Domestic Dogs, Canis familiaris.

    Directory of Open Access Journals (Sweden)

    Éloïse C Déaux

    Full Text Available Evidence of animal multimodal signalling is widespread and compelling. Dogs' aggressive vocalisations (growls and barks have been extensively studied, but without any consideration of the simultaneously produced visual displays. In this study we aimed to categorize dogs' bimodal aggressive signals according to the redundant/non-redundant classification framework. We presented dogs with unimodal (audio or visual or bimodal (audio-visual stimuli and measured their gazing and motor behaviours. Responses did not qualitatively differ between the bimodal and two unimodal contexts, indicating that acoustic and visual signals provide redundant information. We could not further classify the signal as 'equivalent' or 'enhancing' as we found evidence for both subcategories. We discuss our findings in relation to the complex signal framework, and propose several hypotheses for this signal's function.

  18. Visualisation and characterisation of heterogeneous bimodal PDMS networks

    DEFF Research Database (Denmark)

    Bahrt, Frederikke; Daugaard, Anders Egede; Fleury, Clemence

    2014-01-01

    The existence of short-chain domains in heterogeneous bimodal PDMS networks has been confirmed visually, for the first time, through confocal fluorescence microscopy. The networks were prepared using a controlled reaction scheme where short PDMS chains were reacted below the gelation point...... bimodal networks with short-chain domains within a long-chain network. The average sizes of the short-chain domains were found to vary from 2.1 to 5.7 mm depending on the short-chain content. The visualised network structure could be correlated thereafter to the elastic properties, which were determined...... by rheology. All heterogeneous bimodal networks displayed significantly lower moduli than mono-modal PDMS elastomers prepared from the long polymer chains. Low-loss moduli as well as low-sol fractions indicate that low-elastic moduli can be obtained without compromising the network's structure...

  19. Heat Conductor Model and Assessment Results of the Nuclear Reactor Component Thermal-Hydraulic Analysis Code CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young; Jeong, Jae Jun

    2008-09-15

    A thermal-hydraulic analysis module, CUPID has been developed for a design and safety analysis of a nuclear reactor component. In the CUPID code, a two-fluid three-field model is adopted and the governing equations are solved on an unstructured grid to make CUPID advantageous for a flow analysis in complicated geometries. As for the numerical solution scheme, the semi-implicit method was adopted, which has proved to be stable and accurate for most applications of a nuclear reactor transient. A thermal analysis of a heat structure in a nuclear reactor component is very important for the safety analysis of a nuclear reactor as well as that of a fluid. Moreover, a conjugate heat transfer analysis between a fluid and a heat structure is indispensable in many transient situations of a nuclear reactor. For this reason, a heat conductor model was implemented into the CUPID module in this study. This report presents the governing equation, numerical method and verification results of the heat conductor model.

  20. Hydrological and thermal issues concerning a nuclear waste repository in fractured rocks

    International Nuclear Information System (INIS)

    Wang, J.S.Y.

    1991-12-01

    The characterization of the ambient conditions of a potential site and the assessment of the perturbations induced by a nuclear waste repository require hydrological and thermal investigations of the geological formations at different spatial and temporal scales. For high-level wastes, the near-field impacts depend on the heat power of waste packages and the far-field long-term perturbations depend on the cumulative heat released by the emplaced wastes. Surface interim storage of wastes for several decades could lower the near-field impacts but would have relatively small long-term effects if spent fuels were the waste forms for the repository. One major uncertainty in the assessment of repository impacts is from the variation of hydrological properties in heterogeneous media, including the effects of fractures as high-permeability flow paths for containment migration. Under stress, a natural fracture cannot be represented by the parallel plate model. The rock surface roughness, the contact area, and the saturation state in the rock matrix could significantly change the fracture flow. In recent years, the concern of fast flow through fractures in saturated media has extended to the unsaturated zones. The interactions at different scales between fractures and matrix, between fractured matrix unites and porous units, and between formations and faults are discussed

  1. Mars mission opportunity and transit time sensitivity for a nuclear thermal rocket propulsion application

    International Nuclear Information System (INIS)

    Young, A.C.; Mulqueen, J.A.; Nishimuta, E.L.; Emrich, W.J.

    1993-01-01

    President George Bush's 1989 challenge to America to support the Space Exploration Initiative (SEI) of ''Back to the Moon and Human Mission to Mars'' gives the space industry an opportunity to develop effective and efficient space transportation systems. This paper presents stage performance and requirements for a nuclear thermal rocket (NTR) Mars transportation system to support the human Mars mission of the SEI. Two classes of Mars mission profiles are considered in developing the NTR propulsion vehicle performance and requirements. The two Mars mission classes include the opposition class and conjunction class. The opposition class mission is associated with relatively short Mars stay times ranging from 30 to 90 days and total mission duration of 350 to 600 days. The conjunction class mission is associated with much longer Mars stay times ranging from 500 to 600 days and total mission durations of 875 to 1,000 days. Vehicle mass scaling equations are used to determine the NTR stage mass, size, and performance range required for different Mars mission opportunities and for different Mars mission durations. Mission opportunities considered include launch years 2010 to 2018. The 2010 opportunity is the most demanding launch opportunity and the 2018 opportunity is the least demanding opportunity. NTR vehicle mass and size sensitivity to NTR engine thrust level, engine specific impulse, NTR engine thrust-to-weight ratio, and Mars surface payload are presented. NTR propulsion parameter ranges include those associated with NERVA, particle bed reactor (PBR), low-pressure, and ceramic-metal-type engine design

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

    Energy Technology Data Exchange (ETDEWEB)

    farahani, A.A.; Corradini, M.L. [Univ. of Wisconsi, Madison, WI (United States)

    1995-09-01

    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 the 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.

  3. Terrestrial mammal fauna and habitat in environmental assessment reports of thermal and nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Yatake, Hatsuho; Nashimoto, Makoto; Chiba, Shinji [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab

    2000-04-01

    We analyzed the geological distribution of mammals, relationships between ecological distribution of mammals and land use, and vegetation type in the 49 environmental assessment reports of thermal and nuclear power stations in the coastal area of Japan. Seven orders and 17 families of 66 terrestrial mammal species including subspecies were listed from the reports. This is about 40% of the total species of terrestrial mammals observed in Japan. Mammals were divided into 3 groups: distributed in the nationwide, in limited districts, and in limited area. The geological distributions of Insectivora, Rodentia, Chiroptera and naturalized mammals, of which have not been well known, were arranged in a topographic map at the scale of 1:50,000 in this survey. The characteristics of power station sites were classified into 4 categories as follows: Industrial site, Industrial-agricultural mixed site, Industrial-agricultural-forest mixed site, and forest site. The relationships between site categories and species compositions were analyzed. The listed species were fifteen species in the industrial site, however, there were thirty six species in the forest site. The mammal species were classified into six groups by vegetation types of habitat; forest-dwelling, grassland-dwelling, farmland and orchard-dwelling, wide-dwelling except residential area, wide-dwelling mammals including residential area, and residential area-dwelling mammals. (author)

  4. Summary of particle bed reactor designs for the Space Nuclear Thermal Propulsion Program

    Science.gov (United States)

    Powell, J. R.; Ludewig, H.; Todosow, M.

    1993-09-01

    A summary report of the Particle Bed Reactor (PBR) designs considered for the space nuclear thermal propulsion program has been prepared. The first chapters outline the methods of analysis, and their validation. Monte Carlo methods are used for the physics analysis, several new algorithms are used for the fluid dynamics heat transfer and engine system analysis, and commercially available codes are used for the stress analysis. A critical experiment, prototypic of the PBR was used for the physics validation, and blowdown experiments using fuel beds of prototypic dimensions were used to validate the power extraction capabilities from particle beds. In all four different PBR rocket reactor designs were studied to varying degrees of detail. They varied in power from 400 MW to 2000 MW. These designs were all characterized by a negative prompt coefficient, due to Doppler feedback, and the feedback due to moderator heat up varied from slightly negative to slightly positive. In all practical cases, the coolant worth was positive, although core configurations with negative coolant worth could be designed. In all practical cases the thrust/weight ratio was greater than 20.

  5. Mars mission opportunity and transit time sensitivity for a nuclear thermal rocket propulsion application

    Science.gov (United States)

    Young, Archie C.; Mulqueen, John A.; Nishimuta, Ena L.; Emrich, William J.

    1993-01-01

    President George Bush's 1989 challenge to America to support the Space Exploration Initiative (SEI) of ``Back to the Moon and Human Mission to Mars'' gives the space industry an opportunity to develop effective and efficient space transportation systems. This paper presents stage performance and requirements for a nuclear thermal rocket (NTR) Mars transportation system to support the human Mars mission of the SEI. Two classes of Mars mission profiles are considered in developing the NTR propulsion vehicle performance and requirements. The two Mars mission classes include the opposition class and conjunction class. The opposition class mission is associated with relatively short Mars stay times ranging from 30 to 90 days and total mission duration of 350 to 600 days. The conjunction class mission is associated with much longer Mars stay times ranging from 500 to 600 days and total mission durations of 875 to 1,000 days. Vehicle mass scaling equations are used to determine the NTR stage mass, size, and performance range required for different Mars mission opportunities and for different Mars mission durations. Mission opportunities considered include launch years 2010 to 2018. The 2010 opportunity is the most demanding launch opportunity and the 2018 opportunity is the least demanding opportunity. NTR vehicle mass and size sensitivity to NTR engine thrust level, engine specific impulse, NTR engine thrust-to-weight ratio, and Mars surface payload are presented. NTR propulsion parameter ranges include those associated with NERVA, particle bed reactor (PBR), low-pressure, and ceramic-metal-type engine design.

  6. Development of underwater robot for cleaning cooling water intake channels in thermal and nuclear power stations

    International Nuclear Information System (INIS)

    Hirai, Harumi; Ichiryu, Taku; Takenawa, Toshiya.

    1983-01-01

    To the long intake channels for seawater in thermal and nuclear power stations, marine organisms adhere and grow, and cause resistance to the flow, separate and enter into condensers to cause the clogging or corrosion erosion of cooling tubes. At present, the regular cleaning of the channels is carried out by man power, which requires much cost and many days. The underwater robot developed recently performs this cleaning work by remote control from on the ground. The performance and endurance tests of the robot were carried out in an actual channel, and it was able to be successfully put in practical use with good results. The features of this robot are as follows. It achieves the work safely without anyone entering a channel. It can clean all surfaces including ceiling without any additional structure. It can easily move. It can remove shells of 10 cm thickness. It does not require external power source. The system comprises a robot, a power unit, a hose reel, a control wagon and an underwater monitor. The robot is powered by oil hydraulic motors, and controlled through oil hoses. Cleaning is performed with rotary brushes, while it adheres to a wall by water jet power. The construction and performance of the main components and the results of trial operation are reported. (Kako, I.)

  7. Development of a machine treating removed shells and others in thermal and nuclear power stations

    International Nuclear Information System (INIS)

    Daiho, Koichi; Iwao, Takenobu

    1981-01-01

    The living things removed form the cooling water systems in thermal and nuclear power stations, such as shells and jelly fish, have been disposed by burying in the premises, but it is the actual situation that the occurrence of bad smell and the securing of land for burying are the worries. Accordingly, a machine for deodorizing the removed living things was manufactured for trial, and the treatment experiment was carried out in Chita Power Station. This treating machine dries the removed living things around 200 deg C, and makes the deodorizing treatment. The treated products can be utilized effectively as fertilizer, and the prospect to put this machine in practical use as a waste treatment machine of resource re-utilization type was obtained. General Technical Research Institute, Chubu Electric Power Co., Inc., has developed a machine treating abandoned fish for making organic fertilizer, and its principle was applied to the development of this treating machine. The treating capacity of this machine is 1 t/day, and the power consumption is 9.3 kW. The waste oil from power stations of about 15 l/h is used as the fuel. A crusher, a constant feed screw conveyer and a rotary kiln for drying are used. In the treating experiment, about 30 t of shells and others were treated during 51 days. The results are reported. (Kako, I.)

  8. Development of high performance condensers for thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Okouchi, Isao; Takahashi, Sankichi; Tomita, Akira.

    1980-01-01

    As the trend toward the large capacity of thermal and nuclear power plants advances, condensers also become large, and from the viewpoint of energy saving in whole plants, the maintenance of high performance and reliability is strongly desired. Hitachi Ltd. responded to this demand, and repeated the basic investigation with a model condenser on the selection of condenser cooling tubes and their arrangement. As the result, balanced downflow type tube arrangement was developed, which enables smooth steam flow and the improvement of condenser performance by forming the intense flow of turbine exhaust from the upper part of tube nest downward and making the steam flow in the lower part of tube nest into radial form toward air ejecting port. This tube arrangement has been applied to actual machines, and the excellent results have been obtained. In particular, the improvement of the degree of vacuum due to the reduction of steam flow loss is advantageous for increasing the power output of turbines. Thereupon, based on the basic experiment with various models of tube arrangement and the consideration on the operational results of actual machines using this tube arrangement, the features of this technique are reported. The basic construction of tube arrangement, the steam flow in tube nest, the vacuum in condensers, the supercooling of condensate, and actual balanced downflow type condensers are described. (Kako, I.)

  9. Statistical models for thermal ageing of steel materials in nuclear power plants

    International Nuclear Information System (INIS)

    Persoz, M.

    1996-01-01

    Some category of steel materials in nuclear power plants may be subjected to thermal ageing, whose extent depends on the steel chemical composition and the ageing parameters, i.e. temperature and duration. This ageing affects the 'impact strength' of the materials, which is a mechanical property. In order to assess the residual lifetime of these components, a probabilistic study has been launched, which takes into account the scatter over the input parameters of the mechanical model. Predictive formulae for estimating the impact strength of aged materials are important input data of the model. A data base has been created with impact strength results obtained from an ageing program in laboratory and statistical treatments have been undertaken. Two kinds of model have been developed, with non linear regression methods (PROC NLIN, available in SAS/STAT). The first one, using a hyperbolic tangent function, is partly based on physical considerations, and the second one, of an exponential type, is purely statistically built. The difficulties consist in selecting the significant parameters and attributing initial values to the coefficients, which is a requirement of the NLIN procedure. This global statistical analysis has led to general models that are unction of the chemical variables and the ageing parameters. These models are as precise (if not more) as local models that had been developed earlier for some specific values of ageing temperature and ageing duration. This paper describes the data and the methodology used to build the models and analyses the results given by the SAS system. (author)

  10. Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1994-01-01

    The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

  11. Trace metal concentrations in mussels in the outfall zones of thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Krishna Kumar, P.T.; Sekimoto, Hiroshi

    2008-01-01

    Many trace elements (TE) like Mn, Fe, Cu, and Zn, occur naturally in marine environments and these TE accomplish decisive functions in humans to maintain good health. Living organisms like Mytilus galloprovincialis are a rich source of TE and are grown extensively near the industrial water outfalls. Some of these TE tend to be pollutants when their elevated levels produce deleterious effects on the ecological system. As chemical analysis for TE toxicity are expensive, organisms like Mytilus galloprovincialis can be used as monitors of environmental contamination. Most studies reported so far are directed towards the effect of a single environmental factor on marine bivalves. However in the areas receiving mixed effluents from various point and non-point sources, the studies on combined effect of two or more stresses would be a more practical approach. In this paper, We investigate the heavy metal concentrations of mercury, cadmium, lead, zinc, cooper, nickel, manganese, and chromium in Mytilus galloprovincialis to provide information on the pollution of water bodies by thermal and nuclear power plants for the choice of sites from where edible mussels can be harvested. We also propose a chemometric approach developed by us using information theory to mitigate trace element toxicity in the edible part of Mytilus galloprovincialis harvested in these sites. (author)

  12. Terrestrial mammal fauna and habitat in environmental assessment reports of thermal and nuclear power stations

    International Nuclear Information System (INIS)

    Yatake, Hatsuho; Nashimoto, Makoto; Chiba, Shinji

    2000-01-01

    We analyzed the geological distribution of mammals, relationships between ecological distribution of mammals and land use, and vegetation type in the 49 environmental assessment reports of thermal and nuclear power stations in the coastal area of Japan. Seven orders and 17 families of 66 terrestrial mammal species including subspecies were listed from the reports. This is about 40% of the total species of terrestrial mammals observed in Japan. Mammals were divided into 3 groups: distributed in the nationwide, in limited districts, and in limited area. The geological distributions of Insectivora, Rodentia, Chiroptera and naturalized mammals, of which have not been well known, were arranged in a topographic map at the scale of 1:50,000 in this survey. The characteristics of power station sites were classified into 4 categories as follows: Industrial site, Industrial-agricultural mixed site, Industrial-agricultural-forest mixed site, and forest site. The relationships between site categories and species compositions were analyzed. The listed species were fifteen species in the industrial site, however, there were thirty six species in the forest site. The mammal species were classified into six groups by vegetation types of habitat; forest-dwelling, grassland-dwelling, farmland and orchard-dwelling, wide-dwelling except residential area, wide-dwelling mammals including residential area, and residential area-dwelling mammals. (author)

  13. Preprocessor for RELAP5 code, nuclear reactor thermal hydraulics accident analysis program, using Microsoft MS-EXCEL tool

    International Nuclear Information System (INIS)

    Biaty, Patricia Andrea Paladino; Sabundjian, Gaiane

    2005-01-01

    The thermal hydraulic study in accidents and transients analyses in nuclear power plants is realized with some special tools. These programs use the best estimate analyses and have been developed to simulate accidents and transients in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code has been used as tool to licensing the nuclear facilities in our country, which is the objective of this study. The main problem when RELAP5 code is used is a lot of information necessary to simulate thermal hydraulic accidents. Moreover, there is the necessity of a reasonable amount of mathematical operations to calculation of the geometry of the components existents. Therefore, in order to facilitate the manipulation of this information, it is necessary the developing a friendly preprocessor for attainment of the mathematical calculations for RELAP5 code. One of the tools used for some of these calculations is the MS-EXCEL, which will be used in this work. (author)

  14. Elucidation technique on thermal properties data on material for nuclear power

    International Nuclear Information System (INIS)

    Baba, Tetsuya; Matsumoto, Tsuyoshi; Kishimoto, Isao; Taketoshi, Naoyuki; Arai, Teruo

    1999-01-01

    National Research Laboratory of Metrology developed a technology capable of measuring thermal diffusivity with more than 2% in precision at less than 2600degC by using laser flash method, specific heat volume and thermal emissivity with more than 3% in precision at less than 3000degC by using pulse electro-heating method, and thermal conductivity of micro specimen with 3% in precision at a range of room of room temperature to 500degC. On base of such technical potentials, this study aimed at rapidly measuring thermal properties (thermal conductivity, thermal diffusivity specific heat volume, and thermal emissivity) with precision at the highest precision in the world and ranging to ultrahigh temperature under identifying fundamental properties of materials. As a result, a data base on thermal properties capable of collecting all thermal property data obtained at this study and with excellent operability could be developed. (G.K.)

  15. Neutron thermalization in quality control of asphalts content in mixtures for paving. Adaptation of nuclear densimeters for this purpose

    International Nuclear Information System (INIS)

    Bravo R, T.; Montanez M, P.O.

    1995-01-01

    This paper shows how the neutron source of the nuclear densimeters, used for measure the humidity, can be used for measuring and making the quality control of the asphalt percentage in mixtures used for street paving. The measures are based in the neutronic thermalization processes, because the hydrogen is the main part of chemical composition of the asphalts. A calibration method for the equipment is presented. (author). 6 refs, 3 figs, 3 tabs

  16. Detection and analysis of thermal energy loss in the Atucha I nuclear power plant residual heat removal system

    International Nuclear Information System (INIS)

    Berra, Sandra; Guala, Mariana I.; Khon, Hector; Lorenzo, Andrea T.; Raffo Calderon, Maria C.; Urrutia, Guillermo

    1999-01-01

    It is presented the methodology used to detect and to measure energy losses which are existent in the Atucha I nuclear power plant. They were not directly detected, since the magnitude of those was below of the instrumentation precision which is used to measure the electric and thermal power in the plant. To achieve this work temperature special measurements were made. In this way it was possible to quantify the energy losses after operational long periods. (author)

  17. Aspects of stochastic resonance in Josephson junction, bimodal ...

    Indian Academy of Sciences (India)

    the noise amplitude helps to define maximum SNR or peak SNR for an optimum amplitude of input noise. Although .... Here we consider a typical 2-parameter bimodal cubic map defined by. Xn+1 = b + aXn − X3 n. (2) ... due to shuttling with chaotic input of a logistic map (called chaotic resonance) have been reported earlier ...

  18. Interaural bimodal pitch matching with two-formant vowels

    DEFF Research Database (Denmark)

    Guérit, François; Chalupper, Josef; Santurette, Sébastien

    2013-01-01

    For bimodal patients, with a hearing aid (HA) in one ear and a cochlear implant (CI) in the opposite ear, usually a default frequency-to-electrode map is used in the CI. This assumes that the human brain can adapt to interaural place-pitch mismatches. This “one-size-fits-all” method might be part...

  19. The Bimodal Color Distribution of Small Kuiper Belt Objects

    Science.gov (United States)

    Wong, Ian; Brown, Michael E.

    2017-04-01

    We conducted a two-night photometric survey of small Kuiper Belt objects (KBOs) near opposition using the wide-field Hyper Suprime-Cam instrument on the 8.2 m Subaru Telescope. The survey covered about 90 deg2 of sky, with each field imaged in the g and I bands. We detected 356 KBOs, ranging in absolute magnitude from 6.5 to 10.4. Filtering for high-inclination objects within the hot KBO population, we show that the g - I color distribution is strongly bimodal, indicative of two color classes—the red and very red subpopulations. After categorizing objects into the two subpopulations by color, we present the first dedicated analysis of the magnitude distributions of the individual color subpopulations and demonstrate that the two distributions are roughly identical in shape throughout the entire size range covered by our survey. Comparing the color distribution of small hot KBOs with that of Centaurs, we find that they have similar bimodal shapes, thereby providing strong confirmation of previous explanations for the attested bimodality of Centaurs. We also show that the magnitude distributions of the two KBO color subpopulations and the two color subpopulations observed in the Jupiter Trojans are statistically indistinguishable. Finally, we discuss a hypothesis describing the origin of the KBO color bimodality based on our survey results. Based on data collected at Subaru Telescope, which is operated by the National Astronomical Observatory of Japan.

  20. Asymmetric Bimodal Exponential Power Distribution on the Real Line

    Directory of Open Access Journals (Sweden)

    Mehmet Niyazi Çankaya

    2018-01-01

    Full Text Available The asymmetric bimodal exponential power (ABEP distribution is an extension of the generalized gamma distribution to the real line via adding two parameters that fit the shape of peakedness in bimodality on the real line. The special values of peakedness parameters of the distribution are a combination of half Laplace and half normal distributions on the real line. The distribution has two parameters fitting the height of bimodality, so capacity of bimodality is enhanced by using these parameters. Adding a skewness parameter is considered to model asymmetry in data. The location-scale form of this distribution is proposed. The Fisher information matrix of these parameters in ABEP is obtained explicitly. Properties of ABEP are examined. Real data examples are given to illustrate the modelling capacity of ABEP. The replicated artificial data from maximum likelihood estimates of parameters of ABEP and other distributions having an algorithm for artificial data generation procedure are provided to test the similarity with real data. A brief simulation study is presented.

  1. Stochastic resonance and chaotic resonance in bimodal maps: A ...

    Indian Academy of Sciences (India)

    We present the results of an extensive numerical study on the phenomenon of stochastic resonance in a bimodal cubic map. Both Gaussian random noise as well as deterministic chaos are used as input to drive the system between the basins. Our main result is that when two identical systems capable of stochastic ...

  2. THE IMPROVEMENT OF LOW-WASTE TECHNOLOGIES OF WORKING BODY OF WATER PREPARATION AT THERMAL AND NUCLEAR POWER PLANTS

    Directory of Open Access Journals (Sweden)

    K. D. Rymasheuskaya

    2017-01-01

    Full Text Available In the present work the main directions of water desalination technologies improving have been analyzed. Possible techniques of high-quality treatment of water that enable the reduction of amounts of environmentally hazardous substances to be discharged into the hydrosphere are indicated. The purpose of the work was to improve the ecological efficiency and the effectiveness of water treatment equipment at heat power plants when designing new and the modernizing existing water treatment schemes. In order to achieve this goal the following problems have been solved: the one of analyzing the main directions of the improvement of technologies of working body of water preparation at thermal and nuclear power plants; of analyzing the main directions of reduction of total volume of highly mineralized power plant wastewaters; of developing the technological scheme of recycling of concentrate of membrane installations and regenerants of ionite filters in acid and alkali; of developing the technological scheme of transformation of the sludge in pre-processing waste into valuable commodity products. The results of research can be applied for the design of new and the modernization of existing water treatment installations of thermal and nuclear power plants. It will enable to reduce considerably the use of natural water and the amount of chemicals added as well as the volume of wastewater and the concentration of dissolved solids in it. As a consequence, the negative impact of thermal and nuclear power plants on the hydrosphere will be reduced. 

  3. Application of flow network models of SINDA/FLUINT{sup TM} to a nuclear power plant system thermal hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)

  4. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    Energy Technology Data Exchange (ETDEWEB)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable

  5. Lanthanide oxide and phosphate nanoparticles for thermometry and bimodal imaging =

    Science.gov (United States)

    Debasu, Mengistie Leweyehu

    . Finalmente, estudam-se as propriedades de fotoluminescencia correspondentes as conversoes ascendente e descendente de energia em nanocristais de (Gd,Yb,Tb)PO4 sintetizados por via hidrotermica. A relaxividade (ressonancia magnetica) do 1H destes materiais sao investigadas, tendo em vista possiveis aplicacoes em imagem bimodal (luminescencia e ressonancia magnetica nuclear).

  6. CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-08-01

    A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  7. The U.S. Nuclear Regulatory Commission Thermal-Hydraulic Research Program: Maintaining expertise in a changing environment

    International Nuclear Information System (INIS)

    Sheron, B.W.; Shotkin, L.M.; Baratta, A.J.

    1993-01-01

    Throughout the 1970s and early 1980s, the U.S. Nuclear Regulatory Commission's (NRC's) thermal-hydraulic research program enjoyed ample funding, sponsored extensive experimental and analytical development programs, and attracted worldwide expertise. With the completion of the major experimental programs and with the promulgation of the revised emergency core-cooling system rule, both the funding and prominence of thermal-hydraulic research at the NRC have declined in recent years. This has led justifiably to the concern by some that the program may no longer have the minimal elements needed to maintain both expertise and world-class status. The purpose of this article is to describe the NRC's current thermal-hydraulic research program and to show how this program ensures maintenance of a viable, robust research effort and retention of needed expertise and international leadership

  8. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  9. Crystallization of nuclear glass under a thermal gradient: application to the self-crucible produced in the skull melting process

    International Nuclear Information System (INIS)

    Delattre, O.

    2013-01-01

    In the context of the vitrification of high level nuclear waste, a new industrial process has been launched in 2010 at the La Hague factory: The skull melting process. This setup applies thermal gradients to the melt, which leads to the formation of a solid layer of glass: the 'self-crucible'. The question would be to know whether these thermal gradients have an impact or not on the crystallization behaviour of the considered glasses in the self crucible. In order to answer that question, the crystallization of two glass compositions of nuclear interest has been investigated with an image analysis based method in isothermal and thermal gradient heat treatments conditions. The isothermal experiments allow for the quantification (growth speed, nucleation, crystallized fraction) of the crystallization of apatites (660 C-900 C) and powellites (630 C-900 C). The comparison of the results obtained through these two types of experimentations allows us to conclude that there is no impact of the thermal gradient on the crystallization of the studied glass compositions. In order to complete the image analysis study (based on surfaces), in and ex situ microtomography experiments have been performed at ESRF (Grenoble) on the ID19 beamline. This study allowed us to follow the crystallization of apatites in a simplified glass and to confirm the reliability of the image analysis method based on the analysis of surfaces. (author) [fr

  10. Conceptual Engine System Design for NERVA derived 66.7KN and 111.2KN Thrust Nuclear Thermal Rockets

    International Nuclear Information System (INIS)

    Fittje, James E.; Buehrle, Robert J.

    2006-01-01

    The Nuclear Thermal Rocket concept is being evaluated as an advanced propulsion concept for missions to the moon and Mars. A tremendous effort was undertaken during the 1960's and 1970's to develop and test NERVA derived Nuclear Thermal Rockets in the 111.2 KN to 1112 KN pound thrust class. NASA GRC is leveraging this past NTR investment in their vehicle concepts and mission analysis studies, and has been evaluating NERVA derived engines in the 66.7 KN to the 111.2 KN thrust range. The liquid hydrogen propellant feed system, including the turbopumps, is an essential component of the overall operation of this system. The NASA GRC team is evaluating numerous propellant feed system designs with both single and twin turbopumps. The Nuclear Engine System Simulation code is being exercised to analyze thermodynamic cycle points for these selected concepts. This paper will present propellant feed system concepts and the corresponding thermodynamic cycle points for 66.7 KN and 111.2 KN thrust NTR engine systems. A pump out condition for a twin turbopump concept will also be evaluated, and the NESS code will be assessed against the Small Nuclear Rocket Engine preliminary thermodynamic data

  11. First records of thermal neutrons with the spectrometer for time of flight (TOF) in the RP-10 Nuclear Reactor

    International Nuclear Information System (INIS)

    Munive, M.; Baltuano, O; Soto, C; Ravello, Y

    2002-01-01

    To obtain the first spectrum of an emergent beam of neutrons of a nuclear reactor is the main parameter of the characterization in the use of this reactor; one of ways to get this spectrum is for the technique of time of flight, TOF, which registers the time that a neutron need to cover a certain distance, associating this time then to the kinetic energy of the neutron. The kinetic study of the beam of neutrons is carried out on neutron pulses that are generated by a revolving choke called Chopper; and the analysis in the time of the detected pulses is carried out for a system MCS. Using this technique it is achieved the record of the spectra in energy, or in wavelength , of the irradiation facilities No 2 and 4, and of the exit N o 5 of the thermal column of the Nuclear Reactor RP-10 of the Nuclear Center Oscar de la Guerra RACSO, Peru (au)

  12. A HYPOTHESIS FOR THE COLOR BIMODALITY OF JUPITER TROJANS

    International Nuclear Information System (INIS)

    Wong, Ian; Brown, Michael E.

    2016-01-01

    One of the most enigmatic and hitherto unexplained properties of Jupiter Trojans is their bimodal color distribution. This bimodality is indicative of two sub-populations within the Trojans, which have distinct size distributions. In this paper, we present a simple, plausible hypothesis for the origin and evolution of the two Trojan color sub-populations. In the framework of dynamical instability models of early solar system evolution, which suggest a common primordial progenitor population for both Trojans and Kuiper Belt objects, we use observational constraints to assert that the color bimodalities evident in both minor body populations developed within the primordial population prior to the onset of instability. We show that, beginning with an initial composition of rock and ices, location-dependent volatile loss through sublimation in this primordial population could have led to sharp changes in the surface composition with heliocentric distance. We propose that the depletion or retention of H 2 S ice on the surface of these objects was the key factor in creating an initial color bimodality. Objects that retained H 2 S on their surfaces developed characteristically redder colors upon irradiation than those that did not. After the bodies from the primordial population were scattered and emplaced into their current positions, they preserved this primordial color bimodality to the present day. We explore predictions of the volatile loss model—in particular, the effect of collisions within the Trojan population on the size distributions of the two sub-populations—and propose further experimental and observational tests of our hypothesis.

  13. A HYPOTHESIS FOR THE COLOR BIMODALITY OF JUPITER TROJANS

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Ian; Brown, Michael E., E-mail: iwong@caltech.edu [Division of Geological and Planetary Sciences, California Institute of Technology, Pasadena, CA 91125 (United States)

    2016-10-01

    One of the most enigmatic and hitherto unexplained properties of Jupiter Trojans is their bimodal color distribution. This bimodality is indicative of two sub-populations within the Trojans, which have distinct size distributions. In this paper, we present a simple, plausible hypothesis for the origin and evolution of the two Trojan color sub-populations. In the framework of dynamical instability models of early solar system evolution, which suggest a common primordial progenitor population for both Trojans and Kuiper Belt objects, we use observational constraints to assert that the color bimodalities evident in both minor body populations developed within the primordial population prior to the onset of instability. We show that, beginning with an initial composition of rock and ices, location-dependent volatile loss through sublimation in this primordial population could have led to sharp changes in the surface composition with heliocentric distance. We propose that the depletion or retention of H{sub 2}S ice on the surface of these objects was the key factor in creating an initial color bimodality. Objects that retained H{sub 2}S on their surfaces developed characteristically redder colors upon irradiation than those that did not. After the bodies from the primordial population were scattered and emplaced into their current positions, they preserved this primordial color bimodality to the present day. We explore predictions of the volatile loss model—in particular, the effect of collisions within the Trojan population on the size distributions of the two sub-populations—and propose further experimental and observational tests of our hypothesis.

  14. Thermal-Hydraulic Analysis of the Nuclear Power Engineering Corporation Containment Experiments with GOTHIC

    International Nuclear Information System (INIS)

    Wiles, Lawrence E.; George, Thomas L.

    2003-01-01

    GOTHIC version 7.0 was used to model five tests that were conducted in the Nuclear Power Engineering Corporation facility in Japan. The tests involved steam and helium injection into a preheated, spray-moderated, 1/4-scale model of a pressurized water reactor dry containment. Comparison of GOTHIC predictions to measured data for pressure, vapor temperatures, structure surface temperatures, and helium concentrations provided the opportunity to evaluate methods for modeling gas dispersion, drop heat and mass transfer, and surface heat transfer.The test facility includes three floors. The lower two floors are partitioned into a variety of rooms that simulate the lower regions of the modeled containment. On the upper floor, rooms that simulate the steam generator enclosures and the pressurizer enclosure extend into the dome, which represents about two-thirds of the total volume of the containment.The GOTHIC model was defined with 30 control volumes using a mix of lumped parameter volumes and subdivided volumes that employ a three-dimensional mesh. Each volume included several thermal conductors to model the various structures. More than 100 flow paths were used to model the hydraulic connections.Comparison of predictions to data showed that enhanced grid resolution in the vicinity of the steam-helium release point served to limit dispersion of the steam-helium plume. The data comparisons also suggested that spray effectiveness was reduced by drop impact with the containment wall and by the high drop concentration. The data comparisons further suggested that the presence of condensation, sprays, splashing, and other wetting mechanisms should be considered to obtain a reasonable estimate of the effect of liquid films on the structure surfaces

  15. Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

    Energy Technology Data Exchange (ETDEWEB)

    Al-Jobouri, Hussain A., E-mail: hahmed54@gmail.com; Rajab, Mustafa Y., E-mail: mostafaheete@gmail.com [Department of Physics, College of Science, AL-Nahrain University, Baghdad (Iraq)

    2016-03-25

    CR-39 detector which covered with boric acid (H{sub 3}Bo{sub 3}) pellet was irradiated by thermal neutrons from ({sup 241}Am - {sup 9}Be) source with activity 12Ci and neutron flux 10{sup 5} n. cm{sup −2}. s{sup −1}. The irradiation times -T{sub D} for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C° temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLAB software were analyzed and found the following relationships: (a) The irradiation time -T{sub D} has behavior linear relationships with following nuclear track parameters: i) total track number - N{sub T} ii) maximum track number - MRD (relative to track diameter - D{sub T}) at response region range 2.5 µm to 4 µm iii) maximum track number - M{sub D} (without depending on track diameter - D{sub T}). (b) The irradiation time -T{sub D} has behavior logarithmic relationship with maximum track number - M{sub A} (without depending on track area - A{sub T}). The image processing technique principally track diameter - D{sub T} can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.

  16. Thermal stability of the French nuclear waste glass - long term behavior modeling

    International Nuclear Information System (INIS)

    Orlhac, X.

    2000-01-01

    The thermal stability of the French nuclear waste glass was investigated experimentally and by modeling to predict its long-term evolution at low temperature. The crystallization mechanisms were analyzed by studying devitrification in the supercooled liquid. Three main crystalline phases were characterized (CaMoO 4 , CeCO 2 , ZnCr 2 O 4 ). Their crystallisation was TO 4.24 wt%, due to the low concentration of the constituent elements. The nucleation and growth curves showed that platinoid elements catalysed nucleation but did not affect growth, which was governed by volume diffusion. The criteria of classic nucleation theory were applied to determine the thermodynamic and diffusional activation energies. Viscosity measurements illustrate the analogy between the activation energy of viscous flow and diffusion, indicating control of crystallization by viscous flow phenomena. The combined action of nucleation and growth was assessed by TTT plots, revealing a crystallization equilibrium line that enables the crystallized fractions to be predicted over the long term. The authors show that hetero-genetics catalyze the transformation without modifying the maximum crystallized fraction. A kinetic model was developed to describe devitrification in the glass based on the nucleation and growth curves alone. The authors show that the low-temperature growth exhibits scale behavior (between time and temperature) similar to thermo-rheological simplicity. The analogy between the resulting activation energy and that of the viscosity was used to model growth on the basis of viscosity. After validation with a simplified (BaO 2 SiO 2 ) glass, the model was applied to the containment glass. The result indicated that the glass remained completely vitreous after a cooling scenario with the one measured at the glass core. Under isothermal conditions, several million years would be required to reach the maximum theoretical crystallization fraction. (author)

  17. Simulation of thermal behavior of nuclear fuel rod by electrically heated pin

    International Nuclear Information System (INIS)

    Carajilescov, P.

    1985-01-01

    The utilization of electrically heated rods for the simulation of nuclear fuel rods represents an universally adopted method by the nuclear industry to study thermalhydraulic problems. The present work represents the development of a method to obtain the time variation of the electric linear power necessary to simulate a given nuclear power transient in order to yield the same temperature and heat flux conditions in the surface of the electrical heater that would be obtained by the nuclear fuel rod. (Author) [pt

  18. Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yoo, Seung Hun

    2011-02-01

    The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

  19. Results From an International Simulation Study on Coupled Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

    International Nuclear Information System (INIS)

    J. Rutqvist; D. Barr; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q. Liu; Y. Oda; W. Wang; C. Zhang

    2006-01-01

    As part of the ongoing international DECOVALEX project, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near waste emplacement drifts of geological nuclear waste repositories. The simulations were conducted for two generic repository types, one with open and the other with back-filled repository drifts, under higher and lower postclosure temperatures, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses is currently being resolved. In particular, good agreement in the basic thermal-mechanical responses was achieved for both repository types, even though some teams used relatively simplified thermal-elastic heat-conduction models that neglected complex near-field thermal-hydrological processes. The good agreement between the complex and simplified process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level

  20. Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

    2009-01-01

    R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

  1. Thermal properties of clay-based buffer materials for a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    Radhakrishna, H.S.

    1984-06-01

    The thermal properties of three types of bentonite clay, one illite-rich shale and one kaolin mixed with crushed granite were investigated. Thermal conductivity measurements were made over a range of mix proportions, moisture content, density and ambient temperature using the transient heat-probe method. The effects of thermal drying in the buffer zone prior to water uptake were investigated by means of laboratory-scale heater experiments. Illite-rich shale (Sealbond) and kaolin exhibited better compactability and thermal conductivity than the bentonite clays. The thermal conductivity of all types of clay buffers showed a high degree of moisture dependency and relatively no effect due to elevated temperature under high fluid pressure conditions. Bentonite buffers compacted to a dry density of 1200 to 1400 kg/m 3 showed extensive cracking due to differential shrinkage. Addition of crushed granite, and/or compaction to a higher density, reduced the thermal cracking of the buffer material

  2. Effects of burn-off on thermal shock resistances of nuclear carbon materials

    International Nuclear Information System (INIS)

    Kurumada, A.; Oku, T.; Kawamata, K.; Hiraoka, T.; McEnaney, B.

    1997-01-01

    The purpose of this study is to evaluate the effects of burn-off on thermal shock resistance and thermal shock fracture toughness of a carbon fiber reinforced carbon composite (C/C composite) and three fine-grained isotropic graphites. The thermal shock resistance and thermal shock fracture toughness was degraded slightly by air oxidation at 500 o C. The extent of degradations of the thermal shock parameters were less than those of the mechanical and fracture mechanics properties, however, they were larger than that of the thermal diffusivity. In observations of the microstructures of the fracture surfaces after oxidation of the graphites, the size and the number of pores were increased and the fracture surfaces were rough due to oxidation of boundaries of graphite particles. After oxidation of the C/C composite, there were preferential removal of the boundary layer between carbon fiber and pyrolytic carbon matrix and pull out of carbon fiber. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  4. Reactivity impact of {sup 16}O thermal elastic-scattering nuclear data for some numerical and critical benchmark systems

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K. S.; Roubtsov, D. [AECL, Chalk River Laboratories, Chalk River, ON (Canada); Plompen, A. J. M.; Kopecky, S. [EC-JRC, Inst. for Reference Materials and Measurements, Retieseweg 111, 2440 Geel (Belgium)

    2012-07-01

    The thermal neutron-elastic-scattering cross-section data for {sup 16}O used in various modern evaluated-nuclear-data libraries were reviewed and found to be generally too high compared with the best available experimental measurements. Some of the proposed revisions to the ENDF/B-VII.0 {sup 16}O data library and recent results from the TENDL system increase this discrepancy further. The reactivity impact of revising the {sup 16}O data downward to be consistent with the best measurements was tested using the JENDL-3.3 {sup 16}O cross-section values and was found to be very small in MCNP5 simulations of the UO{sub 2} and reactor-recycle MOX-fuel cases of the ANS Doppler-defect numerical benchmark. However, large reactivity differences of up to about 14 mk (1400 pcm) were observed using {sup 16}O data files from several evaluated-nuclear-data libraries in MCNP5 simulations of the Los Alamos National Laboratory HEU heavy-water solution thermal critical experiments, which were performed in the 1950's. The latter result suggests that new measurements using HEU in a heavy-water-moderated critical facility, such as the ZED-2 zero-power reactor at the Chalk River Laboratories, might help to resolve the discrepancy between the {sup 16}O thermal elastic-scattering cross-section values and thereby reduce or better define its uncertainty, although additional assessment work would be needed to confirm this. (authors)

  5. Measurement and analysis of temperature, strain and stress of foundation mat concrete in nuclear and thermal power stations

    International Nuclear Information System (INIS)

    Haraguchi, Akira; Yamakawa, Hidetsugu; Abe, Hirotoshi

    1981-01-01

    The problems of the thermal stress in concrete structures are roughly divided into the initial stress due to setting heat and the stress due to external temperature after hardening. The initial stress exists in every concrete structure, and it is usually neglected in beams and columns, but it must be taken into account in case of the foundation mat structures in nuclear power stations, for example. In this paper, (1) the results of measurement of temperature, strain and stress in each lift at the time of and after placing concrete in the foundation mat of a nuclear power station and the comparison of them with the results of analysis, (2) the results of measurement of the temperature and stress in a foundation mat, which was carried out to rationalize the design method for the raft type foundation mats in thermal power stations, and (3) the results of examination on the analysis model, external force conditions and boundary conditions used for the design are reported. The analysis method for temperature and thermal stress by finite element method, developed by the Central Research Institute of Electric Power Industry, can take the changes in the heat of hydration in placed concrete, the creep phenomenon of concrete and the restraint at construction joints in consideration. It is necessary to collect the data on the measurement of mat concrete and to develop the accurate analysis method. (Kako, I.)

  6. Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

    International Nuclear Information System (INIS)

    Brown, R.; Cobanoglu, M.M.

    1995-01-01

    Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

  7. The possibility of prediction of the lifetime of metallic nuclear fuel elements in a radiation field of thermal nuclear reactors

    International Nuclear Information System (INIS)

    Livne, Z.; Ramon, P.

    1979-01-01

    An attempt is made to clarify the possible causes of failure of irradiated nuclear fuel cartridges, in order to determine the parameters which govern the lifetime of the fuel and a way to predict it. Measurements of mechanical properties of irradiated uranium metal and cladding, can serve as a basis for failure prediction. Testing irradiated fuel elements by bending till fracture enables to evaluate the integral character of the fuel element, along the cross-section, taking into account the difference in brittleness of several zones. It is likely that the bending test, which indicates the behaviour of a stress-strain function, is a faster and more reliable way to determine the mechanical properties of irradiated nuclear fuel. Since the stresses applied to the cladding during irradiation are locally hydrostatic, its postirradiation blow-up provide information on strength and elasticity variations of the irradiated cladding material. (B.G.)

  8. Rapid intensification and the bimodal distribution of tropical cyclone intensity.

    Science.gov (United States)

    Lee, Chia-Ying; Tippett, Michael K; Sobel, Adam H; Camargo, Suzana J

    2016-02-03

    The severity of a tropical cyclone (TC) is often summarized by its lifetime maximum intensity (LMI), and the climatological LMI distribution is a fundamental feature of the climate system. The distinctive bimodality of the LMI distribution means that major storms (LMI >96 kt) are not very rare compared with less intense storms. Rapid intensification (RI) is the dramatic strengthening of a TC in a short time, and is notoriously difficult to forecast or simulate. Here we show that the bimodality of the LMI distribution reflects two types of storms: those that undergo RI during their lifetime (RI storms) and those that do not (non-RI storms). The vast majority (79%) of major storms are RI storms. Few non-RI storms (6%) become major storms. While the importance of RI has been recognized in weather forecasting, our results demonstrate that RI also plays a crucial role in the TC climatology.

  9. The Development of Bimodal Bilingualism: Implications for Linguistic Theory.

    Science.gov (United States)

    Lillo-Martin, Diane; de Quadros, Ronice Müller; Pichler, Deborah Chen

    2016-01-01

    A wide range of linguistic phenomena contribute to our understanding of the architecture of the human linguistic system. In this paper we present a proposal dubbed Language Synthesis to capture bilingual phenomena including code-switching and 'transfer' as automatic consequences of the addition of a second language, using basic concepts of Minimalism and Distributed Morphology. Bimodal bilinguals, who use a sign language and a spoken language, provide a new type of evidence regarding possible bilingual phenomena, namely code-blending, the simultaneous production of (aspects of) a message in both speech and sign. We argue that code-blending also follows naturally once a second articulatory interface is added to the model. Several different types of code-blending are discussed in connection to the predictions of the Synthesis model. Our primary data come from children developing as bimodal bilinguals, but our proposal is intended to capture a wide range of bilingual effects across any language pair.

  10. A novel broadband bi-mode active frequency selective surface

    Science.gov (United States)

    Xu, Yang; Gao, Jinsong; Xu, Nianxi; Shan, Dongzhi; Song, Naitao

    2017-05-01

    A novel broadband bi-mode active frequency selective surface (AFSS) is presented in this paper. The proposed structure is composed of a periodic array of convoluted square patches and Jerusalem Crosses. According to simulation results, the frequency response of AFSS definitely exhibits a mode switch feature between band-pass and band-stop modes when the diodes stay in ON and OFF states. In order to apply a uniform bias to each PIN diode, an ingenious biasing network based on the extension of Wheatstone bridge is adopted in prototype AFSS. The test results are in good agreement with the simulation results. A further physical mechanism of the bi-mode AFSS is shown by contrasting the distribution of electric field on the AFSS patterns for the two working states.

  11. A novel broadband bi-mode active frequency selective surface

    Directory of Open Access Journals (Sweden)

    Yang Xu

    2017-05-01

    Full Text Available A novel broadband bi-mode active frequency selective surface (AFSS is presented in this paper. The proposed structure is composed of a periodic array of convoluted square patches and Jerusalem Crosses. According to simulation results, the frequency response of AFSS definitely exhibits a mode switch feature between band-pass and band-stop modes when the diodes stay in ON and OFF states. In order to apply a uniform bias to each PIN diode, an ingenious biasing network based on the extension of Wheatstone bridge is adopted in prototype AFSS. The test results are in good agreement with the simulation results. A further physical mechanism of the bi-mode AFSS is shown by contrasting the distribution of electric field on the AFSS patterns for the two working states.

  12. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, 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, and 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 papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 2, 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, and 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 papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    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.

  15. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, 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, and 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 papers are indexed separately for inclusion in the Energy Science and Technology Database

  16. Mobile Education: Towards Affective Bi-modal Interaction for Adaptivity

    Directory of Open Access Journals (Sweden)

    Efthymios Alepis

    2009-04-01

    Full Text Available One important field where mobile technology can make significant contributions is education. However one criticism in mobile education is that students receive impersonal teaching. Affective computing may give a solution to this problem. In this paper we describe an affective bi-modal educational system for mobile devices. In our research we describe a novel approach of combining information from two modalities namely the keyboard and the microphone through a multi-criteria decision making theory.

  17. 'Bi-modal' isoscalar giant dipole strength in 58Ni

    International Nuclear Information System (INIS)

    Nayak, B.K.; Garg, U.; Hedden, M.; Koss, M.; Li, T.; Liu, Y.; Madhusudhana Rao, P.V.; Zhu, S.; Itoh, M.; Sakaguchi, H.; Takeda, H.; Uchida, M.; Yasuda, Y.; Yosoi, M.; Fujimura, H.; Fujiwara, M.; Hara, K.; Kawabata, T.; Akimune, H.; Harakeh, M.N.

    2006-01-01

    The strength distribution of the isoscalar giant dipole resonance (ISGDR) in 58 Ni has been obtained over the energy range 10.5-49.5 MeV via extreme forward angle scattering (including 0 deg.) of 386 MeV α particles. We observe a 'bi-modal' E1 strength distribution for the first time in an A<90 nucleus. The observed ISGDR strength distribution is in reasonable agreement with the predictions of a recent RPA calculation

  18. Multifractal Characteristics of Bimodal Mercury Pore Size Distribution Curves

    Science.gov (United States)

    dos Santos Bonini, C.; Alves, M. C.; Paz González, A.

    2012-04-01

    Characterization of Hg pore size distribution (PSDs) curves by monofractal or multifractal analysis has been demonstrated to be an useful tool, which allows a better understanding of the organization of the soil pore space. There are also evidences that multiscale analysis of different segments found in bimodal pore size distributions measured by Hg intrusion can provide further valuable information. In this study we selected bimodal PSDs from samples taken from an experimental area in São Paulo state, Brazil, where a revegetation trial was set up over saprolitic material. The saprolite was left abandoned after decapitation of an Oxisol for building purposes. The field trial consisted of various treatments with different grass species and amendments. Pore size distribution of the sampled aggregates was measured in the equivalent diameter range from 0.005 to about 50 μm and it was characterized by a bimodal pattern, so that two compartments, i.e. 0.005 to 0.2 μm and 0.2 to 50 μm, could be distinguished. The multifractal theory was used to analyse both segments. The scaling properties of these two segments could be fitted reasonably well with multifractal models. Multifractal parameters obtained for equivalent diameters for the segments > 0.2 and pore size distributions studied.

  19. On the effect of segregation on intense bimodal bed load

    Directory of Open Access Journals (Sweden)

    Zrostlík Štěpán

    2017-01-01

    Full Text Available Open-channel two-phase flow above a granular mobile bed is studied experimentally and theoretically. In the two-phase flow, water serves as a carrying liquid for plastic grains transported as collisional contact load in the upper-stage plane bed regime. The investigation evaluates friction- and transport characteristics of the flow under the condition of intense collisional transport of grains and links them with the internal structure of the two-phase flow. The paper focusses on the effect of bimodal solids (mixed two fractions of grains of similar density and different size and shape on the flow characteristics and internal structure. Hence, experimental results obtained for the bimodal mixture are compared with results for individual grain fractions. The experiments show that the bimodal character of the transported solids affects the layered internal structure of the flow as a result of fraction segregation due primarily to gravity (kinetic sieving during transport. The segregation also affects the friction- and transport characteristics of intense bed load. In the paper, the effects are described and quantified.

  20. On the effect of segregation on intense bimodal bed load

    Science.gov (United States)

    Zrostlík, Štěpán; Matoušek, Václav

    Open-channel two-phase flow above a granular mobile bed is studied experimentally and theoretically. In the two-phase flow, water serves as a carrying liquid for plastic grains transported as collisional contact load in the upper-stage plane bed regime. The investigation evaluates friction- and transport characteristics of the flow under the condition of intense collisional transport of grains and links them with the internal structure of the two-phase flow. The paper focusses on the effect of bimodal solids (mixed two fractions of grains of similar density and different size and shape) on the flow characteristics and internal structure. Hence, experimental results obtained for the bimodal mixture are compared with results for individual grain fractions. The experiments show that the bimodal character of the transported solids affects the layered internal structure of the flow as a result of fraction segregation due primarily to gravity (kinetic) sieving during transport. The segregation also affects the friction- and transport characteristics of intense bed load. In the paper, the effects are described and quantified.

  1. Optimization of thermal efficiency of nuclear central power like as PWR

    International Nuclear Information System (INIS)

    Lapa, Nelbia da Silva

    2005-10-01

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  2. A hybrid accident simulation methodology for nuclear power plant by combining thermal-hydraulic program and artificial neural networks

    International Nuclear Information System (INIS)

    Choi, Young Joon

    2004-02-01

    Compact simulators for nuclear power plants can be used as cost-effective training or analysis tools; generally, they demonstrate overall responses of transients or accidents in real time or faster. In the thermal-hydraulic models of compact simulators, governing equations are simplified with reasonable assumptions and empirical correlations, and approximate solutions are obtained by using appropriate numerical schemes. Moreover, many physical control volumes in plant modeling are lumped to reduce the computing time. The simplification of equations and reduction of control volume numbers usually degrade the accuracy of solutions. A hybrid accident simulation methodology is proposed to enhance the capabilities of a compact simulator by introducing artificial neural networks. A simplified thermal-hydraulic program, playing the role of compact simulator, is designed to calculate the overall responses of transients and accidents. Two neural networks are designed and trained with the target values obtained from the analyses of detailed computer codes and trained results are combined with the simplified thermal-hydraulic program to perform the following roles: (I) compensation for inaccuracy of a simplified thermal-hydraulic program occurring from simplified governing equation and small number of physical control volumes: the auto-associative neural network (AANN), trained with the target values obtained from RELAP5/MOD3 code analyses, improves the calculated results of the simplified thermal-hydraulic program, and (II) prediction of the critical parameter usually calculated from the sophisticated computer code: the back propagation neural network (BPN), trained with the target values obtained from COBRA-IV code analyses, predicts the minimum departure from nucleate boiling ratio (DNBR) which is not calculated in simplified thermal-hydraulic program. Simulations for the several accidents are carried out to verify the applicability of the proposed methodology. The

  3. Effect of U-238 and U-235 cross sections on nuclear characteristics of fast and thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1997-03-01

    Benchmark calculation has been made for fast and thermal reactors by using ENDF/B-VI release 2(ENDF/B-VI.2) and JENDL-3.2 nuclear data. Effective multiplication factors (k{sub eff}s) calculated for fast reactors calculated with ENDF/B-VI.2 becomes about 1% larger than the results with JENDL-3.2. The difference in k{sub eff} is caused mainly from the difference in inelastic scattering cross section of U-238. In all thermal benchmark cores, ENDF/B-VI.2 gives smaller multiplication factors than JENDL-3.2. In U-235 cores, the difference is about 0.3%dk and it becomes about 0.6% in TCA U cores. The difference in U-238 data is also important in thermal reactors, while there are found 0.1-0.3% different v values of U isotopes in thermal energy between ENDF/B-VI.2 and JENDL-3.2. (author)

  4. Nuclear power plant accident simulations of gasket materials under simultaneous radiation plus thermal plus mechanical stress conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gillen, K.T.; Malone, G.M.

    1997-07-01

    In order to probe the response of silicone door gasket materials to a postulated severe accident in an Italian nuclear power plant, compression stress relaxation (CSR) and compression set (CS) measurements were conducted under combined radiation (approximately 6 kGy/h) and temperature (up to 230{degrees}C) conditions. By making some reasonable initial assumptions, simplified constant temperature and dose rates were derived that should do a reasonable job of simulating the complex environments for worst-case severe events that combine overall aging plus accidents. Further simplification coupled with thermal-only experiments allowed us to derive thermal-only conditions that can be used to achieve CSR and CS responses similar to those expected from the combined environments that are more difficult to simulate. Although the thermal-only simulations should lead to sealing forces similar to those expected during a severe accident, modulus and density results indicate that significant differences in underlying chemistry are expected for the thermal-only and the combined environment simulations. 15 refs., 31 figs., 15 tabs.

  5. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  6. Prevention and treatment of the Farley-Tihange phenomenon of nuclear auxiliary pipes based on thermal fatigue

    International Nuclear Information System (INIS)

    Cao Feng; Wang Jianjun; Ding Youyuan

    2012-01-01

    Farley-Tihange Phenomenon due to thermal fatigue frequently appears on the downstream area filled with cool and heat water of the residual heat removal heat-exchange equipment and the base metal and welding joint of the RIS and RRA pipes connected with the primary coolant pipe directly in global Nuclear power plants in operation. Which brings unacceptable defects, even worse, LOCA. During the pre-service inspection, autonomic ultrasonic test and radiographic test were done to relative pipes and welds of Unit 3 of the Qinshan Nuclear Power Phase Ⅱ. This article summarizes the recurrent position and potential risks of the Farley-Tihange phenomenon, establishes the fault tree of its failure causes, analyses failure mechanism and models of heat fatigue, presents systematically prevention and treatment methods including in-operation supervision, shut-down inspections, emergent maintenance program and so on. (authors)

  7. Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

    International Nuclear Information System (INIS)

    Sizgek, G. Devlet

    2005-01-01

    A thermal model is constructed and analyses are performed for an 'in-floor' type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels

  8. Determination of the fission coefficients in thermal nuclear reactors for antineutrino detection

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Lenilson M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Cabral, Ronaldo G., E-mail: rgcabral@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Anjos, Joao C.C. dos, E-mail: janjos@cbpf.b [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil). Dept. GLN - G

    2011-07-01

    The nuclear reactors in operation periodically need to change their fuel. It is during this process that these reactors are more vulnerable to occurring of several situations of fuel diversion, thus the monitoring of the nuclear installations is indispensable to avoid events of this nature. Considering this fact, the most promissory technique to be used for the nuclear safeguard for the nonproliferation of nuclear weapons, it is based on the detection and spectroscopy of antineutrino from fissions that occur in the nuclear reactors. The detection and spectroscopy of antineutrino, they both depend on the single contribution for the total number of fission of each actinide in the core reactor, these contributions receive the name of fission coefficients. The goal of this research is to show the computational and mathematical modeling used to determinate these coefficients for PWR reactors. (author)

  9. Nitrogen determination in wheat by neutron activation analysis using fast neutron flux from a thermal nuclear reactor

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

    This is a study of the technique for the determination of nitrogen and other elements in wheat flour through activation analysis with fast neutrons from a thermal nuclear reactor. The study begins with an introduction about the basis of the analytical methods, the equipment used in activation analysis and a brief description of the neutrons source. In the study are included the experiments carried out in order to determine the flux form in the site of irradiation, the N-13 half life and the interference due to the sample composition. (author)

  10. The Impact of Climate Changes on the Thermal Performance of a Proposed Pressurized Water Reactor: Nuclear-Power Plant

    OpenAIRE

    Said M. A. Ibrahim; Mohamed M. A. Ibrahim; Sami. I. Attia

    2014-01-01

    This paper presents a methodology for studying the impact of the cooling water temperature on the thermal performance of a proposed pressurized water reactor nuclear power plant (PWR NPP) through the thermodynamic analysis based on the thermodynamic laws to gain some new aspects into the plant performance. The main findings of this study are that an increase of one degree Celsius in temperature of the coolant extracted from environment is forecasted to decrease by 0.39293 and 0.16% in the pow...

  11. On the radiation safety studies of space nuclear sources at the Scientific-Research Institute of Thermal Processes

    Energy Technology Data Exchange (ETDEWEB)

    Koroteev, A.S.; Gafarov, A.A.; Bakhtin, B.I.; Kosov, A.V. (The Scientific-Research Institute of Thermal Processes, 8, Onezhskaya, Moscow, 125438 (SU))

    1991-01-01

    The main directions and some results of the theoretical and experimental studies, carried out at the Scientific-Research Institute of Thermal Processes on the radiation safety (RS) problem of the space nuclear power sources (SNPS), are stated. The experimental base for SNPS aerodynamic heating and breakup research is described. Some results of studies on the development of RS system for SNPS of Cosmos-954''-type satellites and SNPS Topaz'' are represented. The main directions and results of research on the RS problem of radioisotope SNPS is showen. Some aspects of SNPS possible collisions with space debris in near-earth orbits is examined.

  12. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Basu, Dipankar N.; Bhattacharyya, Souvik; Das, P.K.

    2014-01-01

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  13. Introduction on the thermal-hydraulic analysis codes for nuclear steam generator

    International Nuclear Information System (INIS)

    Yao Yangui; Zu Hongbiao; Yao Weida

    2015-01-01

    This paper describes several typical steam generator thermal-hydraulic analysis programs. Three thermal hydraulic code analysis model, principles and functions of the application are introduced. GENF is a one-dimensional code for steady state analysis, and another one-dimensional code TRANFLOW is used for transients, while ATHOS is a three-dimensional code which can be used to deal with steady as wall as transient analysis. And the code test verification and actual operating parameters verify situations are described. At last, the status and development of SG thermal-hydraulic analysis codes in China are analyzed. (authors)

  14. Thermal Aging Performance of Domestic Cast Austenitic Stainless Steels in Nuclear Power Plants

    Science.gov (United States)

    Chengliang, Li; Xiaoyun, Deng; Zhiying, Yin; Yuangang, Duan

    The domestic cast austenitic stainless steels (CASS) used for PWR main coolant lines face thermal aging phenomenon during long-term service at reactor operating temperatures in the range of 280-320°C. The accelerated thermal aging experiment of CASS at an elevated temperature of 400°C was carried out. The decrease regularity of absorbed energy of CASS at room temperature was obtained by Charpy impact test, and the variation trend of impacted specimen's fracture surfaces of CASS at different aging time was observed by scanning electron microscope. The experimental results show that: after 15,000 hours of accelerated thermal aging, the domestic CASS gradually reaches a thermal aging saturation, and its absorbed energy suffers a dramatic loss, dropping to almost 41-45% of the initial value, but still meets the assessment indicators for unaged state required by design specifications.

  15. Experimental Investigation of Latent Heat Thermal Energy Storage for Bi-Modal Solar Thermal Propulsion

    Science.gov (United States)

    2014-06-01

    be equivalent to the hydrogen exit temperature. Graph reprinted from Frye and Kudija34 The ISUS design provides steady thrust performance, but the...Germanium from 3 K to the Melting Point,” Physical Review , Vol. 134, No. 4A, May 1964, pp. A1058–A1069. 14Shoji, J. M., Frye , P. E., and McClanahan, J

  16. Nuclear Thermal Propulsion (NTP): A Proven, Growth Technology for Fast Transit Human Missions to Mars

    Science.gov (United States)

    Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

    2014-01-01

    The "fast conjunction" long surface stay mission option was selected for NASA's recent Mars Design Reference Architecture (DRA) 5.0 study because it provided adequate time at Mars (approx. 540 days) for the crew to explore the planet's geological diversity while also reducing the "1-way" transit times to and from Mars to approx. 6 months. Short transit times are desirable in order to reduce the debilitating physiological effects on the human body that can result from prolonged exposure to the zero-gravity (0-gE) and radiation environments of space. Recent measurements from the RAD detector attached to the Curiosity rover indicate that astronauts would receive a radiation dose of approx. 0.66 Sv (approx. 66 rem)-the limiting value established by NASA-during their 1-year journey in deep space. Proven nuclear thermal rocket (NTR) technology, with its high thrust and high specific impulse (Isp approx. 900 s), can cut 1-way transit times by as much as 50 percent by increasing the propellant capacity of the Mars transfer vehicle (MTV). No large technology scale-ups in engine size are required for these short transit missions either since the smallest engine tested during the Rover program-the 25 klbf "Pewee" engine is sufficient when used in a clustered arrangement of three to four engines. The "Copernicus" crewed MTV developed for DRA 5.0 is a 0-gE design consisting of three basic components: (1) the NTP stage (NTPS); (2) the crewed payload element; and (3) an integrated "saddle truss" and LH2 propellant drop tank assembly that connects the two elements. With a propellant capacity of approx. 190 t, Copernicus can support 1-way transit times ranging from approx. 150 to 220 days over the 15-year synodic cycle. The paper examines the impact on vehicle design of decreasing transit times for the 2033 mission opportunity. With a fourth "upgraded" SLS/HLV launch, an "in-line" LH2 tank element can be added to Copernicus allowing 1-way transit times of 130 days. To achieve 100

  17. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-08-01

    The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.

  18. Effluent Scrubbing of Engine Exhaust of a Nuclear Thermal Propulsion Engine

    Data.gov (United States)

    National Aeronautics and Space Administration — The Institute for Clean Energy Technology (ICET) at MSU already supports long-standing high-efficiency nuclear particulate air (HEPA) filtration research as well as...

  19. Water/rock interactions and mass transport within a thermal gradient Application to the confinement of high level nuclear waste

    International Nuclear Information System (INIS)

    Poinssot, Ch.; Ecole Normale Superieure, 92 - Fontenay-aux-Roses

    1998-01-01

    The initial stage of a high level nuclear waste disposal will be characterised by a large heat release within the near-field environment of the canisters. This heat flux caused by radioactive decay will lead to an increase of temperature and a subsequent thermal gradient between the 'hot' canisters and the 'cold'geological medium. In addition, this thermal gradient will decrease with time due to the heat decay although it could last hundred years. What will be the consequences of such a thermal field varying both on space and time for the alteration of the different constituents of the near field environment. In particular, what could be the effects on the radionuclides migration in the accidental case of an early breach of a canister during the thermal stage? This study brings significant answers to these questions in the light of a performance assessment study. This work is supported by a triple methodological approach involving experimental studies, modelling calculations and a natural analogues study. This complete work demonstrates that a thermal gradient leads to a large re-distribution of elements within the system: some elements are incorporated in the solid phases of the hot end (Si, Zr, Ca) whereas some others are in those of the cold end (Fe, Al, Zn). The confrontation of the results of very simple experiments with the results of a model built on equilibrium thermodynamics allow us to evidence the probable mechanisms causing this mass transport: out-of-equilibrium thermodiffusion processes coupled to irreversible precipitation. Moreover, the effects of the variation of temperatures with time is studied by the way of a natural system which underwent a similar temperature evolution as a disposal and which was initially rich in uranium: the Jurassic Alpine bauxites. In addition, part of the initial bauxite escaped this temperature transformations due to their incorporation in outer thrusting nappes. They are used as a reference. (author)

  20. STEADY-SHIP: a computer code for three-dimensional nuclear and thermal-hydraulic analyses of marine reactors

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Naito, Yoshitaka; Tokuno, Yukio; Matsui, Yasushi.

    1988-01-01

    A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

  1. Experimental evaluation of different mixing promoter for nuclear fuel element by means of a new thermal tracing technique

    International Nuclear Information System (INIS)

    Silin, Nicolas; Juanico, Luis; Delmastro, Dario

    2004-01-01

    In this work a new experimental method is used to experimentally evaluate the performance of different appendages promoting the turbulent mixing between the coupled subchannels of nuclear fuel elements.The method used will be introduced in another presentation and consists in the generation and measurement of small thermal traces in the refrigerating water flow between the fuel rods.Because it is suitable for heterogeneous and compact subchannels (as Argentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement.The experiments carried out on a small test section proved that the buttons brazed to the fuel rods (similar to the 'turbulence promoters' of the Canflex fuel) had an excellent thermohydraulic performance as compared to different mixing vane designs studied.The thermal traces method developed has shown its potential as a thermohydraulic design tool for the development of advanced nuclear fuels, that eventually incorporate mixing promoter elements. In the case of CARA, and as it includes spacer grids, it could be possible to use them to incorporate these elements without the need of brazing them to the rods (as is the case in Canflex), and therefore without penalizing its integrity [es

  2. Long-Term Preservation of Knowledge in the Area of Nuclear Thermal-Hydraulics - The New STRESA Database

    International Nuclear Information System (INIS)

    Pla, P.; Tanarro, J.; Ammirabile, L.; Strucic, M.; Wastin, F.

    2016-01-01

    The Joint Research Centre (JRC) carried out important projects in the area of thermal-hydraulics (TH) and severe accidents (SA). In the area of Integral Test Facilities (ITF) the JRC LOBI facility and its project (1970-1994) produced data of experiments simulating different accidents and transients in Pressurized Water Reactors (PWR). The JRC was engaged in relevant SA experimental projects: The FARO and KROTOS facilities (1991-2000) simulated Melt Fuel Coolant Interaction (MFCI) phenomena, considering either in-vessel and ex-vessel experiments and potential situations for steam explosions. The STORM facility simulated experiments in the area of aerosol transport. The projects produced large amount of experimental data important to understand the related thermal-hydraulic and severe accident phenomena. In the same way, experimental data became of essential use in the area of system/severe accident code assessment and for the development and improvement of analytical models included in the codes used in safety of Light Water Reactors (LWR). The STRESA (Storage of Thermal REactor Safety Analysis Data) database (DB) was developed by the JRC Ispra site in the year 2000 to store LOBI, FARO, KROTOS and STORM experimental data. Later on, the STRESA database was transferred to and maintained by JRC Petten site. The Nuclear Reactor Safety Assessment Unit (NRSA) of the JRC Petten was engaged during the past 2 years in the design and development of a new STRESA tool. The development of this new STRESA tool was completed and published on-line at the URL: http://stresa.jrc.ec.europa.eu/. The target was to keep the core features of the original STRESA structure but incorporating communication tools broadly used nowadays to align the capabilities of the new information system with the web 2.0. The relevance of issues related with the administration of information systems such as information security, user's data protection or scientific data management has increased in the

  3. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Science.gov (United States)

    Carcreff, H.; Salmon, L.; Lepeltier, V.; Guyot, J. M.; Bouard, E.

    2018-01-01

    Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the "zero method". Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the "zero method" measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between "zero" and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron

  4. Last Improvements of the CALMOS Calorimeter Dedicated to Thermal Neutron Flux and Nuclear Heating Measurements inside the OSIRIS Reactor

    Directory of Open Access Journals (Sweden)

    Carcreff H.

    2018-01-01

    Full Text Available Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the “zero method”. Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the “zero method” measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between “zero” and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions

  5. Status and perspectives of developing the fuel elements for nuclear power plants with thermal and fast reactors

    International Nuclear Information System (INIS)

    Reshetnikov, F.G.

    1981-01-01

    The attained level of fabrication of fuel and fuel elements for thermal and fast reactors is considered. A brief estimation of the fuel element quality is given. New requirements to fuel elements under development and the ways of increasing the thermal reactor fuel element operating life are discussed. The methods of the mixed uranium-plutonium oxide fuel fabrication and perspective carbide and metal fuels are considered in conformity with the fast reactors. The quality of steels, used while fabricating fuel elements and hexahedral fuel assembly shrouds is estimated. It is noted that for the last years new success is achieved in increasing the quality of fuel rods: fuel pellet density is increased and their spread in density is descrised, the content of such unwanted impurities, as fluorine and moisture is essentially reduced. The methods of fuel element sealing are reliably developed. The welding process is automated. Technological process and finished product quality control ensures fabrication of high-quality fuel elements. Premature fuel element tightness failure becomes an occasional phenomenon and the number of such fuel elements is reduced to 0.01%. Two principally different methods for fabricating the mixed fuel - codeposition and mechanical mixing, giving a rather uniform uranium and plutonium distribution in the fuel rod are developed and adopted. The conclusion is made that by common efforts of scientists and designers in collaboration with production workers modern mechanized and automated industrial production of fuel elements for thermal and fast reactors, meeting increasing requirements of the country, nuclear power engineering is created [ru

  6. Bimodal and Gaussian Ising spin glasses in dimension two

    Science.gov (United States)

    Lundow, P. H.; Campbell, I. A.

    2016-02-01

    An analysis is given of numerical simulation data to size L =128 on the archetype square lattice Ising spin glasses (ISGs) with bimodal (±J ) and Gaussian interaction distributions. It is well established that the ordering temperature of both models is zero. The Gaussian model has a nondegenerate ground state and thus a critical exponent η ≡0 , and a continuous distribution of energy levels. For the bimodal model, above a size-dependent crossover temperature T*(L ) there is a regime of effectively continuous energy levels; below T*(L ) there is a distinct regime dominated by the highly degenerate ground state plus an energy gap to the excited states. T*(L ) tends to zero at very large L , leaving only the effectively continuous regime in the thermodynamic limit. The simulation data on both models are analyzed with the conventional scaling variable t =T and with a scaling variable τb=T2/(1 +T2) suitable for zero-temperature transition ISGs, together with appropriate scaling expressions. The data for the temperature dependence of the reduced susceptibility χ (τb,L ) and second moment correlation length ξ (τb,L ) in the thermodynamic limit regime are extrapolated to the τb=0 critical limit. The Gaussian critical exponent estimates from the simulations, η =0 and ν =3.55 (5 ) , are in full agreement with the well-established values in the literature. The bimodal critical exponents, estimated from the thermodynamic limit regime analyses using the same extrapolation protocols as for the Gaussian model, are η =0.20 (2 ) and ν =4.8 (3 ) , distinctly different from the Gaussian critical exponents.

  7. Numerical Simulation of the Thermal Performance of a Dry Storage Cask for Spent Nuclear Fuel

    Directory of Open Access Journals (Sweden)

    Heui-Yung Chang

    2018-01-01

    Full Text Available In this study, the heat flow characteristics and thermal performance of a dry storage cask were investigated via thermal flow experiments and a computational fluid dynamics (CFD simulation. The results indicate that there are many inner circulations in the flow channel of the cask (the channel width is 10 cm. These circulations affect the channel airflow efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures at the top concrete lid and the upper locations of the concrete cask are higher than those permitted by the design specification. The installation of the salt particle collection device has a limited negative effect on the thermal dissipation performance of the dry storage cask.

  8. Thermal performance sensitivity studies in support of material modeling for extended storage of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fort, James A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-08-15

    The work reported here is an investigation of the sensitivity of component temperatures of a storage system, including fuel cladding temperatures, in response to age-related changes that could degrade the design-basis thermal behavior of the system. The purpose of these sensitivity studies is to provide a realistic example of how changes in the physical properties or configuration of the storage system components can affect temperatures and temperature distributions. The magnitudes of these sensitivities can provide guidance for identifying appropriate modeling assumptions for thermal evaluations extending long term storage out beyond 50, 100, 200, and 300 years.

  9. Small Low Mass Advanced PBR's for Bi-Modal Operation

    Science.gov (United States)

    Ludewig, Hans; Todosow, Michael; Powell, James R.

    1994-07-01

    A preliminary assessment is made of a low mass bi-modal reactor for use as a propulsion unit and as a heat source for generating electricity. This reactor is based on the particle bed reactor (PBR) concept. It will be able to generate both thrust and electricity simultaneously. This assessment indicates that the reactor can generate approximately 6.8 (4) N of thrust using hydrogen as a coolant, and 100 KWe using a closed Brayton cycle (CBC) power conversion system. Two cooling paths pass through the reactor allowing simultaneous operation of both modes. The development of all the components for this reactor are within the experience base of the NTP project.

  10. Bifurcation Structures in a Bimodal Piecewise Linear Map

    Directory of Open Access Journals (Sweden)

    Anastasiia Panchuk

    2017-05-01

    Full Text Available In this paper we present an overview of the results concerning dynamics of a piecewise linear bimodal map. The organizing principles of the bifurcation structures in both regular and chaotic domains of the parameter space of the map are discussed. In addition to the previously reported structures, a family of regions closely related to the so-called U-sequence is described. The boundaries of distinct regions belonging to these structures are obtained analytically using the skew tent map and the map replacement technique.

  11. Effect of meta-carborane on segmental dynamics in a bimodal Poly(dimethylsiloxane) network

    Energy Technology Data Exchange (ETDEWEB)

    Lewicki, J; Maxwell, R S; Patel, M; Herberg, J; Swain, A C; Liggat, J; Pethrick, R

    2008-06-11

    Bimodal networks of polydimethylsiloxane (PDMS) filled with varying amounts of icosahedral meta-carborane (m-CB) have been developed and characterized by broadband dielectric spectroscopy (BDS) and static {sup 1}H Multiple Quantum Nuclear Magnetic Resonance (MQ NMR). Both BDS and MQ NMR showed evidence for a decrease in the polymer chain dynamics. BDS spectra quantified a normal-mode relaxation near 40 Hz at 40 C. The frequency maximum observed for filled samples decreased with increasing m-CB content until contents greater than 5 wt. %. The width of the relaxation spectrum increased with the addition of small quantities of filler and decreased with filler contents greater that 5 wt. %. Agglomeration effects were observed at loadings greater than 5 wt % as manifest by the onset of low frequency Maxwell-Wagner-Sillars (MWS) processes. The MQ NMR data allowed the characterization of distributions of the residual dipolar couplings, <{Omega}{sub d}> and thus in the dynamic order parameter, Sb, consistent with the bimodal network architecture expected from the synthesis protocol used. Upon addition of less than 10 wt.% m-CB filler, the mean <{Omega}{sub d}> for the longer chains increased by 46% and the width of the distribution increased by 33%. The mean <{Omega}{sub d}> for the shorter chains increased by much less, indicative of preferential dispersion of the filler particles in the long chain domains of the network structure. We conclude that the mechanism of reinforcement is likely a free volume space filling at low loadings transitioning to complex molecular filler and polymer chain interaction phenomena at higher loadings.

  12. Feasibility study of the vectorization of nuclear codes used at the ENEL Thermal and Nuclear Research Centre

    International Nuclear Information System (INIS)

    Di Pasquantonio, F.

    1987-01-01

    The purpose of this report is that of analyzing tha problems connected with the vectorization and/or multitasking of several computer codes utilized in the ENEL-DSR Centro Ricerca Termica e Nucleare. After some general remarks on vector computers the analysis is focused on some topic relating to vectorization and multitasking of programs written for scalar computers. The priority for vectorization and/or multitasking has been given at the following codes: 1) DOT 4.2 (radiation transport and shielding); 2) QUANDRY (accidental and operating transients in LWR cores); 3) MORSE and MCNP (Monte Carlo codes for radiation transport and shielding); 4) RELAP (accidental and operating transients in LWR plants); 5) TRAP-MELT and NAUA (evaluation of source term). The principal results of the study are the following: 1) For the DOT 4.2 code it is convenients to improve the vectorized version DOT IV/C developed by Swanson introducing the parallel S.O.R. iterative method; 2) For the code QUANDRY it is proposed to introduce the three dimensional red-black mesh point ordering named ''diagonal method''; 3) To implements, on the CRAY X/MP 48, the multitasked version of the code MCNP developed by the Los Alamos National Lboratory; 4) To implements, on the CRAY X/MP 12, the vectorized and optimized version of the codes RELAP5/MOD1-MOD2 developed by J.R.C. EUROATOM-Ispra; 5) For the codes TRAP-MELT and NAUA the insertion of the vectorized routines LSODP na LSODPK for dominanting stiff cases

  13. Modelling of thermally driven groundwater flow in a facility for disposal of spent nuclear fuel in deep boreholes

    Energy Technology Data Exchange (ETDEWEB)

    Marsic, Nico; Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-09-15

    100 to 50 metres, the maximum temperature increase in the rock between the boreholes increased from 5 to 10 deg C and the duration of the this thermal pulse increased. Also, the thermally induced groundwater flow rate increased. However, the travel times for the groundwater from the disposal zone to the mobile fresh water zone above the halo cline remained much longer than the duration of the thermal pulse. Hence, the conclusion from previous studies that the thermal output from the fuel is insufficient to jeopardise the stability of the groundwater stratification is confirmed. It should be noted, though, that some mixing occurs at the halo cline if the permeability of the borehole filling material is assumed to increase. This mixing is less pronounced in the case with the sharper salinity interface. Based on the calculations performed in this study, it can be concluded that boreholes for disposal of spent nuclear fuel should not be spaced closer than 100 metres for the type of canisters assumed in this study. The results also indicated that the properties of the material used for backfilling the boreholes has some importance for the stability of the halo cline.

  14. Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

    Science.gov (United States)

    Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

    2009-12-01

    Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

  15. Deep geological disposal system development; thermal stress analysis and nonlinear structural analysis of spent nuclear fuel disposal canister under sudden rock movement

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Joo; Kim, Jin An; Ha, Jun Yong [Hongik University, Seoul (Korea)

    2002-04-01

    This work constitutes a summary of research and development made for design and dimensioning of the spent nuclear fuel disposal canister. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for the spent nuclear fuel disposal should be secured. Usually this repository is expected to locate at a depth of 500m underground. In this work the thermal stress analysis of the spent nuclear fuel disposal canister in a deep repository at 500m underground is performed for the underground pressure variation. Thermal stresses of the canister due to thermal loads of the heat generation of spent nuclear fuels inside baskets are computed. The thermal stress analysis result shows that even though some high thermal stresses occur due to the heat generation of nuclear fuels inside baskets, the canister is still structurally safe because the maximum stress occurred in the canister is smaller than the yield strength of the cast iron. In this work, the nonlinear structural analysis for the composite structure of the spent nuclear fuel disposal canister and the 50cm thick bentonite buffer is also carried out to predict the collapse of the canister while the sudden rock movement of 10cm is applied on the composite structure. Elastoplastic material model is adopted. Drucker-Prager yield criterion is used for the material yield prediction of the bentonite buffer and von-Mises yield criterion is used for the material yield prediction of the canister(cast iron insert, copper outer shell and lid and bottom). The analysis result shows that even though very large deformations occur beyond the yield point in the bentonite buffer, the canister structure still endures elastic small strains and stresses below the yield strength. Analysis results also show that bending deformations occur in the canister structure due to the shear deformation of the bentonite buffer. 24

  16. Development of a Nuclear Steam Supply System Thermal-Hydraulic Module for the Nuclear Power Plant Simulator Using a Best-Estimate Code, RETRAN

    International Nuclear Information System (INIS)

    Suh, Jae Seung

    2004-08-01

    The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually used very simplified physical models for the real-time simulation of Ness thermal-hydraulic transients, which entails 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, a realistic NSSS thermal-hydraulic program ARTS has been developed, it was based on the RETRAN code for the improvement of the Nuclear Power Plant full-scope simulator. Since ARTS is a generalized code to solve a simultaneous equation system, the smaller time-step size should be used if converged solution could not obtain even in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. The PRT(Pressurizer Relief Tank) is a good example, which requires a dedicated model. The PRT consists of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume may limit the time-step size if it is modeled with a general control volume. To mitigate the time-step size reduction due to convergence failure at this component using RETRAN, the PRT model was developed as a dedicated model. The dedicated model was expected to provide reasonable results without convergence problem in the analysis of the system transients. The ARTS code guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there are some possibilities of calculation failure in the

  17. Thermal energy disposal methods for the proposed nuclear power plant at Vernon

    International Nuclear Information System (INIS)

    Converse, A.O.

    1974-01-01

    Several ways of cooling the proposed nuclear power plant are described, and estimates of the corresponding power costs are presented. The expected river temperatures and dissolved oxygen concentrations are presented as functions of the time of year. The problem of fog formation is discussed. Conclusions and recommendations are presented

  18. Study the dispersion the possible thermal discharges to Juragua nuclear power plant. Preliminary results

    International Nuclear Information System (INIS)

    Munnoz Caravaca, A.; Artega Rodriguez, H.; Diaz Asencio, M.; Cartas Aguila, H.

    1998-01-01

    The present work has as objective to present the results the evaluation to the surface area impact to the waters cooling discharges the Juragua nuclear power plant, by means of hydrodynamic models that described the possible distribution the same ones in the boundary marine means to the location

  19. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  20. Development of a consensus standard for verification and validation of nuclear system thermal-fluids software

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; Schultz, Richard R.; Crane, Ryan L.

    2011-01-01

    With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V and V) of software used to calculate the thermal–hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V and V 30 Committee, under the jurisdiction of the V and V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V and V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. Although software verification will be an important and necessary part of the standard, much of the initial effort of the committee will be focused on the validation of existing software and new models that could be used in the licensing process. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general

  1. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

  2. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  3. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  4. Toward Global Standards on Peaceful uses of Space Nuclear Reactor Power Systems

    Science.gov (United States)

    El-Genk, M. S.

    Space reactor power systems are enabling to future space exploration and outposts on the Moon, Mars and other celestial bodies. They could provide for electrical propulsion, thermal propulsion and bimodal operation of thermal propulsion and electrical power generation. The much higher specific impulse of nuclear thermal and electric propulsion, compared to chemical rockets, could significantly shorten the travel time to Mars and farthest destinations, reducing the exposure of equipment and astronauts to the harmful space radiation. This article addresses safety issues relevant to the design, operation and end-of-life storage of these systems in an attempt to stimulate a discussion of the need to establish specific and globally agreed upon safety standards and guidelines for future peaceful uses.

  5. DECOVALEX III PROJECT. Mathematical Models of Coupled Thermal-Hydro-Mechanical Processes for Nuclear Waste Repositories. Executive Summary

    International Nuclear Information System (INIS)

    Jing, L.; Stephansson, O.; Kautzky, F.

    2005-02-01

    DECOVALEX is an international consortium of governmental agencies associated with the disposal of high-level nuclear waste in a number of countries. The consortium's mission is the DEvelopment of COupled models and their VALidation against EXperiments. Hence the acronym/name DECOVALEX. Currently, agencies from Canada, Finland, France, Germany, Japan, Spain, Switzerland, Sweden, United Kingdom, and the United States are in DECOVALEX. Emplacement of nuclear waste in a repository in geologic media causes a number of physical processes to be intensified in the surrounding rock mass due to the decay heat from the waste. The four main processes of concern are thermal, hydrological, mechanical and chemical. Interactions or coupling between these heat-driven processes must be taken into account in modeling the performance of the repository for such modeling to be meaningful and reliable. DECOVALEX III is organized around four tasks. The FEBEX (Full-scale Engineered Barriers EXperiment) in situ experiment being conducted at the Grimsel site in Switzerland is to be simulated and analyzed in Task 1. Task 2, centered around the Drift Scale Test (DST) at Yucca Mountain in Nevada, USA, has several sub-tasks (Task 2A, Task 2B, Task 2C and Task 2D) to investigate a number of the coupled processes in the DST. Task 3 studies three benchmark problems: a) the effects of thermal-hydrologic-mechanical (THM) coupling on the performance of the near-field of a nuclear waste repository (BMT1); b) the effect of upscaling THM processes on the results of performance assessment (BMT2); and c) the effect of glaciation on rock mass behavior (BMT3). Task 4 is on the direct application of THM coupled process modeling in the performance assessment of nuclear waste repositories in geologic media. This executive summary presents the motivation, structure, objectives, approaches, and the highlights of the main achievements and outstanding issues of the tasks studied in the DECOVALEX III project. The

  6. Degradation of rocks, through cracking caused by differential thermal expansion, in relation to nuclear waste repositories

    International Nuclear Information System (INIS)

    McLaren, J.R.; Davidge, R.W.; Titchell, I.; Sincock, K.; Bromley, A.

    1982-01-01

    Heating to temperatures up to 500 0 C gives a reduction in Young's modulus and increases in permeability of granitic rocks and it is likely that a major reason is grain boundary cracking. The cracking of grain boundary facets in polycrystalline multiphase materials showing anistropic thermal expansion behaviour is controlled by several microstructural factors in addition to the intrinsic thermal and elastic properties. Of specific interest are the relative orientations of the two grains meeting at the facet, and the size of the facet; these factors thus introduce two statistical aspects to the problem and these are introduced to give quantitative data on crack density versus temperature. The theory is compared with experimental measurements of Young's modulus and permeability for various rocks as a function of temperature. There is good qualitative agreement, and the additional (mainly microstructural) data required for a quantitative comparison are defined. 6 figures, 2 tables

  7. 3D numerical modelling of the thermal state of deep geological nuclear waste repositories

    Science.gov (United States)

    Butov, R. A.; Drobyshevsky, N. I.; Moiseenko, E. V.; Tokarev, Yu. N.

    2017-09-01

    One of the important aspects of the high-level radioactive waste (HLW) disposal in deep geological repositories is ensuring the integrity of the engineered barriers which is, among other phenomena, considerably influenced by the thermal loads. As the HLW produce significant amount of heat, the design of the repository should maintain the balance between the cost-effectiveness of the construction and the sufficiency of the safety margins, including those imposed on the thermal conditions of the barriers. The 3D finite-element computer code FENIA was developed as a tool for simulation of thermal processes in deep geological repositories. Further the models for mechanical phenomena and groundwater hydraulics will be added resulting in a fully coupled thermo-hydro-mechanical (THM) solution. The long-term simulations of the thermal state were performed for two possible layouts of the repository. One was based on the proposed project of Russian repository, and another features larger HLW amount within the same space. The obtained results describe the spatial and temporal evolution of the temperature filed inside the repository and in the surrounding rock for 3500 years. These results show that practically all generated heat was ultimately absorbed by the host rock without any significant temperature increase. Still in the short time span even in case of smaller amount of the HLW the temperature maximum exceeds 100 °C, and for larger amount of the HLW the local temperature remains above 100 °C for considerable time. Thus, the substantiation of the long-term stability of the repository would require an extensive study of the materials properties and behaviour in order to remove the excessive conservatism from the simulations and to reduce the uncertainty of the input data.

  8. Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Antoniak, Z.I.

    1986-10-01

    PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data

  9. Thermal characteristics of spent activated carbon generated from air cleaning units in Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hang Rae; So, Ji Yang [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  10. APPLICATION OF MONITORING, DIAGNOSIS, AND PROGNOSIS IN THERMAL PERFORMANCE ANALYSIS FOR NUCLEAR POWER PLANTS

    Directory of Open Access Journals (Sweden)

    HYEONMIN KIM

    2014-12-01

    Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.

  11. Thermal characteristics of spent activated carbon generated from air cleaning units in korean nuclear power plants

    Directory of Open Access Journals (Sweden)

    Ji-Yang So

    2017-06-01

    Full Text Available To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  12. The Economic Potential of Three Nuclear-Renewable Hybrid Energy Systems Providing Thermal Energy to Industry

    Energy Technology Data Exchange (ETDEWEB)

    Ruth, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States); Cutler, Dylan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Flores-Espino, Francisco [National Renewable Energy Lab. (NREL), Golden, CO (United States); Stark, Greg [National Renewable Energy Lab. (NREL), Golden, CO (United States); Jenkin, Thomas [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-12-01

    This report is one of a series of reports that Idaho National Laboratory and National Renewable Energy Laboratory are producing to investigate the technical and economic aspects of nuclear-renewable hybrid energy systems (N-R HESs). Previous reports provided results of an analysis of two N-R HES scenarios. This report builds that analysis with a Texas-synthetic gasoline scenario providing the basis in which the N-R HES sells heat directly to an industrial customer. Subsystems were included that convert electricity to heat, thus allowing the renewable energy subsystem to generate heat and benefit from that revenue stream. Nuclear and renewable energy sources are important to consider in the energy sector's evolution because both are considered to be clean and non-carbon-emitting energy sources.

  13. The Economic Potential of Three Nuclear-Renewable Hybrid Energy Systems Providing Thermal Energy to Industry

    International Nuclear Information System (INIS)

    Ruth, Mark; Cutler, Dylan; Flores-Espino, Francisco; Stark, Greg; Jenkin, Thomas

    2016-01-01

    This report is one of a series of reports that Idaho National Laboratory and National Renewable Energy Laboratory are producing to investigate the technical and economic aspects of nuclear-renewable hybrid energy systems (N-R HESs). Previous reports provided results of an analysis of two N-R HES scenarios. This report builds that analysis with a Texas-synthetic gasoline scenario providing the basis in which the N-R HES sells heat directly to an industrial customer. Subsystems were included that convert electricity to heat, thus allowing the renewable energy subsystem to generate heat and benefit from that revenue stream. Nuclear and renewable energy sources are important to consider in the energy sector's evolution because both are considered to be clean and non-carbon-emitting energy sources.

  14. Thermal tolerances of fish from a reservoir receiving heated effluent from a nuclear reactor

    International Nuclear Information System (INIS)

    Holland, W.E.; Smith, M.H.; Gibbons, J.W.; Brown, D.H.

    1974-01-01

    The heat tolerances of bluegill (Lepomis macrochirus) subjected to heated effluent from a nuclear reactor was compared with those of bluegill living at normal temperatures. Three of the four study areas were located in the Par Pond reservoir system on the Savannah River Plant near Aiken, South Carolina. Results shown that at least one species of warm-water fish can adjust to elevated aquatic temperatures in a natural environment by becoming more tolerant. (U.S.)

  15. Human and mouse switch-like genes share common transcriptional regulatory mechanisms for bimodality

    Directory of Open Access Journals (Sweden)

    Tozeren Aydin

    2008-12-01

    Full Text Available Abstract Background Gene expression is controlled over a wide range at the transcript level through complex interplay between DNA and regulatory proteins, resulting in profiles of gene expression that can be represented as normal, graded, and bimodal (switch-like distributions. We have previously performed genome-scale identification and annotation of genes with switch-like expression at the transcript level in mouse, using large microarray datasets for healthy tissue, in order to study the cellular pathways and regulatory mechanisms involving this class of genes. We showed that a large population of bimodal mouse genes encoding for cell membrane and extracellular matrix proteins is involved in communication pathways. This study expands on previous results by annotating human bimodal genes, investigating their correspondence to bimodality in mouse orthologs and exploring possible regulatory mechanisms that contribute to bimodality in gene expression in human and mouse. Results Fourteen percent of the human genes on the HGU133A array (1847 out of 13076 were identified as bimodal or switch-like. More than 40% were found to have bimodal mouse orthologs. KEGG pathways enriched for bimodal genes included ECM-receptor interaction, focal adhesion, and tight junction, showing strong similarity to the results obtained in mouse. Tissue-specific modes of expression of bimodal genes among brain, heart, and skeletal muscle were common between human and mouse. Promoter analysis revealed a higher than average number of transcription start sites per gene within the set of bimodal genes. Moreover, the bimodal gene set had differentially methylated histones compared to the set of the remaining genes in the genome. Conclusion The fact that bimodal genes were enriched within the cell membrane and extracellular environment make these genes as candidates for biomarkers for tissue specificity. The commonality of the important roles bimodal genes play in tissue

  16. Affordable Development and Demonstration of a Small Nuclear Thermal Rocket (NTR) Engine and Stage: How Small Is Big Enough?

    Science.gov (United States)

    Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg, Abraham; Joyner, Claude R.

    2016-01-01

    The Nuclear Thermal Rocket (NTR) derives its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core. It generates high thrust and has a specific impulse potential of approximately 900 specific impulse - a 100 percent increase over today's best chemical rockets. The Nuclear Thermal Propulsion (NTP) project, funded by NASA's Advanced Exploration Systems (AES) program, includes five key task activities: (1) Recapture, demonstration, and validation of heritage graphite composite (GC) fuel (selected as the Lead Fuel option); (2) Engine Conceptual Design; (3) Operating Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable Development Strategy. During fiscal year (FY) 2014, a preliminary Design Development Test and Evaluation (DDT&E) plan and schedule for NTP development was outlined by the NASA Glenn Research Center (GRC), Department of Energy (DOE) and industry that involved significant system-level demonstration projects that included Ground Technology Demonstration (GTD) tests at the Nevada National Security Site (NNSS), followed by a Flight Technology Demonstration (FTD) mission. To reduce cost for the GTD tests and FTD mission, small NTR engines, in either the 7.5 or 16.5 kilopound-force thrust class, were considered. Both engine options used GC fuel and a common fuel element (FE) design. The small approximately 7.5 kilopound-force criticality-limited engine produces approximately157 thermal megawatts and its core is configured with parallel rows of hexagonal-shaped FEs and tie tubes (TTs) with a FE to TT ratio of approximately 1:1. The larger approximately 16.5 kilopound-force Small Nuclear Rocket Engine (SNRE), developed by Los Alamos National Laboratory (LANL) at the end of the Rover program, produces approximately 367 thermal megawatts and has a FE to TT ratio of approximately 2:1. Although both engines use a common 35-inch (approximately

  17. Final Environmental Impact Statement (EIS) for the Space Nuclear Thermal Propulsion (SNTP) program

    Science.gov (United States)

    1991-09-01

    A program has been proposed to develop the technology and demonstrate the feasibility of a high-temperature particle bed reactor (PBR) propulsion system to be used to power an advanced second stage nuclear rocket engine. The purpose of this Final Environmental Impact Statement (FEIS) is to assess the potential environmental impacts of component development and testing, construction of ground test facilities, and ground testing. Major issues and goals of the program include the achievement and control of predicted nuclear power levels; the development of materials that can withstand the extremely high operating temperatures and hydrogen flow environments; and the reliable control of cryogenic hydrogen and hot gaseous hydrogen propellant. The testing process is designed to minimize radiation exposure to the environment. Environmental impact and mitigation planning are included for the following areas of concern: (1) Population and economy; (2) Land use and infrastructure; (3) Noise; (4) Cultural resources; (5) Safety (non-nuclear); (6) Waste; (7) Topography; (8) Geology; (9) Seismic activity; (10) Water resources; (11) Meteorology/Air quality; (12) Biological resources; (13) Radiological normal operations; (14) Radiological accidents; (15) Soils; and (16) Wildlife habitats.

  18. Thermal coupling in low fields between the nuclear and electronic spins in Tm2+ doped CaF2

    International Nuclear Information System (INIS)

    Urbina, Cristian.

    1977-01-01

    It is shown that in a CaF 2 crystal doped with divalent thulium ions there is in low fields, a thermal coupling between the electron magnetic moments of Tm 2+ and the nuclear moments of 19 F. When these ones have been lowered down to temperature through dynamical high-field polarization and adiabatic demagnetization in succession the resulting polarisation of the formed ones can overstep their original polarization in high field. A trial is given to explain this Zeeman electronic energy cooling through nuclear Zeeman energy with invoking a thermal coupling between both systems through the spin-spin electronic interaction but no theoretical model is developed in view of a quantitative explanation of the dynamics of such a process. The magnetic resonance spectrum of Tm 2 + in low field is also investigated: an important shift and narrowing of the electron resonance line in low field are obtained when 19 F nuclei are very cold. This special spectral characters are explained as due to magnetic interactions between electronic impurities and the neighbouring 19 F nuclei and a theoretical model is developed (based on the local Weiss field approximation) which explains rather well the changes in the spectral shift as a function of the 19 F nucleus temperature. A second theoretical model has also been developed in view of a quantitative explanation of both the narrowing and shift of the spectrum, but its prediction disagree with the experimental results. It is shown that in low fieldsx it is possible to get rid of paramagnetic impurities after they have been reused as reducing agents for 19 F nucleus entropy populating at about 80%, a non magnetic metastable state with these impurities [fr

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

    Science.gov (United States)

    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.

  20. Thermal and nuclear power plants: Competitiveness in the new economic conditions

    Science.gov (United States)

    Aminov, R. Z.; Shkret, A. F.; Garievskii, M. V.

    2017-05-01

    In recent years, the conditions of development and functionality of power generating assets have notably changed. Considering the decline in the price of hydrocarbon fuel on the global market, the efficiency of combined-cycle gas-turbine plants in the European part of Russia is growing in comparison with nuclear power plants. Capital investments in the construction of nuclear power plants have also increased as a result of stiffening the safety requirements. In view of this, there has been an increasing interest in exploration of effective lines of development of generating assets in the European part of Russia, taking consideration of the conditions that may arise in the nearest long-term perspective. In particular, the assessment of comparative efficiency of developing combined-cycle gas-turbine plants (operating on natural gas) in the European part of Russia and nuclear power plants is of academic and practical interest. In this article, we analyze the trends of changes in the regional price of hydrocarbon fuel. Using the prognosis of net-weighted import prices of natural gas in Western European countries—prepared by the International Energy Agency (IEA) and the Energy Research Institute of the Russian Academy of Sciences (ERIRAS)—the prices of natural gas in the European part of Russia equilibrated with import prices of this heat carrier in Western Europe were determined. The methodology of determining the comparative efficiency of combined-cycle gas turbine plants (CCGT) and nuclear power plants (NPP) were described; based on this, the possible development of basic CCGTs and NPPs with regard to the European part of Russia for various scenarios in the prognosis of prices of gaseous fuel in a broad range of change of specific investments in the given generating sources were assessed, and the extents of their comparative efficiency were shown. It was proven that, at specific investments in the construction of new NPPs in the amount of 5000 dollars/kW, nuclear

  1. Nuclear fuels technologies: Thermally induced gallium removal system (TIGRS), fiscal year 1998 research and development test plan

    International Nuclear Information System (INIS)

    Buksa, J.J.; Butt, D.P.; Chidester, K.; DeMuth, S.F.; Havrilla, G.J.; James, C.A.; Kolman, D.G.

    1997-01-01

    This document details the research and development (R and D) activities that will be conducted in Fiscal Year 1998 (FY98) by the Thermally Induced Gallium Removal System (TIGRS) team for the Department of Energy Office of Fissile Materials Disposition. This work is a continuation and extension of experimental activities that have been conducted in support of using weapons-derived plutonium in the fabrication of mixed-oxide (MOX) nuclear fuel for reactor-based plutonium disposition. The ultimate purpose of this work is to demonstrate adequate Thermally Induced Gallium Removal with a prototypic system. This Test Plan presents more than the FY98 R and D efforts in order to frame the Task in its entirety. To achieve the TIGRS Program objectives, R and D activities during the next two years will be focused on (1) process development leading to a prototypic TIGRS design, and (2) prototypic TIGRS design and testing leading to and including a prototypic demonstration of TIGRS operation. Both the process development and system testing efforts will consist of a series of surrogate-based cold tests and plutonium-based hot tests. Some of this testing has already occurred and will continue into FY99

  2. A new multi-scale platform for advanced nuclear thermal-hydraulics status and prospects of the Neptune project

    International Nuclear Information System (INIS)

    Bestion, D.; Boudier, P.; Hervieu, E.; Boucker, M.; Peturaud, P.; Guelfi, A.; Fillion, P.; Grandotto, M.; Herard, J.M.

    2005-01-01

    Full text of publication follows: Further to a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF (Electricite de France) and CEA (Commissariat a l'Energie Atomique) launched in 2001 a new long-term joint development program for the next generation of nuclear reactors simulation tools. The NEPTUNE Project, which constitutes the Thermal-Hydraulics part of this comprehensive program, aims at building a new software platform for advanced two-phase flow thermal-hydraulics allowing easy multi-scale and multi-disciplinary calculations meeting the industrial needs. The NEPTUNE activities include software development, research in physical modeling and numerical methods, the development of advanced instrumentation techniques and performance of new experimental programs. The work focuses on the four different simulation scales: DNS (Direct Numerical Simulation), local CFD (Computational Fluid Dynamics), component (subchannel-type analysis) and system scales. New physical models and numerical methods are being developed for each scale as well as for their coupling. This paper gives an overview of the NEPTUNE activities. It presents the main scientific and technical achievements obtained during Phase 1 (2002-2003) and at the beginning of Phase 2 (2004- 2006). Planned work for the future is also presented. (authors)

  3. Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Suzuki, M. [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan)

    2014-09-15

    The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400 °C for periods of 100–10,000 h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100 h, and gradually for aging durations greater than 1000 h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000 h, had the highest number density after aging for 2000 h, and coarsened after aging for 10,000 h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000 h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition.

  4. Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.; Suzuki, M.

    2014-01-01

    The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400 °C for periods of 100–10,000 h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100 h, and gradually for aging durations greater than 1000 h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000 h, had the highest number density after aging for 2000 h, and coarsened after aging for 10,000 h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000 h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition

  5. Studies of a laser/nuclear thermal-hardened body armor. Final report, 31 Jan 91-30 Sep 91

    Energy Technology Data Exchange (ETDEWEB)

    Misconi, N.Y.; Caldarella, G.J.; Roach, J.F.

    1992-08-01

    The problem of laser/nuclear hardening of body armors and other applications, such as rigid wall, etc, has been investigated in this study. Earlier results from studies of hardening against space systems, which were supported by the Air Force Office of Scientific Research (AFOSR) and carried out by the Principal Investigator during 1984 to 1989 are summarized. The concepts of particle layer and photon multiple scattering inside the layers were utilized in developing a laser shield to protect against laser weapons in the 0.22 to 2.4 micrometer region of the spectrum. Protection against the threats from C02 laser weapons are addressed, and the development of a protective shield is detailed. It is now possible to apply a coating that will protect against laser/nuclear threats and reduction of solar loads for 0.22 to 16 micrometers of the spectrum. Applications are expected for rigid walls (Army containers), human body armor, thermal jackets for military hardware, etc. Finally, a mathematical model was created to help predict how the laser hardening material will behave under specific constraints that have not yet been tested in the laboratory. Also, this model can be used to extrapolate the performance of similar materials/coatings in the mid- to far-infrared wavelengths and also predict the broadband performance.

  6. A Novel, Safe, and Environmentally Friendly Technology for Water Production Through Recovery of Rejected Thermal Energy From Nuclear Power Plants

    International Nuclear Information System (INIS)

    Khalil, Yehia F.; Elimelech, Menachem

    2006-01-01

    In this work, we describe a novel design that utilizes seawater and a portion of rejected heat from a nuclear plant's steam cycle to operate a water desalination system using forward osmosis technology. Water produced from this process is of sufficient quality to be readily used to supply plant demands for continuous makeup water. The proposed process minimizes the environmental concerns associated with thermal pollution of public waters and the resulting adverse impact on marine ecology. To demonstrate the technical feasibility of this conceptual design of a water treatment process, we discuss a case study as an example to describe how the proposed design can be implemented in a nuclear power station with a once--through cooling system that discharges rejected heat to an open sound seawater as its ultimate heat sink. In this case study, the station uses a leased (vendor owned and operated) onsite water treatment system that demineralizes and polishes up to 500-gpm of city water (at 100 ppm TDS) to supply high-quality makeup water (< 0.01 ppm TDS) to the plant steam system. The objectives of implementing the new design are three fold: 1) forego current practice of using city water as the source of plant makeup water, thereby reducing the nuclear station's impact on the region's potable water supply by roughly 100 million gallons/year, 2) minimize the adverse impact of discharging rejected heat into the open sound seawater and, hence, protect the marine ecology, and 3) eliminate the reliance on external vendor that owns and operates the onsite water treatment system, thereby saving an annual fixed cost of $600 K plus 6 cents per 1,000 gallons of pure water. The design will also eliminate the need for using two double-path reverse osmosis (RO) units that consume 425 kW/h of electric power to operate two RO pumps (480 V, 281.6 HP, and 317.4 amps). (authors)

  7. The path to clean energy: direct coupling of nuclear and renewable technologies for thermal and electrical applications

    Energy Technology Data Exchange (ETDEWEB)

    Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Fuel Performance and Design; Boardman, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Process and Decision Systems; Ruth, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States). Strategic Energy Analysis Center

    2015-07-01

    The U.S. Department of Energy (DOE) recognizes the need to transform the energy infrastructure of the U.S. and elsewhere to systems that can significantly reduce environmental impacts in an efficient and economically viable manner while utilizing both clean energy generation sources and hydrocarbon resources. Thus, DOE is supporting research and development that could lead to more efficient utilization of clean nuclear and renewable energy generation sources. A concept being advanced by the DOE Offices of Nuclear Energy (NE) and Energy Efficiency and Renewable Energy (EERE) is tighter coupling of nuclear and renewable energy sources in a manner that better optimizes energy use for the combined electricity, industrial manufacturing, and the transportation sectors. This integration concept has been referred to as a “hybrid system” that is capable of providing energy (thermal or electrical) where it is needed, when it is needed. For the purposes of this work, the hybrid system would integrate two or more energy resources to generate two or more products, one of which must be an energy commodity, such as electricity or transportation fuel. This definition requires coupling of subsystems ‘‘behind’’ the electrical transmission bus, where energy flows are dynamically apportioned as necessary to meet demand and the system has a single connection to the grid that provides dispatchable electricity as required while capital intensive generation assets operate at full capacity. Development of integrated energy systems for an “energy park” must carefully consider the intended location and the associated regional resources, traditional industrial processes, energy delivery infrastructure, and markets to identify viable region-specific system configurations. This paper will provide an overview of the current status of regional hybrid energy system design, development and application of dynamic analysis tools to assess technical and economic performance, and

  8. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  9. Increasing of prediction reliability of calcium carbonate scale formation in heat exchanger of secondary coolant circuits of thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Tret'yakov, O.V.; Kritskij, V.G.; Styazhkin, P.S.

    1991-01-01

    Calcium carbonate scale formation in the secondary circuit heat exchanger of thermal and nuclear power plants is investigated. A model of calcium-carbonate scale formation providing quite reliable prediction of process running and the possibility of its control affecting the parameters of hydrochemical regime (HCR) is developed. The results can be used when designing the automatic-control system of HCR

  10. Natural dynamic stability of thermal and nuclear power sets at 3000 and 1500 rev/min

    International Nuclear Information System (INIS)

    Caucheteux, G.

    1974-01-01

    The natural dynamic stability time limits were determined for the main types of power set, using a micro-network. By definition, the characteristic feature of natural dynamic stability is that speed governing is in effective, i.e. a constant driving torque is applied to the machine shaft. The time limits give an order of magnitude of the performance to be expected from future power sets connected to the system in the 1980s, for which low-inertia conventional sets running at 3000 rev/min or high-inertia nuclear power sets at 1500 rev/min are envisaged [fr

  11. Microstructural and Fractographic Characterization of a Thermally Embrittled Nuclear Grade Steel: Part I - Annealing

    Directory of Open Access Journals (Sweden)

    Tarpani José R.

    2002-01-01

    Full Text Available A nuclear reactor pressure vessel steel was submitted to different annealing heat treatments aimed at simulating neutron irradiation damage. The obtained microstructures were mechanically tested with subsequent metallographic and fractographic characterization. The relevant microstructural and fractographic aspects were employed in the interpretation of the mechanical behavior of the microstructures in both quasi-static (J-R curve and dynamic (Charpy impact loading regimes. A well defined relationship was determined between the elastic-plastic fracture toughness parameter J-integral and the Charpy impact energy for very most of the microstructures.

  12. The climatic imprint of bimodal distributions in vegetation cover for western Africa

    NARCIS (Netherlands)

    Yin, Z.; Dekker, S. C.; van den Hurk, B. J. J. M.; Dijkstra, H. A.

    2016-01-01

    Observed bimodal distributions of woody cover in western Africa provide evidence that alternative ecosystem states may exist under the same precipitation regimes. In this study, we show that bimodality can also be observed in mean annual shortwave radiation and above-ground biomass, which might

  13. Gaze-independent ERP-BCIs: augmenting performance through location-congruent bimodal stimuli

    Science.gov (United States)

    Thurlings, Marieke E.; Brouwer, Anne-Marie; Van Erp, Jan B. F.; Werkhoven, Peter

    2014-01-01

    Gaze-independent event-related potential (ERP) based brain-computer interfaces (BCIs) yield relatively low BCI performance and traditionally employ unimodal stimuli. Bimodal ERP-BCIs may increase BCI performance due to multisensory integration or summation in the brain. An additional advantage of bimodal BCIs may be that the user can choose which modality or modalities to attend to. We studied bimodal, visual-tactile, gaze-independent BCIs and investigated whether or not ERP components’ tAUCs and subsequent classification accuracies are increased for (1) bimodal vs. unimodal stimuli; (2) location-congruent vs. location-incongruent bimodal stimuli; and (3) attending to both modalities vs. to either one modality. We observed an enhanced bimodal (compared to unimodal) P300 tAUC, which appeared to be positively affected by location-congruency (p = 0.056) and resulted in higher classification accuracies. Attending either to one or to both modalities of the bimodal location-congruent stimuli resulted in differences between ERP components, but not in classification performance. We conclude that location-congruent bimodal stimuli improve ERP-BCIs, and offer the user the possibility to switch the attended modality without losing performance. PMID:25249947

  14. Nonlatching positive feedback enables robust bimodality by decoupling expression noise from the mean

    Science.gov (United States)

    Razooky, Brandon S.; Cao, Youfang; Hansen, Maike M. K.; Perelson, Alan S.; Simpson, Michael L.

    2017-01-01

    Fundamental to biological decision-making is the ability to generate bimodal expression patterns where 2 alternate expression states simultaneously exist. Here, we use a combination of single-cell analysis and mathematical modeling to examine the sources of bimodality in the transcriptional program controlling HIV’s fate decision between active replication and viral latency. We find that the HIV transactivator of transcription (Tat) protein manipulates the intrinsic toggling of HIV’s promoter, the long terminal repeat (LTR), to generate bimodal ON-OFF expression and that transcriptional positive feedback from Tat shifts and expands the regime of LTR bimodality. This result holds for both minimal synthetic viral circuits and full-length virus. Strikingly, computational analysis indicates that the Tat circuit’s noncooperative “nonlatching” feedback architecture is optimized to slow the promoter’s toggling and generate bimodality by stochastic extinction of Tat. In contrast to the standard Poisson model, theory and experiment show that nonlatching positive feedback substantially dampens the inverse noise-mean relationship to maintain stochastic bimodality despite increasing mean expression levels. Given the rapid evolution of HIV, the presence of a circuit optimized to robustly generate bimodal expression appears consistent with the hypothesis that HIV’s decision between active replication and latency provides a viral fitness advantage. More broadly, the results suggest that positive-feedback circuits may have evolved not only for signal amplification but also for robustly generating bimodality by decoupling expression fluctuations (noise) from mean expression levels. PMID:29045398

  15. Nonlatching positive feedback enables robust bimodality by decoupling expression noise from the mean

    Energy Technology Data Exchange (ETDEWEB)

    Razooky, Brandon S. [Rockefeller Univ., New York, NY (United States). Lab. of Virology and Infectious Disease; Gladstone Institutes (Virology and Immunology), San Francisco, CA (United States); Univ. of California, San Francisco, CA (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Center for Nanophase Materials Science (CNMS); Univ. of Tennessee, Knoxville, TN (United States). Bredesen Center for Interdisciplinary; Cao, Youfang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hansen, Maike M. K. [Gladstone Institutes (Virology and Immunology), San Francisco, CA (United States); Perelson, Alan S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Simpson, Michael L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Center for Nanophase Materials Science (CNMS); Univ. of Tennessee, Knoxville, TN (United States). Bredesen Center for Interdisciplinary; Weinberger, Leor S. [Gladstone Institutes (Virology and Immunology), San Francisco, CA (United States); Univ. of California, San Francisco, CA (United States). Dept. of Biochemistry and Biophysics; Univ. of California, San Francisco, CA (United States). QB3: California Inst. of Quantitative Biosciences; Univ. of California, San Francisco, CA (United States). Dept. of Pharmaceutical Chemistry

    2017-10-18

    Fundamental to biological decision-making is the ability to generate bimodal expression patterns where two alternate expression states simultaneously exist. Here in this study, we use a combination of single-cell analysis and mathematical modeling to examine the sources of bimodality in the transcriptional program controlling HIV’s fate decision between active replication and viral latency. We find that the HIV Tat protein manipulates the intrinsic toggling of HIV’s promoter, the LTR, to generate bimodal ON-OFF expression, and that transcriptional positive feedback from Tat shifts and expands the regime of LTR bimodality. This result holds for both minimal synthetic viral circuits and full-length virus. Strikingly, computational analysis indicates that the Tat circuit’s non-cooperative ‘non-latching’ feedback architecture is optimized to slow the promoter’s toggling and generate bimodality by stochastic extinction of Tat. In contrast to the standard Poisson model, theory and experiment show that non-latching positive feedback substantially dampens the inverse noise-mean relationship to maintain stochastic bimodality despite increasing mean-expression levels. Given the rapid evolution of HIV, the presence of a circuit optimized to robustly generate bimodal expression appears consistent with the hypothesis that HIV’s decision between active replication and latency provides a viral fitness advantage. More broadly, the results suggest that positive-feedback circuits may have evolved not only for signal amplification but also for robustly generating bimodality by decoupling expression fluctuations (noise) from mean expression levels.

  16. Bimodal distribution of glucose is not universally useful for diagnosing diabetes

    DEFF Research Database (Denmark)

    Vistisen, Dorte; Colagiuri, Stephen; Borch-Johnsen, Knut

    2009-01-01

    OBJECTIVE: Bimodality in the distribution of glucose has been used to define the cut point for the diagnosis of diabetes. Previous studies on bimodality have primarily been in populations with a high prevalence of type 2 diabetes, including one study in a white Caucasian population. All studies i...

  17. Characterization of polymer-type ionic conductors using nuclear magnetic resonance and thermal analysis. Humidity sensor

    International Nuclear Information System (INIS)

    Cavalcante, Maria Goretti.

    1992-04-01

    We report a study using Nuclear Magnetic Resonance (NMR), Thermogravimetry Analysis, Differential Scanning Calorimetry and Infrared Spectroscopy in polymeric complexes formed poly(ethylene oxide), (PEO), and lithium salts. These complexes have have shown a large potential for technological applications in batteries, sensors, etc. We developed and characterized humidity sensors and discussed how the humidity affects the conformation of the complexes, the mobility of ionic species, and the polymeric chains. The results indicate that the hydration affects the conformation of polymeric complexes by plasticizing the water, which induces a volumetric expansion in the PEO chain. The processes was completely reversible for the level of hydration studied. NMR was used to distinguish the movement of polymeric chains from the movement of the ionic species. From the analysis of the second moment of resonance lines from the study of the nuclear relaxation we were able to estimate the average distance between the ionic species and the proton in the complexes chains. The behaviour of spin -lattice relaxation of hydrogen and fluorine in the P(EO) - Li B F, as a function of temperature and frequency reflects the nature of the disorder and the complexity of the ionic conduction process in these materials. (author). 91 refs., 69 figs., 2 tabs

  18. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Science.gov (United States)

    Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

    2018-01-01

    The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  19. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Directory of Open Access Journals (Sweden)

    Spaccapaniccia C.

    2018-01-01

    Full Text Available The Belgian nuclear research institute (SCK•CEN is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS. The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE. The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC. After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection. The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  20. Isomap nonlinear dimensionality reduction and bimodality of Asian monsoon convection

    Science.gov (United States)

    Hannachi, A.; Turner, A. G.

    2013-04-01

    It is known that the empirical orthogonal function method is unable to detect possible nonlinear structure in climate data. Here, isometric feature mapping (Isomap), as a tool for nonlinear dimensionality reduction, is applied to 1958-2001 ERA-40 sea-level pressure anomalies to study nonlinearity of the Asian summer monsoon intraseasonal variability. Using the leading two Isomap time series, the probability density function is shown to be bimodal. A two-dimensional bivariate Gaussian mixture model is then applied to identify the monsoon phases, the obtained regimes representing enhanced and suppressed phases, respectively. The relationship with the large-scale seasonal mean monsoon indicates that the frequency of monsoon regime occurrence is significantly perturbed in agreement with conceptual ideas, with preference for enhanced convection on intraseasonal time scales during large-scale strong monsoons. Trend analysis suggests a shift in concentration of monsoon convection, with less emphasis on South Asia and more on the East China Sea.