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Sample records for ebr-ii blanket elements

  1. Comparison of measured and calculated composition of irradiated EBR-II blanket assemblies.

    SciTech Connect

    Grimm, K. N.

    1998-07-13

    In anticipation of processing irradiated EBR-II depleted uranium blanket subassemblies in the Fuel Conditioning Facility (FCF) at ANL-West, it has been possible to obtain a limited set of destructive chemical analyses of samples from a single EBR-II blanket subassembly. Comparison of calculated values with these measurements is being used to validate a depletion methodology based on a limited number of generic models of EBR-II to simulate the irradiation history of these subassemblies. Initial comparisons indicate these methods are adequate to meet the operations and material control and accountancy (MC and A) requirements for the FCF, but also indicate several shortcomings which may be corrected or improved.

  2. Disposition of Unirradiated Sodium Bonded EBR-II Driver Fuel Elements and HEU Scrap: Work Performed for FY 2007

    SciTech Connect

    Karen A Moore

    2007-04-01

    Specific surplus high enriched uranium (HEU) materials at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) will be transferred to a designated off-site receiving facility. The DOE High Enriched Uranium Disposition Program Office (HDPO) will determine which materials, if any, will be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for nuclear reactors. These surplus HEU materials include approximately 7200 kg unirradiated sodium-bonded EBR-II driver fuel elements, and nearly 800 kg of HEU casting scrap from the process which formed various sodium-bonded fuels (including the EBR-II driver elements). Before the driver fuel can be packaged for shipment, the fuel elements will require removal of the sodium bond. The HEU scrap will also require repackaging in preparation for off-site transport. Preliminary work on this task was authorized by BWXT Y-12 on Nov 6, 2006 and performed in three areas: • Facility Modifications • Safety Documentation • Project Management

  3. EBR-II Data Digitization

    SciTech Connect

    Yoon, Su-Jong; Rabiti, Cristian; Sackett, John

    2014-08-01

    1. Objectives To produce a validation database out of those recorded signals it will be necessary also to identify the documents need to reconstruct the status of reactor at the time of the beginning of the recordings. This should comprehends the core loading specification (assemblies type and location and burn-up) along with this data the assemblies drawings and the core drawings will be identified. The first task of the project will be identify the location of the sensors, with respect the reactor plant layout, and the physical quantities recorded by the Experimental Breeder Reactor-II (EBR-II) data acquisition system. This first task will allow guiding and prioritizing the selection of drawings needed to numerically reproduce those signals. 1.1 Scopes and Deliverables The deliverables of this project are the list of sensors in EBR-II system, the identification of storing location of those sensors, identification of a core isotopic composition at the moment of the start of system recording. Information of the sensors in EBR-II reactor system was summarized from the EBR-II system design descriptions listed in Section 1.2.

  4. Deactivation of the EBR-II complex

    SciTech Connect

    Michelbacher, J.A.; Earle, O.K.; Henslee, S.P.

    1997-12-31

    In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D&D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D&D plan has necessitated this current action. The EBR-II is a pool-type reactor. The primary system contains approximately 87,000 gallons of sodium, while the secondary system has 13,000 gallons. In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility has been built to react the sodium to a dry carbonate powder in a two stage process. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in the primary and secondary systems must be either reacted or inerted to preclude future concerns with sodium-air reactions that generate explosive mixtures of hydrogen and leave corrosive compounds. Residual amounts of sodium on components will effectively {open_quotes}solder{close_quotes} components in place, making future operation or removal unfeasible.

  5. EBR-II and TREAT Digitization Project

    SciTech Connect

    Griffith, George W.; Rabiti, Cristian

    2015-09-01

    Digitizing the technical drawings for EBR-II and TREAT provides multiple benefits. Moving the scanned or hard copy drawings to modern 3-D CAD (Computer Aided Drawing) format saves data that could be lost over time. The 3-D drawings produce models that can interface with other drawings to make complex assemblies. The 3-D CAD format can also include detailed material properties and parametric coding that can tie critical dimensions together allowing easier modification. Creating the new files from the old drawings has found multiple inconsistencies that are being flagged or corrected improving understanding of the reactor(s).

  6. EBR-II Primary Tank Wash-Water Alternatives Evaluation

    SciTech Connect

    Demmer, R. L.; Heintzelman, J. B.; Merservey, R. H.; Squires, L. N.

    2008-05-01

    The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. It was shut down in 1994; the fuel was removed by 1996; and the bulk of sodium metal coolant was removed from the reactor by 2001. Approximately 1100 kg of residual sodium remained in the primary system after draining the bulk sodium. To stabilize the remaining sodium, both the primary and secondary systems were treated with a purge of moist carbon dioxide. Most of the residual sodium reacted with the carbon dioxide and water vapor to form a passivation layer of primarily sodium bicarbonate. The passivation treatment was stopped in 2005 and the primary system is maintained under a blanket of dry carbon dioxide. Approximately 670 kg of sodium metal remains in the primary system in locations that were inaccessible to passivation treatment or in pools of sodium that were too deep for complete penetration of the passivation treatment. The EBR-II reactor was permitted by the Idaho Department of Environmental Quality (DEQ) in 2002 under a RCRA permit that requires removal of all remaining sodium in the primary and secondary systems by 2022. The proposed baseline closure method would remove the large components from the primary tank, fill the primary system with water, react the remaining sodium with the water and dissolve the reaction products in the wash water. This method would generate a minimum of 100,000 gallons of caustic, liquid, low level radioactive, hazardous waste water that must be disposed of in a permitted facility. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to look at alternatives that could meet the RCRA permit clean closure requirements and minimize the quantity of hazardous waste generated by the cleanup process. The workshop convened a panel of national and international sodium cleanup specialists, subject matter experts from the INL, and the EBR-II Wash Water Project team that organized the workshop. The

  7. Embedded computer systems for control applications in EBR-II

    SciTech Connect

    Carlson, R.B.; Start, S.E.

    1993-03-01

    The purpose of this paper is to describe the embedded computer systems approach taken at Experimental Breeder Reactor II (EBR-II) for non-safety related systems. The hardware and software structures for typical embedded systems are presented The embedded systems development process is described. Three examples are given which illustrate typical embedded computer applications in EBR-II.

  8. Embedded computer systems for control applications in EBR-II

    SciTech Connect

    Carlson, R.B.; Start, S.E.

    1993-01-01

    The purpose of this paper is to describe the embedded computer systems approach taken at Experimental Breeder Reactor II (EBR-II) for non-safety related systems. The hardware and software structures for typical embedded systems are presented The embedded systems development process is described. Three examples are given which illustrate typical embedded computer applications in EBR-II.

  9. Reliability and extended-life potential of EBR-II

    SciTech Connect

    King, R W

    1985-01-01

    Although the longlife potential of liquid-metal-cooled reactors (LMRs) has been only partially demonstrated, many factors point to the potential for exceptionally long life. EBR-II has the opportunity to become the first LMR to achieve an operational lifetime of 30 years or more. In 1984 a study of the extended-life potential of EBR-II identified the factors that contribute to the continued successful operation of EBR-II as a power reactor and experimental facility. Also identified were factors that could cause disruptions in the useful life of the facility. Although no factors were found that would inherently limit the life of EBR-II, measures were identified that could help ensure continued plant availability. These measures include the implementation of more effective surveillance, diagnostic, and control systems to complement the inherent safety and reliability features of EBR-II. An operating lifetime of well beyond 30 years is certainly feasible.

  10. EBR-II: twenty years of operating experience

    SciTech Connect

    Lentz, G.L.; Buschman, H.W.; Smith, R.N.

    1985-01-01

    Experimental Breeder Reactor No. 2 (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. For the last 20 years EBR-II has operated safely, has demonstrated stable operating characteristics, has shown excellent performance of its sodium components, and has had an excellent plant factor. These years of operating experience provide a valuable resource to the nuclear community for the development and design of future liquid metal fast reactors. This report provides a brief description of the EBR-II plant and its early operating experience, describes some recent problems of interest to the nuclear community, and also mentions some of the significant operating achievements of EBR-II. Finally, a few words and speculations on EBR-II's future are offered. 4 figs., 1 tab.

  11. Fault tree analysis of the EBR-II reactor shutdown system

    SciTech Connect

    Kamal, S.A.; Hill, D.J.

    1992-01-01

    As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized in order to reduce scram failure probability. Therefore, heavy emphasis is placed in the fault trees on the common cause failures (CCFs) among similar mechanical components of the control and safety rods and among similar electrical components in redundant detection channels and shutdown strings. Generic beta-factors that cover all types of similar components and reflect redundancy level are used to model the CCFs. Human errors are addressed in the fault trees in two major areas: errors that would prevent the automatic scram channels from detecting the abnormal events and errors that would prevent utilization of the manual scram capability. The fault tree analysis of the EBR-II shutdown system has provided not only a systematic process for calculating the probabilities of system failures but also useful insights into the system and how its elements interact during transient events that require shutdown.

  12. Fault tree analysis of the EBR-II reactor shutdown system

    SciTech Connect

    Kamal, S.A.; Hill, D.J.

    1992-12-01

    As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized in order to reduce scram failure probability. Therefore, heavy emphasis is placed in the fault trees on the common cause failures (CCFs) among similar mechanical components of the control and safety rods and among similar electrical components in redundant detection channels and shutdown strings. Generic beta-factors that cover all types of similar components and reflect redundancy level are used to model the CCFs. Human errors are addressed in the fault trees in two major areas: errors that would prevent the automatic scram channels from detecting the abnormal events and errors that would prevent utilization of the manual scram capability. The fault tree analysis of the EBR-II shutdown system has provided not only a systematic process for calculating the probabilities of system failures but also useful insights into the system and how its elements interact during transient events that require shutdown.

  13. Degradation of EBR-II driver fuel during wet storage

    SciTech Connect

    Pahl, R. G.

    2000-03-09

    Characterization data are reported for sodium bonded EBR-II reactor fuel which had been stored underwater in containers since the 1981--1982 timeframe. Ten stainless steel storage containers, which had leaked water during storage due to improper sealing, were retrieved from the ICPP-603 storage basin at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide sludge filled the bottom of the container. Headspace gas sampling determined that greater than 99% hydrogen was present. Cesium 137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a concentrated caustic solution of NaOH.

  14. System modelling to support accelerated fuel transfer rate at EBR-II

    SciTech Connect

    Imel, G.R.; Houshyar, A.; Planchon, H.P.; Cutforth, D.C.

    1995-06-01

    The Experimental Breeder Reactor-II (EBR-II) ia a 62.5 MW(th) liquid metal reactor operated by Argonne National Laboratory for The United States Department of Energy. The reactor is located near Idaho Falls, Idaho at the Argonne-West site (ANL-W). Full power operation was achieved in 1964,- the reactor operated continuously since that time until October 1994 in a variety of configurations depending on the programmatic mission. A three year program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. It was intended to operate the reactor during the three year blanket unloading program, followed by about a half year of driver fuel unloading. However, in the summer of 1994, Congress dictacted that EBR-II be shut down October 1, and complete defueling without operation. To assist in the planning for resources needed for this defueling campaign, a mathematical model of the fuel handling sequence was developed utilizing the appropriate reliability factors and inherent mm constraints of each stage of the process. The model allows predictions of transfer rates under different scenarios. Additionally, it has facilitated planning of maintenance activities, as well as optimization of resources regarding manpower and modification effort. The model and its application is described in this paper.

  15. The EBR-II X501 Minor Actinide Burning Experiment

    SciTech Connect

    M. K. Meyer; S. L. Hayes; W. J. Carmack; H. Tsai

    2009-07-01

    The X501 experiment was conducted in EBR-II as part of the IFR (Integral Fast Reactor) program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data, and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few minor actinide-bearing fuel irradiation tests conducted worldwide and knowledge can be gained by understanding the changes in fuel behavior due to addition of MA’s. Of primary interest are the affect of the MA’s on fuel-cladding-chemical-interaction, and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995-1996, and currently represents a set of observations rather than a complete understanding of fuel behavior. This paper provides a summary of the X501 fabrication, characterization, irradiation, and post irradiation examination.

  16. The EBR-II X501 Minor Actinide Burning Experiment

    SciTech Connect

    W. J. Carmack; M. K. Meyer; S. L. Hayes; H. Tsai

    2008-01-01

    The X501 experiment was conducted in EBR II as part of the Integral Fast Reactor program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few MA bearing fuel irradiation tests conducted worldwide, and knowledge can be gained by understanding the changes in fuel behavior due to addition of MAs. Of primary interest are the effect of the MAs on fuel cladding chemical interaction and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995–1996 and, currently, represents a set of observations rather than a complete understanding of fuel behavior. This report provides a summary of the X501 fabrication, characterization, irradiation, and post irradiation examination.

  17. The EBR-II X501 Minor Actinide Burning Experiment

    SciTech Connect

    Jon Carmack; S. L. Hayes; M. K. Meyer; H. Tsai

    2008-06-01

    The X501 experiment was conducted in EBR-II as part of the IFR (Integral Fast Reactor) program to demonstrate minor actinide burning through the use of a homogeneous recycle scheme. The X501 subassembly contained two metallic fuel elements loaded with relatively small quantities of americium and neptunium. Interest in the behavior of minor actinides (MA) during fuel irradiation has prompted further examination of existing X501 data, and generation of new data where needed in support of the U.S. waste transmutation effort. The X501 experiment is one of the few minor actinide-bearing fuel irradiation tests conducted worldwide and knowledge can be gained by understanding the changes in fuel behavior due to addition of MA’s. Of primary interest are the affect of the MA’s on fuel-cladding-chemical-interaction, and the redistribution behavior of americium. The quantity of helium gas release from the fuel and any effects of helium on fuel performance are also of interest. It must be stressed that information presented at this time is based on the limited PIE conducted in 1995-1996, and currently represents a set of observations rather than a complete understanding of fuel behavior.

  18. Benchmark specifications for EBR-II shutdown heat removal tests

    SciTech Connect

    Sofu, T.; Briggs, L. L.

    2012-07-01

    Argonne National Laboratory (ANL) is hosting an IAEA-coordinated research project on benchmark analyses of sodium-cooled fast reactor passive safety tests performed at the Experimental Breeder Reactor-II (EBR-II). The benchmark project involves analysis of a protected and an unprotected loss of flow tests conducted during an extensive testing program within the framework of the U.S. Integral Fast Reactor program to demonstrate the inherently safety features of EBR-II as a pool-type, sodium-cooled fast reactor prototype. The project is intended to improve the participants' design and safety analysis capabilities for sodium-cooled fast reactors through validation and qualification of safety analysis codes and methods. This paper provides a description of the EBR-II tests included in the program, and outlines the benchmark specifications being prepared to support the IAEA-coordinated research project. (authors)

  19. The COBRA-1B irradiation experiment in EBR-II

    SciTech Connect

    Tsai, H.; Hins, A.G.; Strain, R.V.; Smith, D.L.

    1994-09-01

    The objective of the forthcoming COBRA-1B experiment in EBR-II is to evaluate the effects of fast neutron irradiation on the physical and mechanical properties of candidate fusion structural materials. Of special interest in this experiment will be ITER-relevant temperature and exposure for the test specimens. Approximately 50% of the irradiation test volume will be devoted to vanadium-alloy specimens. Design of the COBRA-1B irradiation experiment began in this reporting period and is in progress. The target reactor insertion date for COBRA-1B is September 1994. Technical and programmatic feasibility approval for the experiment has been granted by EBR-II Operations.

  20. The EBR-II Probabilistic Risk Assessment: Results and insights

    SciTech Connect

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1993-12-31

    This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1. 6 10{sup {minus}6} yr{sup {minus}1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The probability of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquake) is 3.6 10{sup {minus}6} yr{sup {minus}1}. overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double, vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

  1. The EBR-II Probabilistic Risk Assessment: Results and insights

    SciTech Connect

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1993-01-01

    This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1. 6 10[sup [minus]6] yr[sup [minus]1], even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The probability of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquake) is 3.6 10[sup [minus]6] yr[sup [minus]1]. overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double, vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

  2. Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

    SciTech Connect

    Koenig, F.F.; Dean, E.M.

    1985-01-01

    In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program.

  3. Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

    SciTech Connect

    Lell, R. M.; Pope, C. L.

    2000-03-14

    Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles.

  4. Recent operating experiences and programs at EBR-II

    SciTech Connect

    Lentz, G.L.

    1984-01-01

    Experimental Breeder Reactor No. II (EBR-II) is a pool-type, unmoderated, sodium-cooled reactor with a design power of 62.5 MWt and an electrical generation capability of 20 MW. It has been operated by Argonne National Laboratory for the US government for almost 20 years. During that time, it has operated safely and has demonstrated stable operating characteristics, high availability, and excellent performance of its sodium components. The 20 years of operating experience of EBR-II is a valuable resource to the nuclear community for the development and design of future LMFBR's. Since past operating experience has been extensively reported, this report will focus on recent programs and events.

  5. Use of a steam leak simulator in EBR-II

    SciTech Connect

    McKee, J.M.; Osterhout, M.M. Batten, R.L.

    1984-01-01

    A steam leak simulator has been installed on EBR-II to periodically test and calibrate the steam-generator leak detection system. Measured amounts of molten anhydrous sodium hydroxide are injected at controlled rates simulating leaks in the range of 0.024 to 0.16 g H/sub 2/O/s. Experience with 11 injections over an 18 month period is described.

  6. Metallic fuels: The EBR-II legacy and recent advances

    SciTech Connect

    Douglas L. Porter; Steven L. Hayes; J. Rory Kennedy

    2012-09-01

    Experimental Breeder Reactor – II (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.

  7. A transient overpower experiment in EBR-II

    SciTech Connect

    Herzog, J.P.; Tsai, H.; Dean, E.M.; Aoyama, T.; Yamamoto, K.

    1994-03-01

    The TOPI-IE test was a transient overpower test on irradiate mixed-oxide fuel pins in the Experimental Breeder Reactor-II (EBR-II). The test, the fifth in a series, was part of a cooperative program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan to conduct operational transient testing on mixed-oxide fuel pins in the metal-fueled EBR-II. The principle objective of the TOPI-1E test was to assess breaching margins for irradiated mixed-oxide fuel pins over the Plant Protection System (PPS) thresholds during a slow, extended overpower transient. This paper describes the effect of the TOPI-1E experiment on reactor components and the impact of the experiment on the long-term operability of the reactor. The paper discusses the role that SASSYS played in the pre-test safety analysis of the experiment. The ability of SASSYS to model transient overpower events is detailed by comparisons of data from the experiment with computed reactor variables from a SASSYS post-test simulation of the experiment.

  8. An integrated plant-life extension program for EBR-II (Experimental Breeder Reactor)

    SciTech Connect

    King, R.W.

    1986-01-01

    An integrated plant-life extension program is being developed and implemented at EBR-II. The program plan has five primary areas of focus, and is structured to take advantage of inherent features of the liquid-metal-cooled reactor that give it potential for very long life. The program is centered around development and increased use of computer-based software systems for surveillance, diagnostics, prognostics, data handling, and knowledge transfer. Even though the program is only partially implemented, benefits are already being realized in the form of increased understanding of plant system status and performance due to development of diagnostic data-handling software for manipulation of plant sensor data, and improved force monitoring and protection of the remotely operated fuel handling system. The eventual integration of the elements of the program is a key feature that is expected to enhance the overall effectiveness of the program.

  9. Data handling at EBR-II (Experimental Breeder Reactor II) for advanced diagnostics and control work

    SciTech Connect

    Lindsay, R.W.; Schorzman, L.W.

    1988-01-01

    Improved control and diagnostics systems are being developed for nuclear and other applications. The Experimental Breeder Reactor II (EBR-II) Division of Argonne National Laboratory has embarked on a project to upgrade the EBR-II control and data handling systems. The nature of the work at EBR-II requires that reactor plant data be readily available for experimenters, and that the plant control systems be flexible to accommodate testing and development needs. In addition, operational concerns require that improved operator interfaces and computerized diagnostics be included in the reactor plant control system. The EBR-II systems have been upgraded to incorporate new data handling computers, new digital plant process controllers, and new displays and diagnostics are being developed and tested for permanent use. In addition, improved engineering surveillance will be possible with the new systems.

  10. Technical assessment of continued wet storage of EBR-II fuel

    SciTech Connect

    Pahl, R.G.; Franklin, E.M.; Ebner, M.A.

    1996-05-01

    A technical assessment of the continued wet storage of EBR-II fuel has been made. Previous experience has shown that in-basin cladding failure occurs by intergranular attack of sensitized cladding, likely assisted by basin water chlorides. Subsequent fuel oxidation is rapid and leads to loss of configuration and release of fission products. The current inventory of EBR-II fuel stored in the ICPP basins is at risk from similar corrosion reactions.

  11. Automated start-up of EBR-II: A preview

    SciTech Connect

    Kisner, R.A.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) and Argonne National Laboratory (ANL) are undertaking a joint project to develop control philosophies, strategies, and algorithms for computer control of the start-up mode of the Experimental Breeder Reactor II (EBR-II). The major objective of this project is to show that advanced liquid-metal reactor (LMR) plants can be operated from low power to full power using computer control. Development of an automated control system with this objective in view will help resolve specific issues and provide proof through demonstration that automatic control for plant start-up is feasible. This paper describes the approach that will be used to develop such a system and some of the features it is expected to have. Structured, rule-based methods, which will provide start-up capability from a variety of initial plant conditions and degrees of equipment operability, will be used for accomplishing mode changes during plant start-up. Several innovative features will be incorporated such as signal, command, and strategy validation to maximize reliability, flexibility to accommodate a wide range of plant conditions, and overall utility. Continuous control design will utilize figures of merit to evaluate how well the controller meets the mission requirements. The operator interface will have unique ''look ahead'' features to let the operator see what will happen next. 15 refs., 7 figs., 1 tab.

  12. EBR-II transient operation and test capabilities

    SciTech Connect

    Seidel, B.R.; Cutforth, D.C.; Lentz, G.L.; Lambert, J.D.B.

    1983-01-01

    Experimental fuel pins intended for eventual use in LMFBR's have been irradiated for many years in fast test reactors. A wealth of data have been obtained on their performance under steady-state conditions, and fuel-pin performance codes have been developed to predict their behavior. In parallel, safety tests of fuel pins to explore behavior under accident conditions have been performed in transient reactors like TREAT in the US, and CABRI in France. These two types of testing generally have had different aims and have tended to produce results which are not reconcilable with a common modeling code, such as a LIFE or COMETHE, in the middle ground between normal and off-normal conditions. But as the licensing and commercialization of LMFBR's approaches, the attention and needs of the fuel-pin designer and licenser have focused on this middle ground between steady-state and accident testing of fuel pins and subassemblies. Preparations and now capability for operational reliability testing at EBR-II have been the subject of papers at recent conferences. This paper updates the status of those preparations to the present time when the ORT program is about to begin.

  13. Application of PCT to the EBR II ceramic waste form.

    SciTech Connect

    Ebert, W. L.; Lewis, M. A.; Johnson, S. G.

    2002-01-10

    We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF.

  14. JAEA Fatigue Analysis of EBR-II Duplex Tubing

    SciTech Connect

    J. H. Jackson; D. L. Porter; W. R. Lloyd; N. Kisohara

    2011-03-01

    Small, notched three-point bend specimens machined from duplex tubes, which were extracted from an EBR-II superheater, were fatigued through the nickel interlayer to determine propensity for crack arrest within this interlayer. Several of these specimens were fatigued in the near threshold, and steady state regimes of Paris Law behavior. Additionally, two specimens were fatigued to the edge of the nickel interlayer and then monotonically loaded. Micro-hardness profiles of the nickel interlayer were also measured. Fatigue behavior was found to be similar to previous studies in that arrest was only noted in the near threshold Paris regime (attributed to the presence of voids) and in the steady state regime exhibited an acceleration of crack growth rate through the nickel interlayer followed by a slight retardation. Monotonic loading resulted in crack branching or delamination along the interlayer. Although archival material was not available for this study, the hardness of the nickel interlayer was determined to have been lowered slightly during service by comparison to the expected hardness of a similar nickel braze prepared as specified for fabrication of these tubes.

  15. Off-normal performance of EBR-II (Experimental Breeder Reactor) driver fuel

    SciTech Connect

    Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

    1986-09-01

    The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel.

  16. Vanadium alloy irradiation experiment X530 in EBR-II{sup *}

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.

    1995-04-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994.

  17. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    SciTech Connect

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-05-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components. Operation of EBR-II has produced a wealth of information. As an irradiation facility, EBR-II has generated specific information on the behavior of oxide, carbide, and metallic fuels. As an LMFBR power plant, EBR-II has produced general information related to plant-systems and equipment design, plant safety, plant availability, and plant maintenance.

  18. Remote, under-sodium fuel handling experience at EBR-II

    SciTech Connect

    King, R.W.; Planchon, H.P.

    1995-05-01

    The EBR-II is a pool-type design; the reactor fuel handling components and entire primary-sodium coolant system are submerged in the primary tank, which is 26 feet in diameter, 26 feet high, and contains 86,000 gallons of sodium. Since the reactor is submerged in sodium, fuel handling operations must be performed blind, making exact positioning and precision control of the fuel handling system components essential. EBR-II operated for 30 years, and the fuel handling system has performed approximately 25,000 fuel transfer operations in that time. Due to termination of the IFR program, EBR-II was shut down on September 30, 1994. In preparation for decommissioning, all fuel in the reactor will be transferred out of EBR-II to interim storage. This intensive fuel handling campaign will last approximately two years, and the number of transfers will be equivalent to the fuel handling done over about nine years of normal reactor operation. With this demand on the system, system reliability will be extremely important. Because of this increased demand, and considering that the system has been operating for about 32 years, system upgrades to increase reliability and efficiency are proceeding. Upgrades to the system to install new digital, solid state controls, and to take advantage of new visualization technology, are underway. Future reactor designs using liquid metal coolant will be able to incorporate imaging technology now being investigated, such as ultraviolet laser imaging and ultrasonic imaging.

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

    SciTech Connect

    Steven R. Sherman; Collin J. Knight

    2006-03-01

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

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

    SciTech Connect

    Steven R. Sherman

    2005-04-01

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

  1. Thermal-hydraulic-structural behavior of the EBR-II IHX for overpower transients

    SciTech Connect

    Mohr, D.; Chang, L.K.; Lee, M.J.; Feldman, E.E.

    1982-01-01

    A detailed study has been made of the effects of the Operational Reliability Testing (ORT) program on major plant components of the Experimental Breeder Reactor No. II (EBR-II). This paper describes the integrated thermal-hydraulic-structural analyses conducted for the intermediate heat exchanger (IHX) with the aid of the NATDEMO, THTB, and ANSYS codes. An extensive analysis revealed the stress limiting area to be the junction between the upper head and upper tube sheet. The analyses indicate, however, that the EBR-II IHX, the major plant component most affected by the ORT program, will be able to withstand the thermal stress and accumulated fatigue damage during the lifetime of the plant including the ORT program.

  2. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  3. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-20

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  4. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    NASA Astrophysics Data System (ADS)

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  5. Experimental Breeder Reactor II (EBR-II): Instrumentation for core surveillance

    SciTech Connect

    Christensen, L.J.

    1989-01-01

    EBR-II has operated for 25 years in support of several major programs. During this time period, several of the original, non-replaceable, flow sensors, RDT sensors and thermocouples have failed in the primary system. This has led to the development of new sensors and the use of calculated values using computer models of the plant. It is important for the next generation of LMR reactors to minimize or eliminate the use of non-replaceable sensors. EBR-II is perhaps the best modeled reactor in the world, thanks to a dedicated T-H analysis program. The success of this program relied on excellent measurements of temperature and flow in subassemblies in the core. The instrumented subassemblies of the XX series provided that measurement capability. From this test series, EBR-II calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. From this, reactor designers can now design with confidence, inherently safe reactors. 11 refs., 8 figs.

  6. Swelling, microstructural development and helium effects in type 316 stainless steel irradiated in HFIR and EBR-II

    SciTech Connect

    Maziasz, P.J.; Grossbeck, M.L.

    1981-01-01

    This work examines the swelling and microstructural development of a single heat of 20%-cold-worked type 316 stainless steel irradiated to produce displacement damage and a high, continuous helium generation rate, in the High Flux Isotope Reactor (HFIR). Similar irradiation of the same heat of steel in the Experimental Breeder Reactor (EBR)-II is used as a base line for comparing displacement damage accompanying a very low continuous helium generation rate. At temperatures above and below the void swelling regime (approx. 350 to 625/sup 0/C) swelling is greater in HFIR than in EBR-II. In the temprature range of 350 to 625/sup 0/C, cavity formation, precipitation and dislocation recovery are both enhanced and accelerated in HFIR, often causing swelling at lower dose than in EBR-II. In HFIR, however, cavities appear to be bubbles rather than voids. They are about 10 times smaller and 20 to 50 times more numerous than voids in EBR-II. Thus, the swelling becomes greater in EBR-II than in HFIR for 20%-CW 316 in the void swelling temperature ranges as fluence increases. Such differences in swelling and microstructural behavior must be understood in order to anticipate the behavior of materials during fusion irradiation.

  7. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    SciTech Connect

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  8. Microchemical and microstructural evolution of AISI 304 stainless steel irradiated in EBR-II at PWR-relevant dpa rates

    NASA Astrophysics Data System (ADS)

    Dong, Y.; Sencer, B. H.; Garner, F. A.; Marquis, E. A.

    2015-12-01

    AISI 304 stainless steel was irradiated at 416 °C and 450 °C at a 4.4 × 10-9 and 3.05 × 10-7 dpa/s to ∼0.4 and ∼28 dpa, respectively, in the reflector of the EBR-II fast reactor. Both unirradiated and irradiated conditions were examined using standard and scanning transmission electron microscopy, energy dispersive spectroscopy, and atom probe tomography on very small specimens produced by focused ion beam milling. These results are compared with previous electron microscopy examination of 3 mm disks from essentially the same material. By comparing a very low dose specimen with a much higher dose specimen, both derived from a single reactor assembly, it has been demonstrated that the coupled microstructural and microchemical evolution of dislocation loops and other sinks begins very early, with elemental segregation producing at these sinks what appears to be measurable precursors to fully formed precipitates found at higher doses. The nature of these sinks and their possible precursors are examined in detail.

  9. Operating experience of the EBR-II intermediate heat exchanger and the steam generator system

    SciTech Connect

    Buschman, H.W.; Longua, K.J.; Penney, W.H.

    1983-01-01

    Experimental Breeder Reactor-II (EBR-II) is an experimental liquid metal fast breeder reactor located at the Idaho National Engineering Laboratory. It consists of an unmoderated, heterogeneous, sodium-cooled reactor with a nominal thermal power output of 62.5 MW; an intermediate closed loop of secondary sodium coolant; and a steam plant that produces 20 MW of electrical power through a conventional turbine generator. The EBR-II heat transport system continues to operate satisfactorily after 18 years. This represents about 89,000 hours of steaming, which results in a total integrated thermal power production of about 215,000 MWd. In this time, the steam generator has experienced over 580 plant startups and 349 reactor scrams. The plant capacity factor for the past five years has been in excess of 70%, and in fact has averaged almost 60% over the last thirteen years. This excellent record is partly attributable to the trouble-free operation of the steam generator which, aside from an initial construction tube-to-tubesheet weld defect, has had a plant availability of 100%.

  10. Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report

    SciTech Connect

    Pahl, R. G.

    2000-04-17

    Characterization data is reported for sodium bonded Experimental Breeder Reactor II (EBR-II) fuel which had been stored underwater in containers since the late 1970's. Sixteen stainless steel storage containers were retrieved from the ICPP-603 storage pool at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. Ten of the containers had leaked water due to improper sealing. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide particulate formed and filled the bottom of the container. Headspace gas analysis determined that greater than 99% hydrogen was present. Cesium-137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a caustic solution of NaOH.

  11. Expert system applications in support of system diagnostics and prognostics at EBR-II

    SciTech Connect

    Lehto, W.K.; Gross, K.C.; Argonne National Lab., IL )

    1989-01-01

    Expert systems have been developed to aid in the monitoring and diagnostics of the Experimental Breeder Reactor-II (EBR-II) at the Idaho National Engineering Laboratory (INEL) in Idaho Falls, Idaho. Systems have been developed for failed fuel surveillance and diagnostics and reactor coolant pump monitoring and diagnostics. A third project is being done jointly by ANL-W and EG G Idaho to develop a transient analysis system to enhance overall plant diagnostic and prognostic capability. The failed fuel surveillance and diagnosis system monitors, processes, and interprets information from nine key plant sensors. It displays to the reactor operator diagnostic information needed to make proper decisions regarding technical specification conformance during reactor operation with failed fuel. 8 refs., 9 figs., 2 tabs.

  12. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    SciTech Connect

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-01-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components.

  13. Response of EBR-II to a complete loss of primary forced flow during power operation

    SciTech Connect

    Singer, R.M.; Gillette, J.L.; Mohr, D.; Tokar, J.V.; Sullivan, J.E.; Dean, E.M.

    1980-01-01

    Detailed measurements of the thermal, hydraulic, and neutronic response of EBR-II to a complete loss of primary forced flow followed by a PPS-activated scram are presented. The experimental results clearly indicate a smooth transition to natural convective flow with a quite modest incore temperature transient. The accompanying calculations using the NATDEMO code agree quite well with the measured temperatures and flow rates throughout the primary system. The only region of the plant where a significant discrepancy between the measurements and calculations occurred was in the IHX. The reasons for this result could not be definitively determined, but it is speculated that the one-dimensional assumptions used in the modeling may not be valid in the IHX during buoyancy driver flows.

  14. The EBR-II materials-surveillance program. 5: Results of SURV-5.

    SciTech Connect

    Ruther, W.E.; Staffon, J.D.; Carlson, B.G.; Allen, T.R.

    1998-01-01

    In March of 1965, a set of surveillance (SURV) samples was placed in the EBR-II reactor to determine the effect of irradiation, thermal aging, and sodium corrosion on reactor materials. Eight subassemblies were placed into row 12 positions of EBR-II to determine the effect of irradiation at 370 C. Two subassemblies were placed into the primary sodium basket to determine the effect of thermal aging at 370 C. One half of all samples were exposed to primary system sodium while one half were sealed in capsules with a helium atmosphere. Fifteen different structural materials were tested in the SURV program. In this work, the properties of these materials irradiated at 370 C to a total fluence of 3.2 {times} 10{sup 22} n/cm{sup 2} were determined. These materials are the fifth set of irradiated subassemblies to be examined as part of the SURV program (SURV-5). The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, and fracture resistance. Of all the alloys examined in SURV-5, only Berylco-25 showed any significant weight loss. Stainless steel (both 304 and 347) had the largest density decrease, although the density decrease from irradiation for all alloys was less than 0.4 percent. The microstructure of both Berylco-25 and the aluminum-bronze alloy was altered significantly. Iron- and nickel-base alloys showed little change in microstructure. Austenitic steels (304 and 347) harden with irradiation. The hardness of Inconel X750 did not change significantly with irradiation. The ultimate tensile strength of Inconel X750, 304 stainless steel, 420 stainless steel and welded 304 changed little due to a fluence increase from 2.2 {times} 10{sup 22} n/cm{sup 2} (the maximum fluence of the SURV-4 samples) to 3.2 {times} 10{sup 22} n/cm{sup 2}.

  15. Comparison of swelling and cavity microstructural development for type 316 stainless steel irradiated in EBR-II and HFIR

    SciTech Connect

    Maziasz, P.J.

    1983-01-01

    Comparison of swelling and cavity microstructures for one heat of 20% cold-worked (CW) type 316 stainless steel (316) irradiated at 500 to 650/sup 0/C in EBR-II (up to 75 dpa) and HFIR (up to 61 dpa) suggests that void growth and swelling are suppressed by the higher helium generation found in HFIR. Instead of voids, many small bubbles develop in the CW 316 in HFIR and resist conversion to voids. However, similar comparison of solution-annealed (SA) 316 irradiated in EBR-II and HFIR at 500 to 550/sup 0/C leads to an opposite conclusion; void swelling is enhanced by helium in HFIR. Many more bubbles nucleate in SA 316 at low fluence in HFIR compared to EBR-II, but bimodel distributions and rapid coarsening eventually lead to high swelling due to high concentrations of matrix ands precipitate-associated voids in HFIR. A key to the swelling resistance of the CW 316 in HFIR appears to be the development of a sufficiently cavity-dominated sink system in the early stages of evolution.

  16. Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes

    SciTech Connect

    Terry C. Totemeier

    2006-12-01

    Fatigue crack growth tests were performed on 2¼Cr-1Mo steel specimens machined from ex-service Experimental Breeder Reactor – II (EBR-II) superheater duplex tubes. The tubes had been metallurgically bonded with a 100 µm thick Ni interlayer; the specimens incorporated this bond layer. Tests were performed at room temperature in air and at 400°C in air and humid Ar; cracks were grown at varied levels of constant ?K. Crack growth tests at a range of ?K were also performed on specimens machined from the shell of the superheater. In all conditions the presence of the Ni interlayer was found to result in a net retardation of growth as the crack passed through the interlayer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous interlayer. Full crack arrest was only observed in a single test performed at near-threshold ?K level (12 MPa?m) at 400°C. In this case the crack tip was blunted by oxidation of the base steel at the steel-interlayer interface.

  17. SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS

    SciTech Connect

    Matthew C. Morrison; Kenneth J. Bateman; Michael F. Simpson

    2010-11-01

    ABSTRACT SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS Matthew C. Morrison, Kenneth J. Bateman, Michael F. Simpson Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 The ceramic waste process is the intended method for disposing of waste salt electrolyte, which contains fission products from the fuel-processing electrorefiners (ER) at the INL. When mixed and processed with other materials, the waste salt can be stored in a durable ceramic waste form (CWF). The development of the CWF has recently progressed from small-scale testing and characterization to full-scale implementation and experimentation using surrogate materials in lieu of the ER electrolyte. Two full-scale (378 kg and 383 kg) CWF test runs have been successfully completed with final densities of 2.2 g/cm3 and 2.1 g/cm3, respectively. The purpose of the first CWF was to establish material preparation parameters. The emphasis of the second pre-qualification test run was to evaluate a preliminary multi-section CWF container design. Other considerations were to finalize material preparation parameters, measure the material height as it consolidates in the furnace, and identify when cracking occurs during the CWF cooldown process.

  18. Midplane and off-midplane axial leakage simulation of heterogeneous subassemblies in EBR-II

    SciTech Connect

    Grimm, K.N.; Meneghetti, D.

    1984-01-01

    Generally EBR-II XY geometry and one-dimensional (1D) cylindrical neutron flux calculations using transport theory analysis assume energy independent DB/sup 2/-type absorptions to simulate effects of axial leakages. This assumption, while generally resulting in satisfactory eigenvalues and high- and intermediate-energy flux spectra, gives large errors in the low-energy flux spectra where the flux levels are smaller. These midplane errors, and more importantly the off-midplane errors, can be reduced by using a more realistic leakage model: space and energy dependent leakage absorption cross sections. Analyses have been reported in which transport theory methods using row-wise azimuthally-homogeneous RZ-geometry boundary angular fluxes to calculate space and energy dependent leakage absorptions which were then used in subsequent 1D cylindrical simulations of RZ calculations. The present paper extends the study to include heterogeneous core loading configurations. This study contains modeling of heterogeneous XYZ loadings using heterogeneous XY geometry and space and energy dependent leakage absorptions. Because of the complexities arising from the three-dimensional analysis, the results presented here use diffusion theory. Although the actual negative leakage absorption values can be used in the CITATION diffusion theory code, it was found that the ..sigma../sub s/(1..-->..g) method gave better results in the core region of these studies.

  19. Visual imagery and the user model applied to fuel handling at EBR-II

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-06-01

    The material presented in this paper is based on two studies involving visual display designs and the user`s perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ``comfort parameters`` and ``perspective reality`` of the user`s model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator`s perspective of the fuel handling system of Argonne`s Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting.

  20. Tensile properties of vanadium alloys irradiated at 390{degrees}C in EBR-II

    SciTech Connect

    Chung, H.M.; Tsai, H.C.; Nowicki, L.J.

    1997-08-01

    Vanadium alloys were irradiated in Li-bonded stainless steel capsules to {approx}390{degrees}C in the EBR-II X-530 experiment. This report presents results of postirradiation tests of tensile properties of two large-scale (100 and 500 kg) heats of V-4Cr-Ti and laboratory (15-30 kg) heats of boron-doped V-4Cr-4Ti, V-8Cr-6Ti, V-5Ti, and V-3Ti-1Si alloys. Tensile specimens, divided into two groups, were irradiated in two different capsules under nominally similar conditions. The 500-kg heat (No. 832665) and the 100-kg heat (VX-8) of V-4Cr-4Ti irradiated in one of the subcapsules exhibited complete loss of work-hardening capability, which was manifested by very low uniform plastic strain. In contrast, the 100-kg heat of V-4Cr-4Ti irradiated in another subcapsule exhibited good tensile properties (uniform plastic strain 2.8-4.0%). A laboratory heat of V-3Ti-1Si irradiated in the latter subcapsule also exhibited good tensile properties. These results indicate that work-hardening capability at low irradiation temperatures varies significantly from heat to heat and is influenced by nominally small differences in irradiation conditions.

  1. Initiating the D&D Project for the EBR-II

    SciTech Connect

    Rick Demmer

    2010-08-01

    A novel decommissioning project is underway to close the Experimental Breeder Reactor-II (EBR-II) “fast” reactor at the Idaho National Laboratory (INL), Materials and Fuels Complex (MFC) facility near Idaho Falls, ID. The facility was placed in cold shutdown in 1994 and work began on the removal of the metallic sodium coolant. The bulk of the sodium was drained and treated beginning in 2001. The residual sodium heel was chemically passivated to render it less reactive in 2005 using a novel carbon dioxide treatment. Approximately 700 kg of metallic sodium and 3500 kg of sodium bicarbonate remain in the facility. A RCRA Waste Treatment Permit, issued in 2002 by the State of Idaho Department of Environmental Quality, requires annual progress toward closure of the facility, and that all regulated materials be removed or deactivated, and the waste products removed by 2022. The baseline sodium removal technology would result in about 100,000 gallons of low-level waste solution requiring treatment along with separate handling of the large components (intermediate heat exchanger, rotating plug, etc) outside of the primary tank.

  2. Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes

    SciTech Connect

    T.C. Totemeier; D.M. Wachs; D.L. Porter

    2008-05-01

    Fatigue crack growth and impact tests were performed on 2¼Cr-1Mo steel specimens machined from ex-service Experimental Breeder Reactor – II (EBR-II) superheater duplex tubes. The tubes had been metallurgically bonded with a 100 µm thick Ni layer; the specimens incorporated this bond layer. Impact tests were performed at temperatures from –50 to 400°C; cracks propagating from the V-notch were arrested by delamination at the bond layer for all tests with one exception at –50°C. Fatigue crack growth tests were performed at room temperature in air and at 400°C in air and humid Ar; cracks were grown at varied levels of constant ?K. In all conditions the presence of the Ni bond layer was found to result in a net retardation of growth as the crack passed through the layer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous bond layer. Full crack arrest was only observed in a single test performed at near-threshold ?K level (12 MPa?m) at 400°C. In this case the crack tip was blunted by oxidation of the base steel at the steel-nickel interface.

  3. Static leach tests with the EBR-II metallic waste form.

    SciTech Connect

    Ebert, W. L.; Lewis, M. A.; Barber, T. L.; DiSanto, T.; Johnson, S. G.

    2004-01-09

    A metallic waste form (MWF) will be used to immobilize contaminated cladding hulls recovered after electrometallurgical treatment of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor-II (EBR-II). Tests were conducted to determine if the high-level waste (HLW) glass degradation model developed for total system performance assessment (TSPA) calculations for the Yucca Mountain repository system can be used to represent the degradation of disposed MWF. Static tests were conducted at 50, 70, and 90 C with monolithic samples of MWF in pH buffer solutions spiked with NaCl at a MWF surface-to-solution volume ratio of about 200 m{sup -1}. Test specimens were prepared from a surrogate MWF ingot containing about 10 mass% U. Solutions were exchanged after 14, 28, and 70 days. The cumulative amount of U released into solution through 70 days was used to calculate the MWF degradation rate for each test condition. The rate was independent of temperature. The rate was highest in acidic solutions, lowest in neutral solutions, and intermediate in alkaline solutions. The uranium release rate from a breached canister, which is the product of the MWF degradation rate and the surface area of two MWF ingots in a canister, was compared with the release rate calculated with the HLW glass degradation model for a glass log at the same temperature and pH values. The uranium release rates measured for MWF are less than the degradation rates calculated for HLW glass (compared on a mass per time basis).

  4. Prediction of stainless steel activation in experimental breeder reactor 2 (EBR-II) reflector and blanket subassemblies

    SciTech Connect

    Bunde, K.A.

    1996-12-31

    Stainless steel structural components in nuclear reactors become radioactive wastes when no longer useful. Prior to disposal, certain physical attributes must be analyzed. These attributes include structural integrity, chemical stability, and the radioactive material content among others. The focus of this work is the estimation of the radioactive material content of stainless steel wastes from a research reactor operated by Argonne National Laboratory.

  5. Flibe blanket concept for transmuting transuranic elements and long lived fission products.

    SciTech Connect

    Gohar, Y.

    2000-11-15

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  6. Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM

    SciTech Connect

    Hu, Rui; Sumner, Tyler S.

    2016-01-01

    An advanced system analysis tool SAM is being developed for fast-running, improved-fidelity, and wholeplant transient analyses at Argonne National Laboratory under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. As an important part of code development, companion validation activities are being conducted to ensure the performance and validity of the SAM code. This paper presents the benchmark simulations of two EBR-II tests, SHRT-45R and BOP- 302R, whose data are available through the support of DOE-NE’s Advanced Reactor Technology (ART) program. The code predictions of major primary coolant system parameter are compared with the test results. Additionally, the SAS4A/SASSYS-1 code simulation results are also included for a code-to-code comparison.

  7. Experimental and analytical study of loss-of-flow transients in EBR-II occurring at decay power levels

    SciTech Connect

    Chang, L.K.; Mohr, D.; Feldman, E.E.; Betten, P.R.; Planchon, H.P.

    1985-01-01

    A series of eight loss-of-flow (LOF) tests have been conducted in EBR-II to study the transition between forced and natural convective flows following a variety of loss-of-primary-pumping power conditions from decay heat levels. Comparisons of measurements and pretest/posttest predictions were made on a selected test. Good agreements between measurements and predictions was found prior to and just after the flow reaching its minimum, but the agreement is not as good after that point. The temperatures are consistent with the flow response and the assumed decay power. The measured results indicate that the flows of driver and the instrumented subassemblies are too much in the analytical model in the natural convective region. Although a parametric study on secondary flow, turbulent-laminar flow transition, heat transfer ability of the intermediate heat exchange at low flow and flow mixing in the primary tank has been performed to determine their effects on the flow, the cause of the discrepancy at very low flow level is still unknown.

  8. Radiation Damage Calculations for the FUBR and BEATRIX Irradiations of Lithium Compounds in EBR-II and FFTF

    SciTech Connect

    LR Greenwood

    1999-06-17

    The Fusion Breeder Reactor (FUBR) and Breeder Exchange Matrix (BEATRIX) experiments were cooperative efforts by members of the International Energy Agency to investigate the irradiation behavior of solid breeder materials for tritium production to support future fusion reactors. Lithium ceramic materials including Li{sub 2}O, LiAlO{sub 2}, Li{sub 4}SiO{sub 4}, and Li{sub 2}ZrO{sub 3} with varying {sup 6}Li enrichments from 0 to 95% were irradiated in a series of experiments in the Experimental Breeder Reactor (EBR II) and in the Fast Flux Test Facility (FFTF) over a period of about 10 years from 1982 to 1992. These experiments were characterized in terms of the nominal fast neutron fluences and measured {sup 6}Li burnup factors, as determined by either mass spectrometry or helium measurements. Radiation damage in these compounds is caused by both the {sup 6}Li-burnup reaction and by all other possible neutron reactions with the atoms in the compound materials. In this report, displacements per atom (dpa) values have been calculated for each type of material in each of the various irradiations that were conducted. Values up to 11% {sup 6}Li-burnup and 130 dpa are predicted for the longest irradiations. The dpa cross sections were calculated for each compound using the SPECOMP computer code. Details of the dpa calculations are presented in the report. Total dpa factors were determined with the SPECTER computer code by averaging the dpa cross sections over the measured or calculated neutron flux spectra for each series of irradiations. Using these new calculations, previously measured radiation damage effects in these lithium compounds can be compared or correlated with other irradiation data on the basis of the dpa factor as well as {sup 6}Li-burnup.

  9. Microstructural characterization and density change of 304 stainless steel reflector blocks after long-term irradiation in EBR-II

    NASA Astrophysics Data System (ADS)

    Huang, Y.; Wiezorek, J. M. K.; Garner, F. A.; Freyer, P. D.; Okita, T.; Sagisaka, M.; Isobe, Y.; Allen, T. R.

    2015-10-01

    While thin reactor structural components such as cladding and ducts do not experience significant gradients in dpa rate, gamma heating rate, temperature or stress, thick components can develop strong local variations in void swelling and irradiation creep in response to gradients in these variables. In this study we conducted microstructural investigations by transmission electron microscopy of two 52 mm thick 304-type stainless steel hex-blocks irradiated for 12 years in the EBR-II reactor with accumulated doses ranging from ∼0.4 to 33 dpa. Spatial variations in the populations of voids, precipitates, Frank loops and dislocation lines have been determined for 304 stainless steel sections exposed to different temperatures, different dpa levels and at different dpa rates, demonstrating the existence of spatial gradients in the resulting void swelling. The microstructural measurements compare very well with complementary density change measurements regarding void swelling gradients in the 304 stainless steel hex-block components. The TEM studies revealed that the original cold-worked-state microstructure of the unirradiated blocks was completely erased by irradiation, replaced by high densities of interstitial Frank loops, voids and carbide precipitates at both the lowest and highest doses. At large dose levels the amount of volumetric void swelling correlated directly with the gamma heating gradient-related temperature increase (e.g. for 28 dpa, ∼2% swelling at 418 °C and ∼2.9% swelling at 448 °C). Under approximately iso-thermal local conditions, volumetric void swelling was found to increase with dose level (e.g. ∼0.2% swelling at 0.4 dpa, ∼0.5% swelling at 4 dpa and ∼2% swelling at 28 dpa). Carbide precipitate formation levels were found to be relatively independent of both dpa level and temperature and induced a measurable densification. Void swelling was dominant at the higher dose levels and caused measurable decreases in density. Void swelling

  10. The dependence of helium generation rate on nickel content of Fe Cr Ni alloys irradiated to high dpa levels in EBR-II

    NASA Astrophysics Data System (ADS)

    Garner, F. A.; Oliver, B. M.; Greenwood, L. R.

    1998-10-01

    Fusion-relevant helium-effects experiments conducted on austenitic steels in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF) fast reactor had to recognize the contributions of both the high neutron energy (n,α) reactions and that of the 58Ni(n,γ) 59Ni(n,α) reaction sequence with low energy neutrons. An experiment conducted in the harder neutron spectra found within the core of Experimental Breeder Reactor-II (EBR-II) has shown that the helium in this reactor was generated almost exclusively from the interaction of high energy neutrons with the natural isotopes of nickel. There was very little contribution from 59Ni. The helium production was found to scale directly with the nickel content over the range 25-75% Ni. Even at very high neutron exposures, the helium production in such reactors can be predicted within 5% accuracy on the basis of high energy reactions, as demonstrated by an experiment conducted on three Fe-15Cr-Ni ternary alloys irradiated to doses of 75-131 dpa in EBR-II.

  11. Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

    SciTech Connect

    S. D. Herrmann; L. A. Wurth; N. J. Gese

    2013-09-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

  12. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    SciTech Connect

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  13. Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

    SciTech Connect

    Hermann, S.D.; Gese, N.J.; Wurth, L.A.

    2013-07-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.

  14. Validation of the integration of CFD and SAS4A/SASSYS-1: Analysis of EBR-II shutdown heat removal test 17

    SciTech Connect

    Thomas, J. W.; Fanning, T. H.; Vilim, R.; Briggs, L. L.

    2012-07-01

    Recent analyses have demonstrated the need to model multidimensional phenomena, particularly thermal stratification in outlet plena, during safety analyses of loss-of-flow transients of certain liquid-metal cooled reactor designs. Therefore, Argonne's reactor systems safety code SAS4A/SASSYS-1 is being enhanced by integrating 3D computational fluid dynamics models of the plena. A validation exercise of the new tool is being performed by analyzing the protected loss-of-flow event demonstrated by the EBR-II Shutdown Heat Removal Test 17. In this analysis, the behavior of the coolant in the cold pool is modeled using the CFD code STAR-CCM+, while the remainder of the cooling system and the reactor core are modeled with SAS4A/SASSYS-1. This paper summarizes the code integration strategy and provides the predicted 3D temperature and velocity distributions inside the cold pool during SHRT-17. The results of the coupled analysis should be considered preliminary at this stage, as the exercise pointed to the need to improve the CFD model of the cold pool tank. (authors)

  15. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    SciTech Connect

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  16. Blanket technology workshop report

    NASA Technical Reports Server (NTRS)

    Scott-Monck, J. A.

    1980-01-01

    The solar array blanket, defined as a substrate covered with interconnected and glassed solar cells, but excluding the necessary support structure, deployment, and orientation devices is considered. The interactions between the blanket and the structure that is used to package, deploy, support and, if necessary restow it, are addressed along with systems constraints such as spacecraft configuration, size, and payload requirements. The influence on blanket design is emphasized. The three main mission classes considered are low Earth orbital (LEO), intermediate, or LEO to GEO transfer, and geosynchronous (GEO). Although interplanetary missions could be considered to be a separate class, their requirements, primarily power per unit mass, are generally close enough to geosynchronous missions to allow this mission class to be included within the third type. Examination of the critical elements of each class coupled with considerations of the shuttle capabilities is used to define the type of blanket technology most likely required to support missions that will be flown starting in 1990.

  17. Thermomechanical analysis of the ITER breeding blanket

    SciTech Connect

    Majumdar, S.; Gruhn, H.; Gohar, Y.; Giegerich, M.

    1997-03-01

    Thermomechanical performance of the ITER breeding blanket is an important design issue because it requires first, that the thermal expansion mismatch between the blanket structure and the blankets internals (such as, beryllium multiplier and tritium breeders) can be accommodated without creating high stresses, and second, that the thermomechanical deformation of various interfaces within the blanket does not create high resistance to heat flow and consequent unacceptably high temperatures in the blanket materials. Thermomechanical analysis of a single beryllium block sandwiched between two stainless steel plates was carried out using the finite element code ABAQUS to illustrate the importance of elastic deformation on the temperature distributions. Such an analysis for the whole ITER blanket needs to be conducted in the future. Uncertainties in the thermomechanical contact analysis can be reduced by bonding the beryllium blocks to the stainless steel plates by a thin soft interfacial layer.

  18. TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

    NASA Astrophysics Data System (ADS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2010-10-01

    A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10 25 n m -2 ( E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.

  19. Thermionic fuel element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. Data from accelerated tests in FETF and EBR-II show component lifetimes longer than 7 yr. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are currently operating in the TRIGA reactor, the oldest having accumulated 15,000 hr of irradiation as of 1 October 1990.

  20. Thermionic fuel element verification program—overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dutt, Dale S.; Dahlberg, Richard C.; Wood, John T.

    1991-01-01

    TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. It is jointly funded by SIDO and DOE. Data from accelerated tests in FFTF and EBR-II show component lifetimes longer than 7 years. Alumina insulators have shown good performance at high fast fluence. Graphite-cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are current operating in the TRIGA reactor, the oldest having accumulated 15,000 hours of irradiation as of 1 October 1990.

  1. ITER convertible blanket evaluation

    SciTech Connect

    Wong, C.P.C.; Cheng, E.

    1995-09-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate.

  2. Materials for breeding blankets

    SciTech Connect

    Mattas, R.F.; Billone, M.C.

    1995-09-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified.

  3. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    SciTech Connect

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-04-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability.

  4. Multivariable optimization of fusion reactor blankets

    SciTech Connect

    Meier, W.R.

    1984-04-01

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% /sup 6/Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO/sub 2/ breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO/sub 2/ breeding blanket enriched to 34% /sup 6/Li.

  5. Composite flexible blanket insulation

    NASA Technical Reports Server (NTRS)

    Kourtides, Demetrius A. (Inventor); Lowe, David M. (Inventor)

    1994-01-01

    An improved composite flexible blanket insulation is presented comprising top silicon carbide having an interlock design, wherein the reflective shield is composed of single or double aluminized polyimide and wherein the polyimide film has a honeycomb pattern.

  6. Line Blanketing in Przybylski's Star

    NASA Astrophysics Data System (ADS)

    Cowley, C. R.; Kupka, F.; Mathys, G.

    1999-12-01

    Przybylski's star (HD 101065) may be the most heavily blanketed star known. It therefore provides a test of our techniques for line blanketing. The current abstract draws on a paper in preparation by CRC, T. Ryabchikova, F. Kupka, G. Mathys, and D. J. Bord, based on ESO spectra obtained by GM. Unfortunately, the atomic species that provide the majority of the line blanketing in Przybylski's star does not have enough atomic data for realistic calculations of the blanketing. We therefore discuss three models in which iron-group elements were articifically elevated in abundance in the calculation of opacity used to construct the models. We thank Drs. R. L. Kurucz, and Bengt Edvardsson for calculating respectively Models 1 (dashed [Fe/H]=+3) and 2 (dot-dash, [Fe/H]=+2) at our request. Model 3 (line, [Fe/H]) was calculated by FK, using the Canuto-Mazzitelli formalism. Figure 1 (www.astro.lsa.umich.edu/usrs/cowley/models.gif), shows these 3 models in good agreement with one another, and clearly different from a standard solar-abundance Atlas9 model (dashed) with the same effective temperature. All three models are scaled to Te=6600K. The blanketed models have little or no convection, and show the lowered boundary temperature of classical picket-fence models. The true boundary temperature may be still lower than in these numerical models. Abundances from Pr I and Nd I are systematically higher than those from the corresponding second spectra, as are those from Pr III and Nd III. It was noted long ago by Przybylski and others that the Balmer profiles had cores indicative of temperatures of some 6000K; the wings could be fit with much higher temperatures--perhaps as high as 7500K. Molecular species have been sought but not identified. Calculations show CN and CH lines would be very weak, even if the temperature between log(tau5000)=-3.5 and -5.4 were allowed to drop to 3000K.

  7. ITER breeding blanket design

    SciTech Connect

    Gohar, Y.; Cardella, A.; Ioki, K.; Lousteau, D.; Mohri, K.; Raffray, R.; Zolti, E.

    1995-12-31

    A breeding blanket design has been developed for ITER to provide the necessary tritium fuel to achieve the technical objectives of the Enhanced Performance Phase. It uses a ceramic breeder and water coolant for compatibility with the ITER machine design of the Basic Performance Phase. Lithium zirconate and lithium oxide am the selected ceramic breeders based on the current data base. Enriched lithium and beryllium neutron multiplier are used for both breeders. Both forms of beryllium material, blocks and pebbles are used at different blanket locations based on thermo-mechanical considerations and beryllium thickness requirements. Type 316LN austenitic steel is used as structural material similar to the shielding blanket. Design issues and required R&D data are identified during the development of the design.

  8. Thermionic Fuel Element Verification Program - Overview

    NASA Astrophysics Data System (ADS)

    Bohl, Richard J.; Dahlberg, Richard C.; Dutt, Dale S.; Wood, John T.

    The Thermionic Fuel Element (TFE) Verification program was established in 1986 to resolve the technology concerns raised in Phase 1 of the SP-100 program, namely, the performance and lifetime of thermionic fuel elements in a fast spectrum reactor. The program builds directly on an extensive database developed in the 1960s and early 1970s in an AEC/NASA-sponsored program, when TFEs were developed and tested at design conditions for over 10,000 h. The current effort has reestablished that technology and is extending the lifetime up to 7 to 10 yr. A TFE lifetime of more than 2 yr has been demonstrated in the TRIGA reactor. Component lifetimes of more than 10 yr have been demonstrated in accelerated tests in the FFTF (Richland) and EBR-II (Idaho) test reactors. Program completion is scheduled for FY-95.

  9. Thermal insulation blanket material

    NASA Technical Reports Server (NTRS)

    Pusch, R. H.

    1982-01-01

    A study was conducted to provide a tailorable advanced blanket insulation based on a woven design having an integrally woven core structure. A highly pure quartz yarn was selected for weaving and the cells formed were filled with a microquartz felt insulation.

  10. Blanket integrated blocking diodes

    NASA Astrophysics Data System (ADS)

    Uebele, P.; Kasper, C.; Rasch, K.-D.

    1986-11-01

    Two types of large area protection diodes for integration in solar arrays were developed in planar technology. For application in a bus voltage concept of V sub bus = 80 V a p-doped blanket integrated blocking diode (p-IBD) was developed with V sub rev = 120 V, whereas for the high voltage concept of V sub bus = 160 V a n-IBD with V sub rev = 250 V was developed. Application as blanket integrated shunt diodes is recommended. The optimized rearside diffusion provides a low forward voltage drop in the temperature range of minus 100 to plus 150 C. As a consequence of planar technology metallized coverglasses have to be used to minimize the photocurrent.

  11. Fusion Blanket Development in FDF

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Smith, J. P.; Stambaugh, R. D.

    2008-11-01

    To satisfy the electricity and tritium self-sufficiency missions of a Fusion Development Facility (FDF), suitable blanket designs will need to be evaluated, selected and developed. To demonstrate closure of the fusion fuel cycle, 2-3 main tritium breeding blankets will be used to cover most of the available chamber surface area in order to reach the project goal of achieving a tritium breeding ratio, TBR > 1. To demonstrate the feasibility of electricity and tritium production for subsequent devices such as the fusion demonstration power reactor (DEMO), several advanced test blankets will need to be selected and tested on the FDF to demonstrate high coolant outlet temperature necessary for efficient electricity production. Since the design goals for the main and test blankets are different, the design criteria of these blankets will also be different. The considerations in performing the evaluation of blanket and structural material options in concert with the maintenance approach for the FDF will be reported in this paper.

  12. Tailorable Advanced Blanket Insulation (TABI)

    NASA Technical Reports Server (NTRS)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

    Single layer and multilayer insulating blankets for high-temperature service fabricated without sewing. TABI woven fabric made of aluminoborosilicate. Triangular-cross-section flutes of core filled with silica batting. Flexible blanket formed into curved shapes, providing high-temperature and high-heat-flux insulation.

  13. Transmutations of elements under irradiation and its impact on alloys composition

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1994-09-01

    This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V-4Cr-4Ti), a reduced activation martensitic steel (Fe-9Cr-2WVTa), a high conductivity copper alloy (Cu-Cr-Zr), and the SiC compound. The fission irradiation facilities considered include the EBR-II (Experimental Breeder Reactor) fast reactor, and two high flux mixed spectrum reactors, HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The transmutation and dpa rates that occur in these test reactors are compared with the calculated transmutation and dpa rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions [(n,x{alpha}) and (n,xp)] are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from the ones predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time. As examples, the transmutation rates of W, Ta, V, Cu, among others, differ considerably when the irradiation is performed under a mixed spectrum reactor`s and fusion first wall`s spectrum. Fast reactors (EBR-II) provide the only possibility for obtaining high damage rates without producing significant compositional effects in vanadium alloys, ferritic steels and copper alloys.

  14. Blanket comparison and selection study. Volume II

    SciTech Connect

    Not Available

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies. (MOW)

  15. Progress on DCLL Blanket Concept

    SciTech Connect

    Wong, Clement; Abdou, M.; Katoh, Yutai; Kurtz, Richard J.; Lumsdaine, A.; Marriott, Edward P.; Merrill, Brad; Morley, Neil; Pint, Bruce A.; Sawan, M.; Smolentsev, S.; Williams, Brian; Willms, Scott; Youssef, M.

    2013-09-01

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

  16. Adaptive robust control of the EBR-II reactor

    SciTech Connect

    Power, M.A.; Edwards, R.M.

    1996-05-01

    Simulation results are presented for an adaptive H{sub {infinity}} controller, a fixed H{sub {infinity}} controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H{sub {infinity}} controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H{sub {infinity}} and classical controllers. This makes for a superior and more robust controller.

  17. JAEA Fatigue Analysis of EBR-II Duplex Tubing

    SciTech Connect

    J. H. Jackson; D. L. Porter; W. R. Lloyd

    2009-07-01

    This work addresses questions brought up concerning the mechanisms associated with fatigue crack growth retardation and/or arrest within the nickel bond layer in duplex 2¼ Cr-1Mo steel superheater tubes. Previous work performed at the Idaho National Laboratory (INL) indicated that the nickel bond layer did not function as a crack arrestor during fatigue crack propagation with the exception of one, isolated case involving an exceptionally low fatigue load and a high temperature (400 0C) environment. Since it is atypical for a fatigue crack to propagate from a relatively soft material (the nickel bond layer) to a harder material (the 2¼ Cr-1Mo steel) there has been speculation that the nickel bond layer was hardened in service. Additionally, there are questions surrounding the nature of the fatigue crack propagation within the nickel bond layer; specifically with regard to the presence of voids seen on micrographs of the bond layer and oxidation within the steel along the edge of the nickel bond layer. There is uncertainty as to the effect of these voids and/or oxide barriers with respect to potential fatigue crack arrest.

  18. US solid breeder blanket design for ITER

    SciTech Connect

    Gohar, Y.; Attaya, H.; Billone, M.; Lin, C.; Johnson, C.; Majumdar, S.; Smith, D. ); Goranson, P.; Nelson, B.; Williamson, D.; Baker, C. ); Raffray, A.; Badawi, A.; Gorbis, Z.; Ying, A.; Abdou, M. ); Sviatoslavsky, I.; Blanchard, J.; Mogahed, E.; Sawan, M.; Kulcinski, G. )

    1990-09-01

    The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.

  19. Packed fluidized bed blanket for fusion reactor

    DOEpatents

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  20. Toughened Thermal Blanket for MMOD Protection

    NASA Technical Reports Server (NTRS)

    Christiansen, Eric L.; Lear, Dana M.

    2014-01-01

    Thermal blankets are used extensively on spacecraft to provide passive thermal control of spacecraft hardware from thermal extremes encountered in space. Toughened thermal blankets have been developed that greatly improve protection from hypervelocity micrometeoroid and orbital debris (MMOD) impacts. These blankets can be outfitted if so desired with a reliable means to determine the location, depth and extent of MMOD impact damage by incorporating an impact sensitive piezoelectric film. Improved MMOD protection of thermal blankets was obtained by adding selective materials at various locations within the thermal blanket. As given in Figure 1, three types of materials were added to the thermal blanket to enhance its MMOD performance: (1) disrupter layers, near the outside of the blanket to improve breakup of the projectile, (2) standoff layers, in the middle of the blanket to provide an area or gap that the broken-up projectile can expand, and (3) stopper layers, near the back of the blanket where the projectile debris is captured and stopped. The best suited materials for these different layers vary. Density and thickness is important for the disrupter layer (higher densities generally result in better projectile breakup), whereas a highstrength to weight ratio is useful for the stopper layer, to improve the slowing and capture of debris particles.

  1. Proposed fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor.

  2. ARIES-IV Nested Shell Blanket Design

    SciTech Connect

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design.

  3. Method of fabricating a multilayer insulation blanket

    DOEpatents

    Gonczy, John D.; Niemann, Ralph C.; Boroski, William N.

    1993-01-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  4. Multilayer insulation blanket, fabricating apparatus and method

    DOEpatents

    Gonczy, John D.; Niemann, Ralph C.; Boroski, William N.

    1992-01-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  5. Method of fabricating a multilayer insulation blanket

    DOEpatents

    Gonczy, J.D.; Niemann, R.C.; Boroski, W.N.

    1993-07-06

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel.

  6. Multilayer insulation blanket, fabricating apparatus and method

    DOEpatents

    Gonczy, J.D.; Niemann, R.C.; Boroski, W.N.

    1992-09-01

    An improved multilayer insulation blanket for insulating cryogenic structures operating at very low temperatures is disclosed. An apparatus and method for fabricating the improved blanket are also disclosed. In the improved blanket, each successive layer of insulating material is greater in length and width than the preceding layer so as to accommodate thermal contraction of the layers closest to the cryogenic structure. The fabricating apparatus has a rotatable cylindrical mandrel having an outer surface of fixed radius that is substantially arcuate, preferably convex, in cross-section. The method of fabricating the improved blanket comprises (a) winding a continuous sheet of thermally reflective material around the circumference of the mandrel to form multiple layers, (b) binding the layers along two lines substantially parallel to the edges of the circumference of the mandrel, (c) cutting the layers along a line parallel to the axle of the mandrel, and (d) removing the bound layers from the mandrel. 7 figs.

  7. Multifractal Framework Based on Blanket Method

    PubMed Central

    Paskaš, Milorad P.; Reljin, Irini S.; Reljin, Branimir D.

    2014-01-01

    This paper proposes two local multifractal measures motivated by blanket method for calculation of fractal dimension. They cover both fractal approaches familiar in image processing. The first two measures (proposed Methods 1 and 3) support model of image with embedded dimension three, while the other supports model of image embedded in space of dimension three (proposed Method 2). While the classical blanket method provides only one value for an image (fractal dimension) multifractal spectrum obtained by any of the proposed measures gives a whole range of dimensional values. This means that proposed multifractal blanket model generalizes classical (monofractal) blanket method and other versions of this monofractal approach implemented locally. Proposed measures are validated on Brodatz image database through texture classification. All proposed methods give similar classification results, while average computation time of Method 3 is substantially longer. PMID:24578664

  8. Axially staggered seed-blanket reactor fuel module construction

    DOEpatents

    Cowell, Gary K.; DiGuiseppe, Carl P.

    1985-01-01

    A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

  9. Current Trends of Blanket Research and Deveopment in Japan 3.Blanket Designs in Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Sagara, Akio; Enoeda, Mikio; Nishio, Satoshi; Kozaki, Yasuji

    The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

  10. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  11. Flute stabilization by a cold line-tied blanket

    SciTech Connect

    Segal, D.; Wickham, M.; Rynn, N.

    1982-09-01

    The curvature-driven flute instability in an axisymmetric mirror was stabilized by an annular line-tied plasma blanket. A significant temperature difference was maintained between core and blanket. Theoretical calculations support the experimental observations.

  12. Insulation Blankets for High-Temperature Use

    NASA Technical Reports Server (NTRS)

    Goldstein, H.; Leiser, D.; Sawko, P. M.; Larson, H. K.; Estrella, C.; Smith, M.; Pitoniak, F. J.

    1986-01-01

    Insulating blanket resists temperatures up to 1,500 degrees F (815 degrees C). Useful where high-temperature resistance, flexibility, and ease of installation are important - for example, insulation for odd-shaped furnaces and high-temperature ducts, curtains for furnace openings and fire control, and conveyor belts in hot processes. Blanket is quilted composite consisting of two face sheets: outer one of silica, inner one of silica or other glass cloth with center filling of pure silica glass felt sewn together with silica glass threads.

  13. Lightweight IMM PV Flexible Blanket Assembly

    NASA Technical Reports Server (NTRS)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

  14. 32 CFR 318.14 - Blanket routine uses.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 2 2010-07-01 2010-07-01 false Blanket routine uses. 318.14 Section 318.14 National Defense Department of Defense (Continued) OFFICE OF THE SECRETARY OF DEFENSE (CONTINUED) PRIVACY PROGRAM DEFENSE THREAT REDUCTION AGENCY PRIVACY PROGRAM § 318.14 Blanket routine uses. (a) Blanket...

  15. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  16. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  17. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    .... Where a manufacturer of tobacco products operates more than one factory in the same region he may, in... provisions of § 40.134, for any or all of the factories in the same region. The total amount of any blanket... factory covered by the bond. (72 Stat. 1421; 26 U.S.C. 5711)...

  18. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  19. 27 CFR 40.67 - Blanket bond.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    .... Where a manufacturer of tobacco products operates more than one factory he may, in lieu of filing... provisions of § 40.134, for any or all of the factories. The total amount of any blanket bond given under this section shall be available for the satisfaction of any liability incurred at any factory...

  20. Aerogel Blanket Insulation Materials for Cryogenic Applications

    NASA Technical Reports Server (NTRS)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

  1. Advanced Polymer For Multilayer Insulating Blankets

    NASA Technical Reports Server (NTRS)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  2. Fidget Blankets: A Sensory Stimulation Outreach Program.

    PubMed

    Kroustos, Kelly Reilly; Trautwein, Heidi; Kerns, Rachel; Sobota, Kristen Finley

    2016-01-01

    Behavioral and Psychological Symptoms of Dementia (BPSD) include behaviors such as aberrant motor behavior, agitation, anxiety, apathy, delusions, depression, disinhibition, elation, hallucinations, irritability, and sleep or appetite changes. A student-led project to provide sensory stimulation in the form of "fidget blankets" developed into a community outreach program. The goal was to decrease the use of antipsychotics used for BPSD. PMID:27250073

  3. Thermal insulation blanket material. Final Report

    SciTech Connect

    Pusch, R.H.

    1982-06-01

    A study was conducted to provide a tailorable advanced blanket insulation based on a woven design having an integrally woven core structure. A highly pure quartz yarn was selected for weaving and the cells formed were filled with a microquartz felt insulation.

  4. 18 CFR 157.203 - Blanket certification.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket certification. 157.203 Section 157.203 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY REGULATIONS UNDER NATURAL GAS ACT APPLICATIONS FOR CERTIFICATES OF PUBLIC CONVENIENCE AND NECESSITY AND FOR ORDERS...

  5. Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

    DOEpatents

    Cowell, G.K.; DiGuiseppe, C.P.

    1982-10-28

    A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

  6. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    SciTech Connect

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab.

  7. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  8. Novel method for sludge blanket measurements.

    PubMed

    Schewerda, J; Förster, G; Heinrichmeier, J

    2014-01-01

    The most widely used methods for sludge blanket measurements are based on acoustic or optic principles. In operation, both methods are expensive and often maintenance-intensive. Therefore a novel, reliable and simple method for sludge blanket measurement is proposed. It is based on the differential pressure measurement in the sludge zone compared with the differential pressure in the clear water zone, so that it is possible to measure the upper and the lower sludge level in a tank. Full-scale tests of this method were done in the secondary clarifier at the waste water treatment plant in Hecklingen, Germany. The result shows a good approximation of the manually measured sludge level. PMID:24569276

  9. Chicxulub Ejecta Blanket Deposits From Belize

    NASA Technical Reports Server (NTRS)

    Ocampo, A.

    1995-01-01

    The Chicxulub impact into a thick sequence of carbonates and sulfates released over a trillion tons of volatiles. The importance of the explosive release of such a large mass of volatiles has been greatly underestimated in studies of ejecta depositional processes. Proximal Chicxulub ejecta blanket deposits recent discovered on Albion Island in Belize provide a key to understanding the role of volatile-rich target material during large impact events.

  10. A light blanket for intraoperative photodynamic therapy

    NASA Astrophysics Data System (ADS)

    Hu, Yida; Wang, Ken; Zhu, Timothy C.

    2009-06-01

    A novel light source - light blanket composed of a series of parallel cylindrical diffusing fibers (CDF) is designed to substitute the hand-held point source in the PDT treatment of the malignant pleural or intraperitoneal diseases. It achieves more uniform light delivery and less operation time in operating room. The preliminary experiment was performed for a 9cmx9cm light blanket composed of 8 9-cm CDFs. The linear diffusers were placed in parallel fingerlike pockets. The blanket is filled with 0.2 % intralipid scattering medium to improve the uniformity of light distribution. 0.3-mm aluminum foil is used to shield and reflect the light transmission. The full width of the profile of light distribution at half maximum along the perpendicular direction is 7.9cm and 8.1cm with no intralipid and with intralipid. The peak value of the light fluence rate profiles per input power is 11.7mW/cm2/W and 8.6mW/cm2/W respectively. The distribution of light field is scanned using the isotropic detector and the motorized platform. The average fluence rate per input power is 8.6 mW/cm2/W and the standard deviation is 1.6 mW/cm2/W for the scan in air, 7.4 mW/cm2/W and 1.1 mW/cm2/W for the scan with the intralipid layer. The average fluence rate per input power and the standard deviation are 20.0 mW/cm2/W and 2.6 mW/cm2/W respectively in the tissue mimic phantom test. The light blanket design produces a reasonably uniform field for effective light coverage and is flexible to confirm to anatomic structures in intraoperative PDT. It also has great potential value for superficial PDT treatment in clinical application.

  11. Detection of Breeding Blankets Using Antineutrinos

    NASA Astrophysics Data System (ADS)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  12. High power density self-cooled lithium-vanadium blanket.

    SciTech Connect

    Gohar, Y.; Majumdar, S.; Smith, D.

    1999-07-01

    A self-cooled lithium-vanadium blanket concept capable of operating with 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple blanket configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in the blanket performance. Spectral shifter, moderator, and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.

  13. Thin Thermal-Insulation Blankets for Very High Temperatures

    NASA Technical Reports Server (NTRS)

    Choi, Michael K.

    2003-01-01

    Thermal-insulation blankets of a proposed type would be exceptionally thin and would endure temperatures up to 2,100 C. These blankets were originally intended to protect components of the NASA Solar Probe spacecraft against radiant heating at its planned closest approach to the Sun (a distance of 4 solar radii). These blankets could also be used on Earth to provide thermal protection in special applications (especially in vacuum chambers) for which conventional thermal-insulation blankets would be too thick or would not perform adequately.

  14. Neutronics analysis of deuterium-tritium-driven experimental hybrid blankets

    SciTech Connect

    Sahin, S.; Kumar, A.

    1984-07-01

    At the Swiss Federal Institute of Technology, an experimental fusion and fusion-fission (hybrid) reactor facility is near completion. Experiments are scheduled to begin in February 1984. The experimental cavity leads one to plan experiments mostly with blankets in plane geometry. Five different hybrid blanket modules in plane geometry are analyzed with two different left boundary conditions representing varying experimental situations. Numbers I and II represent energy and fissile fuel producing blankets, whereas number III is mainly a fissile fuel producing blanket. Numbers IV and V are actinide burning blankets. It is shown that the overall neutronic performance, such as k /sub eff/ , energy multiplication factor M, fusile and fissile breeding, of a hybrid blanket with transplutonium actinide fuel is already better than that of a UO/sub 2/ or ThO/sub 2/ hybrid blanket. Furthermore, the transplutonium actinide waste is partly converted into precious nuclear fuel of a new type, such as /sup 242m/ Am and /sup 245/Cm. An experimental blanket with a vacuum left boundary has a harder neutron spectrum, and also excessive neutron leakage from the front surface and the lateral surfaces, as compared to that in the blanket in confinement geometry. It leads to the poorer neutronic performance of the former.

  15. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  16. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    SciTech Connect

    Finn, P.A.

    1985-01-01

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared.

  17. High temperature - low mass solar blanket

    NASA Technical Reports Server (NTRS)

    Mesch, H. G.

    1979-01-01

    Interconnect materials and designs for use with ultrathin silicon solar cells are discussed, as well as the results of an investigation of the applicability of parallel-gap resistance welding for interconnecting these cells. Data relating contact pull strength and cell electrical degradation to variations in welding parameters such as time, voltage and pressure are presented. Methods for bonding ultrathin cells to flexible substances and for bonding thin (75 micrometers) covers to these cells are described. Also, factors influencing fabrication yield and approaches for increasing yield are discussed. The results of vacuum thermal cycling and thermal soak tests on prototype ultrathin cell test coupons and one solar module blanket are presented.

  18. Specific welds for test blanket modules

    NASA Astrophysics Data System (ADS)

    Rieth, Michael; Rey, Jörg

    2009-04-01

    Fabrication and assembling test blanket modules needs a variety of different welding techniques. Therefore, an evaluation of plate joining for breeder units by tungsten-inert-gas, laser, and electron beam welding was performed by qualification of relevant mechanical properties like hardness, charpy, and creep strength. The focus was laid on the study of post-weld heat treatments at lowest possible temperatures and for maximum recovery of the joints. The most important result is that thin EUROFER plates may be welded by EB or laser techniques without the necessity of post-welding heat treatments that include an austenitization step.

  19. 75 FR 51482 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-20

    ... publishing the notice in the Federal Register of March 11, 2010 (75 FR 11557). The hearing was held in... COMMISSION Woven Electric Blankets From China Determination On the basis of the record \\1\\ developed in the... United States is materially injured by reason of imports from China of woven electric blankets,...

  20. Security Blankets and Children's Security of Attachment to Their Mothers.

    ERIC Educational Resources Information Center

    Donate-Bartfield, Evelyn L.; Passman, Richard H.

    This study investigated the relations between toddlers' degree of attachment to their mothers and their development of an attachment to a security blanket. Seventy-four 18-month-olds were separated from their mothers three times; the third time the toddlers were left for 5 minutes in an unfamiliar playroom with their blanket and with a stranger.…

  1. Overview of the TFTB lithium blanket module program

    SciTech Connect

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an approx. 80-cm/sup 3/ module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program.

  2. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 18 Conservation of Power and Water Resources 1 2011-04-01 2011-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  3. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 18 Conservation of Power and Water Resources 1 2014-04-01 2014-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  4. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  5. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 18 Conservation of Power and Water Resources 1 2012-04-01 2012-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  6. 18 CFR 284.402 - Blanket marketing certificates.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 18 Conservation of Power and Water Resources 1 2013-04-01 2013-04-01 false Blanket marketing certificates. 284.402 Section 284.402 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... RELATED AUTHORITIES Certain Sales for Resale by Non-interstate Pipelines § 284.402 Blanket...

  7. Diffusive heat blanketing envelopes of neutron stars

    NASA Astrophysics Data System (ADS)

    Beznogov, M. V.; Potekhin, A. Y.; Yakovlev, D. G.

    2016-06-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H-He, He-C, C-Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density ρb ˜ 108 - 1010 g cm-3) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses ΔM of lighter ions in the envelope. Our principal result is that the Ts-Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on ΔM. The obtained relations are approximated by analytic expressions which are convenient for modelling the evolution of neutron stars.

  8. Development of blanket box structure fabrication technology

    SciTech Connect

    Mohri, K.; Sata, S.; Kawaguchi, I.

    1994-12-31

    Fabrication studies have been performed for first wall and blanket box structure in the Fusion Experimental Reactor designed in Japan. The first wall must have internal cooling channels to remove volumetric heat loading by neutron wall load and surface heat loading from the plasma. The blanket which is higher than 10 m and 1 m wide withstands enormous electromagnetic load (about 10 MN/m). And a fabrication accuracy is required in the order of 10 mm from the machine configuration and remote assembling standpoints. To make cooling channels inside the first wall and to reduce the deformation during fabrication, the authors adopted advance techniques Hot Isostatic Pressing method (HIP) and Electron Beam Welding (EBW) respectively. Evaluation studies for the bondability of the HIP bonding joint have been performed. To evaluate the bondability, the mechanical properties such as tensile strength, impact value, low cycle fatigue strength and creep strength of the bonded part were investigated using HIP bonded test specimens. And the detectability of ultrasonic detection tests were also studied on them.

  9. The TFTR lithium blanket module program

    SciTech Connect

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-02-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li/sub 2/O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li/sub 2/O pellets with satisfactory reproducibility were developed using purified Li/sub 2/O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g).

  10. MIT LMFBR blanket research project. Final summary report

    SciTech Connect

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record.

  11. Neutron dosimetry for the Lithium-Blanket-Module program

    SciTech Connect

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.; Schultz, E.K.

    1982-01-01

    The Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the Tokamak fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to prototypical fusion reactor blanket conditions, and (2) to obtain tritium breeding and power production performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory.

  12. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    SciTech Connect

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory (INEL).

  13. Neutronic design for the TFTR lithium blanket module

    SciTech Connect

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design.

  14. US technical report for the ITER blanket/shield

    NASA Astrophysics Data System (ADS)

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li2O) and lithium zirconate (Li2ZrO3) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  15. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    SciTech Connect

    Not Available

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  16. Analyses of Hubble Space Telescope Aluminized-Teflon Multilayer Insulation Blankets Retrieved After 19 Years of Space Exposure

    NASA Technical Reports Server (NTRS)

    de Groh, Kim K.; Perry, Bruce A.; Mohammed, Jelila S.; Banks, Bruce

    2015-01-01

    Since its launch in April 1990, the Hubble Space Telescope (HST) has made many important observations from its vantage point in low Earth orbit (LEO). However, as seen during five servicing missions, the outer layer of multilayer insulation (MLI) has become increasingly embrittled and has cracked in many areas. In May 2009, during the 5th servicing mission (called SM4), two MLI blankets were replaced with new insulation and the space-exposed MLI blankets were retrieved for degradation analyses by teams at NASA Glenn Research Center (GRC) and NASA Goddard Space Flight Center (GSFC). The retrieved MLI blankets were from Equipment Bay 8, which received direct sunlight, and Equipment Bay 5, which received grazing sunlight. Each blanket was divided into several regions based on environmental exposure and/or physical appearance. The aluminized-Teflon (DuPont, Wilmington, DE) fluorinated ethylene propylene (Al-FEP) outer layers of the retrieved MLI blankets have been analyzed for changes in optical, physical, and mechanical properties, along with chemical and morphological changes. Pristine and as-retrieved samples (materials) were heat treated to help understand degradation mechanisms. When compared to pristine material, the analyses have shown how the Al-FEP was severely affected by the space environment. Most notably, the Al-FEP was highly embrittled, fracturing like glass at strains of 1 to 8 percent. Across all measured properties, more significant degradation was observed for Bay 8 material as compared to Bay 5 material. This paper reviews the tensile and bend-test properties, density, thickness, solar absorptance, thermal emittance, x-ray photoelectron spectroscopy (XPS) and energy dispersive spectroscopy (EDS) elemental composition measurements, surface and crack morphologies, and atomic oxygen erosion yields of the Al-FEP outer layer of the retrieved HST blankets after 19 years of space exposure.

  17. Spacecraft thermal blanket cleaning: Vacuum bake of gaseous flow purging

    NASA Technical Reports Server (NTRS)

    Scialdone, John J.

    1990-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours, In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  18. Spacecraft thermal blanket cleaning - Vacuum baking or gaseous flow purging

    NASA Technical Reports Server (NTRS)

    Scialdone, John J.

    1992-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours. In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  19. Design analyses of self-cooled liquid metal blankets

    SciTech Connect

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations.

  20. Overview of EU activities on DEMO liquid metal breeder blanket

    SciTech Connect

    Giancarli, L.; Proust, E.

    1994-12-31

    The European test-blanket development programme, started in 1988, is aiming at the selection by 1995 of two DEMO-relevant blanket lines to be tested in ITER. At present, four lines of blanket are under development, two of them using solid and the other two liquid breeder materials. As far as liquid breeders are concerned, two lines of blankets have been selected within the European Union, the water-cooled lithium-lead (the eutectic Pb-17Li) blankets and the dual-coolant Pb-17Li blankets. Designs have been developed considering an agreed set of DEMO specifications, such as, for instance, a fusion power of 2,200 MW, a neutron wall-loading of 2MW/m{sup 2}, a life-time of 20,000 hours, and the use of martensitic steel as a structural material. Moreover, an experimental program has been set up in order to address the main critical issues for each line. The present paper gives an overview of both design and experimental activities within the European Union concerning these two lines of liquid breeder blankets.

  1. Parametric Weight Comparison of Advanced Metallic, Ceramic Tile, and Ceramic Blanket Thermal Protection Systems

    NASA Technical Reports Server (NTRS)

    Myers, David E.; Martin, Carl J.; Blosser, Max L.

    2000-01-01

    A parametric weight assessment of advanced metallic panel, ceramic blanket, and ceramic tile thermal protection systems (TPS) was conducted using an implicit, one-dimensional (I-D) finite element sizing code. This sizing code contained models to account for coatings fasteners, adhesives, and strain isolation pads. Atmospheric entry heating profiles for two vehicles, the Access to Space (ATS) vehicle and a proposed Reusable Launch Vehicle (RLV), were used to ensure that the trends were not unique to a certain trajectory. Ten TPS concepts were compared for a range of applied heat loads and substructural heat capacities to identify general trends. This study found the blanket TPS concepts have the lightest weights over the majority of their applicable ranges, and current technology ceramic tiles and metallic TPS concepts have similar weights. A proposed, state-of-the-art metallic system which uses a higher temperature alloy and efficient multilayer insulation was predicted to be significantly lighter than the ceramic tile stems and approaches blanket TPS weights for higher integrated heat loads.

  2. The excitation of plasma lines in blanketing sporadic E

    NASA Technical Reports Server (NTRS)

    Gordon, W. E.; Carlson, H. C.

    1976-01-01

    Enhanced plasma lines in blanketing sporadic E have been excited by a powerful HF radio wave illuminating the E region over the Arecibo Observatory. The plasma lines are observed by the incoherent scatter radar at the observatory. They originate in the sporadic E layer when the blanketing frequency exceeds the exciting frequency, a result which confirms that the plasma is overdense for the exciting frequency. Around the time when the blanketing frequency falls through the exciting frequency, large fluctuations in the plasma line intensities are observed, and thus the possibility of overdense patches drifting through the sampled volume is suggested.

  3. Development of fusion blanket technology for the DEMO reactor.

    PubMed

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

  4. Hubble Space Telescope Thermal Blanket Repair Design and Implementation

    NASA Technical Reports Server (NTRS)

    Ousley, Wes; Skladany, Joseph; Dell, Lawrence

    2000-01-01

    Substantial damage to the outer layer of Hubble Space Telescope (HST) thermal blankets was observed during the February 1997 servicing mission. After six years in LEO, many areas of the aluminized Teflon(R) outer blanket layer had significant cracks, and some material was peeled away to expose inner layers to solar flux. After the mission, the failure mechanism was determined, and repair materials and priorities were selected for follow-on missions. This paper focuses on the thermal, mechanical, and EVA design requirements for the blanket repair, the creative solutions developed for these unique problems, hardware development, and testing.

  5. Disinfection of woollen blankets in steam at subatmospheric pressure

    PubMed Central

    Alder, V. G.; Gillespie, W. A.

    1961-01-01

    Blankets may be disinfected in steam at subatmospheric pressures by temperatures below boiling point inside a suitably adapted autoclave chamber. The chamber and its contents are thoroughly evacuated of air so as to allow rapid heat penetration, and steam is admitted to a pressure of 10 in. Hg below atmospheric pressure, which corresponds to a temperature of 89°C. Woollen blankets treated 50 times by this process were undamaged. Vegetative organisms were destroyed but not spores. The method is suitable for large-scale disinfection of blankets and for disinfecting various other articles which would be damaged at higher temperatures. PMID:13860203

  6. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    SciTech Connect

    B.R. Westphal; K.C. Marsden; W.M. McCartin; S.M. Frank; D.D. Keiser, Jr.; T.S. Yoo; D. Vaden; D.G. Cummings; K.J. Bateman; J. J. Giglio; T. P. O'Holleran; P. A. Hahn; M. N. Patterson

    2013-03-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  7. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    NASA Astrophysics Data System (ADS)

    Westphal, Brian R.; Frank, S. M.; McCartin, W. M.; Cummings, D. G.; Giglio, J. J.; O'Holleran, T. P.; Hahn, P. A.; Yoo, T. S.; Marsden, K. C.; Bateman, K. J.; Patterson, M. N.

    2015-01-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 °C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  8. Cassini/Titan-4 Acoustic Blanket Development and Testing

    NASA Technical Reports Server (NTRS)

    Hughes, William O.; McNelis, Anne M.

    1996-01-01

    NASA Lewis Research Center recently led a multi-organizational effort to develop and test verify new acoustic blankets. These blankets support NASA's goal in reducing the Titan-4 payload fairing internal acoustic environment to allowable levels for the Cassini spacecraft. To accomplish this goal a two phase acoustic test program was utilized. Phase One consisted of testing numerous blanket designs in a flat panel configuration. Phase Two consisted of testing the most promising designs out of Phase One in a full scale cylindrical payload fairing. This paper will summarize this highly successful test program by providing the rationale and results for each test phase, the impacts of this testing on the Cassini mission, as well as providing some general information on blanket designs.

  9. Fusion blanket for high-efficiency power cycles

    SciTech Connect

    Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500/sup 0/C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO/sub 2/ interior (cooled by Ar) utilizing Li/sub 2/O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230/sup 0/C leading to an overall efficiency estimate of 55 to 60% for this reference case.

  10. Performance of uncoated AFRSI blankets during multiple Space Shuttle flights

    NASA Astrophysics Data System (ADS)

    Sawko, Paul M.; Goldstein, Howard E.

    1992-04-01

    Uncoated Advanced Flexible Reusable Surface Insulation (AFRSI) blankets were successfully flown on seven consecutive flights of the Space Shuttle Orbiter OV-099 (Challenger). In six of the eight locations monitored (forward windshield, forward canopy, mid-fuselage, upper wing, rudder/speed brake, and vertical tail) the AFRSI blankets performed well during the ascent and reentry exposure to the thermal and aeroacoustic environments. Several of the uncoated AFRSI blankets that sustained minor damage, such as fraying or broken threads, could be repaired by sewing or by patching with a surface coating called C-9. The chief reasons for replacing or completely coating a blanket were fabric embrittlement and fabric abrasion caused by wind erosion. This occurred in the orbiter maneuvering system (OMS) pod sidewall and the forward mid-fuselage locations.

  11. Thin Thermal-Insulation Blankets for Very High Temperatures

    NASA Technical Reports Server (NTRS)

    Choi, Michael K.

    2003-01-01

    Thermal-insulation blankets of a proposed type would be exceptionally thin and would endure temperatures up to 2,100 C. These blankets were originally intended to protect components of the NASA Solar Probe spacecraft against radiant heating at its planned closest approach to the Sun (a distance of 4 solar radii). These blankets could also be used on Earth to provide thermal protection in special applications (especially in vacuum chambers) for which conventional thermal-insulation blankets would be too thick or would not perform adequately. A blanket according to the proposal (see figure) would be made of molybdenum, titanium nitride, and carbon- carbon composite mesh, which melt at temperatures of 2,610, 2,930, and 2,130 C, respectively. The emittance of molybdenum is 0.24, while that of titanium nitride is 0.03. Carbon-carbon composite mesh is a thermal insulator. Typically, the blanket would include 0.25-mil (.0.00635-mm)-thick hot-side and cold-side cover layers of molybdenum. Titanium nitride would be vapor-deposited on both surfaces of each cover layer. Between the cover layers there would be 10 inner layers of 0.15-mil (.0.0038-mm)-thick molybdenum with vapor-deposited titanium nitride on both sides of each layer. The thickness of each titanium nitride coat would be about 1,000 A. The cover and inner layers would be interspersed with 0.25-mil (0.00635-mm)-thick layers of carbon-carbon composite mesh. The blanket would have total thickness of 4.75 mils (approximately equal to 0.121 mm) and an areal mass density of 0.7 kilograms per square meter. One could, of course, increase the thermal- insulation capability of the blanket by increasing number of inner layers (thereby unavoidably increasing the total thickness and mass density).

  12. Flexible, Thin-Film Solar-Cell Blanket

    NASA Technical Reports Server (NTRS)

    Stella, Paul M.

    1992-01-01

    Much of available area used to absorb solar energy. Proposed blanket of solar photovoltaic cells mounted on exterior surface of equipment it powers. Readily conforms to irregular shapes. Does not require separate supporting structure and saves space. Not added on to equipment but constitutes an integral part of it. Interconnection wiring deposited on sheet photolithographically or by other suitable masking/fabrication methods. Complete blanket, including cells and interconnections, fabricated as rigid unit directly on, and supported by, nonplanar surface to be covered.

  13. End-of-life destructive examinations of Zircaloy maximum depletion blanket fuel plates from the Shippingport PWR Core 2

    SciTech Connect

    Clayton, J.C.; Kammenzind, B.F.; Senio, P.; Sherman, J.

    1993-10-01

    Destructive examinations were performed on four Shippingport PWR Core 2 maximum fluence and depletion blanket plates for surface integrity, corrosion oxide thickness, and hydrogen absorption of the Zircaloy-4 cladding. The Shippingport PWR Core 2 operated for 23,360 effective full power hours (EFPH) (62,235 hot hours) at an average coolant temperature of 536{degrees}F (280{degrees}C) and a peak neutron flux of 0.6{times}10{sup 14}n/cm{sup 2}/s. The end-of-life examination program included measurements on three PWR-2 beta-quenched blanket fuel plates and one alpha-annealed blanket end plate. The examinations consisted of optical and scanning electron microscopy (SEM) inspections, direct metallographic oxide thickness measurements, and hydrogen extraction analyses on a joined element pair from the peak fluence (132{times}10{sup 20} n/cm{sup 2}), maximum depletion (13.5{times}10{sup 20} fissions/cc)PWR-2 blanket cluster.

  14. U.S. Plans and Strategy for ITER Blanket Testing

    SciTech Connect

    Abdou, M.; Sze, D.; Wong, C.; Sawan, M.; Ying, A.; Morley, N.B.; Malang, S

    2005-04-15

    Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R and D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R and D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder 'temperature window' and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanket concept have been selected for the US ITER-TBM.

  15. Blanket of Snow Covers Salt Lake City

    NASA Technical Reports Server (NTRS)

    2002-01-01

    On December 23, 2001, less than two months before the start of the 2002 Winter Olympics, snow blankets Salt Lake City and the surrounding area. The Great Salt Lake, on the left hand side of the image above, often contributes to the region's snowfall through the 'lake-effect.' As cold air passes over a large body of water it both warms and absorbs moisture. The warm air then rises (like a hot air balloon) and cools again. As it cools, the water vapor condenses out, resulting in snowfall. Just to the east (right) of the Great Salt Lake the mountains of the Wasatch Range lift air from the lake even higher, enhancing the lake-effect, resulting in an average snowfall of 64 inches a year in Salt Lake City and 140 inches in Park City, which is located at the foot of the Wasatch Front. For more information about the lake-effect, read Lake-Effect Snowfalls. Image courtesy Jacques Descloitres, MODIS Land Rapid Response Team at NASA GSFC

  16. Flow characteristics of the Cascade granular blanket

    SciTech Connect

    Pitts, J.H.; Walton, O.R.

    1985-07-01

    Analysis of a single granule on a rotating cone shows that for the 35/sup 0/ half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer.

  17. Flow characteristics of the Cascade granular blanket

    SciTech Connect

    Pitts, J.H.; Walton, O.R.

    1985-04-15

    Analysis of a single granule on a rotating cone shows that for the 35/sup 0/ half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer.

  18. APT Blanket System Loss-of-Helium-Gas Accident Based on Initial Conceptual Design - Helium Supply Rupture into Blanket Module

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    The model results are used to determine if beam power shutdown is necessary (or not) as a result of the LOHGA accident to maintain the blanket system well below any of the thermal-hydraulic constraints imposed on the design. The results also provide boundary conditions to the detailed bin model to study the detailed temperature response of the hot blanket module structure. The results for these two cases are documented in the report.

  19. High Power Density Blanket Design Study for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Huang, J. H.; Zhu, Y. K.; Deng, P. Zh.

    2003-06-01

    A conceptual design study of a high power density blanket has been carried out. The Fusion Experimental Breeder, FEB, is adopted as the reference reactor. The neutron wall loading is 0.5 MW/m2. The blanket is cooled by 10 MPa helium in tube. The concept of LiPb eutectic/transuranium oxide suspension is adopted. The neutronics design is performed to provide the design basis, and it gives an energy multiplication of 37 and a flattened power density distribution with a peak value of 70 W/m3. Multiple cooling panels are introduced to reduce the peak temperature of the blanket. In spite of up to 15 cooling panels, the blanket module is calculated using the ANSYS code and analytically as well. The results are consistent with each other and can meet the thermal criteria. However, structural calculation results from ANSYS did not satisfy the criterion: The blanket structure design is then improved by using curved cooling panels to model the structure in detail. Temperature distribution is obtained using the Pro/Mechanica code. Detailed structural analyses are also done by this code. Some satisfactory results are obtained.

  20. Overview of design activities for Li/V blankets

    SciTech Connect

    Sze, D.K.; Mattas, R.F.

    1997-12-31

    Recent fusion power plant design studies in the US have been conducted within the ARIES project. The most recent design of Li/V blankets was conducted as part of the ARIES-RS design. The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design.

  1. Progress in developing high performance solar blankets and arrays

    NASA Technical Reports Server (NTRS)

    Scott-Monck, J.

    1982-01-01

    The development of high efficiency, ultrathin silicon solar cells offers both opportunity and challenge. It is possible to consider 400 W/kg blanket designs by using this cell in conjuction with flexible substrates, ultrathin covers and welded interconnects. By designing array structure which is mechanically and dynamically compatible with very low mass blankets, solar arrays with a specific power approaching 200 W/kg are achievable. Further improvements in blanket performance (higher power and lower mass per unit area), which could come from the implementation of higher efficiency cells operating at lower temperatures (silicon or GaAs), and the use of encapsulants, would result in the development of 300 W/kg solar arrays.

  2. The evolution of US helium-cooled blankets

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Cheng, E. T.; Schultz, K. R.

    1991-08-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America. These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket confirmation for helium-cooled fusion power and experimental reactors.

  3. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    SciTech Connect

    Colon-Mercado, H.; Babineau, D.; Elvington, M.; Garcia-Diaz, B.; Teprovich, J.; Vaquer, A.

    2015-10-13

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed.  The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. 

  4. Overview of the TFTR Lithium Blanket Module Program

    SciTech Connect

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an {approximately}80-cm{sup 3} module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program. Construction of the LBM provided unique development and manufacturing experience with the mass production of reactor-representative lithium oxide pellets and fuel rods. Neutron activation and tritium assay data from present irradiation experiments with a point-neutron source and future experiments with the TFTR geometrically extended neutron source will reveal the ability to neutronics codes and models to characterize individual blanket module performance in a fusion device assembly.

  5. Experimental impacts into Teflon targets and LDEF thermal blankets

    NASA Technical Reports Server (NTRS)

    Hoerz, F.; Cintala, M. J.; Zolensky, M. E.; Bernhard, R. P.; See, T. H.

    1994-01-01

    The Long Duration Exposure Facility (LDEF) exposed approximately 20 sq m of identical thermal protective blankets, predominantly on the Ultra-Heavy Cosmic Ray Experiment (UHCRE). Approximately 700 penetration holes greater than 300 micron in diameter were individually documented, while thousands of smaller penetrations and craters occurred in these blankets. As a result of their 5.7 year exposure and because they pointed into a variety of different directions relative to the orbital motion of the nonspinning LDEF platform, these blankets can reveal important dynamic aspects of the hypervelocity particle environment in near-earth orbit. The blankets were composed of an outer teflon layer (approximately 125 micron thick), followed by a vapor-deposited rear mirror of silver (less than 1000 A thick) that was backed with an organic binder and a thermal protective paint (approximately 50 to 75 micron thick), resulting in a cumulative thickness (T) of approximately 175 to 200 microns for the entire blanket. Many penetrations resulted in highly variable delaminations of the teflon/metal or metal/organic binder interfaces that manifest themselves as 'dark' halos or rings, because of subsequent oxidation of the exposed silver mirror. The variety of these dark albedo features is bewildering, ranging from totally absent, to broad halos, to sharp single or multiple rings. Over the past year experiments were conducted over a wide range of velocities (i.e., 1 to 7 km/s) to address velocity dependent aspects of cratering and penetrations of teflon targets. In addition, experiments were performed with real LDEF thermal blankets to duplicate the LDEF delaminations and to investigate a possible relationship of initial impact conditions on the wide variety of dark halo and ring features.

  6. 75 FR 50991 - Antidumping Duty Order: Certain Woven Electric Blankets From the People's Republic of China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-18

    ...Based on affirmative final determinations by the Department of Commerce (the ``Department'') and the International Trade Commission (``ITC''), the Department is issuing an antidumping duty order on certain woven electric blankets (``woven electric blankets'') from the People's Republic of China...

  7. Study of Automated Module Fabrication for Lightweight Solar Blanket Utilization

    NASA Technical Reports Server (NTRS)

    Gibson, C. E.

    1979-01-01

    Cost-effective automated techniques for accomplishing the titled purpose; based on existing in-house capability are described. As a measure of the considered automation, the production of a 50 kilowatt solar array blanket, exclusive of support and deployment structure, within an eight-month fabrication period was used. Solar cells considered for this blanket were 2 x 4 x .02 cm wrap-around cells, 2 x 2 x .005 cm and 3 x 3 x .005 cm standard bar contact thin cells, all welded contacts. Existing fabrication processes are described, the rationale for each process is discussed, and the capability for further automation is discussed.

  8. Overview of the TFTR Lithium Blanket Module program

    SciTech Connect

    Jassby, D.L.

    1986-11-01

    The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests.

  9. Neutron Dosimetry Tokamak Fusion Test Reactor Lithium Blanket Module

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderl, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment.

  10. Blanket illumination vs scanned-mosaicking imaging schemes for wide-area photoacoustic tomography

    NASA Astrophysics Data System (ADS)

    Barber, Quinn; Harrison, Tyler; Zemp, Roger J.

    2015-03-01

    We compare scanned-mosaicking and blanket illumination schemes for wide-field photoacoustic tomography with potential applications to breast imaging. For each illumination, a locally high-SNR image patch is reconstructed then mosaicked with image patches from other illuminations. Because the beam is not diffused over the entire area, the fluence of the beam can be maximized, therefore maximizing the signal generated. Moreover, the imaging can potentially still be done fast enough within a breath-hold. A Monte Carlo simulation as a function of beam-spot size and depth is performed to quantify this signal gain. We experimentally test both schemes using a 256-element Imasonic ring array on a tissue-mimicking phantom. We were able to verify the simulated signal gain of 2.9x under 0.5 cm of tissue with the experimental data, and measured the signal gain decrease expected when imaging deeper into the tissue. We also measured the effectiveness of averaging the diffused beam versus the scanned-mosaicking approach, and observed that for the same scan times and limited laser power output, scanned-mosaicking was able to produce a higher SNR than the blanket illumination approach. We have shown that this technique will allow wide-area PAT to utilize the maximum SNR available from any system while minimizing the number of acquisitions to reach this SNR.

  11. 75 FR 11557 - Woven Electric Blankets From China

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-11

    ... permitted by section 201.8 of the Commission's rules, as amended, 67 FR 68036 (November 8, 2002). Even where... specified in II (C) of the Commission's Handbook on Electronic Filing Procedures, 67 FR 68168, 68173... COMMISSION Woven Electric Blankets From China AGENCY: United States International Trade Commission....

  12. Nitrogen sparging and blanketing of water storage tanks

    SciTech Connect

    Jonas, O.

    2000-04-01

    In many industrial processes, including most utility and industrial steam systems, good deaerated makeup and condensate water is stored in open-to-air storage tanks where it is contaminated by oxygen, carbon dioxide (CO{sub 2}), and dirt before it is used. This contamination can be prevented by nitrogen sparging and blanketing of storage tanks.

  13. Unified first wall-blanket structure for plasma device applications

    DOEpatents

    Gruen, Dieter M.

    1987-01-01

    A plasma device for use in controlling nuclear reactions within the plasma including a first wall and blanket formed in a one-piece structure composed of a solid solution containing copper and lithium and melting above about 500.degree. C.

  14. Fusion-reactor blanket-material safety-compatibility studies

    SciTech Connect

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/ and LiTiO/sub 3/) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li/sub 17/Pb/sub 83/ alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li/sub 17/Pb/sub 83/ alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns.

  15. Unified first wall - blanket structure for plasma device applications

    DOEpatents

    Gruen, D.M.

    A plasma device is described for use in controlling nuclear reactions within the plasma including a first wall and blanket formed in a one-piece structure composed of a solid solution containing copper and lithium and melting above about 500/sup 0/C.

  16. First-wall/blanket materials selection for STARFIRE tokamak reactor

    SciTech Connect

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed.

  17. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... United States, the authorization is conditioned on the holding company, consistent with 18 CFR 385.2005(b... 18 Conservation of Power and Water Resources 1 2013-04-01 2013-04-01 false Applicability, definitions, and blanket authorizations. 33.1 Section 33.1 Conservation of Power and Water Resources...

  18. GREEN CANE TRASH BLANKETS: INFLUENCE ON RATOON CROPS IN LOUISIANA

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Approximately 75% of Louisiana's 2000 sugarcane crop was harvested with a chopper harvester. A significant portion of the chopper-harvested sugarcane was harvested green, especially early in the season. Information on the impact of the post-harvest, green-cane residue blankets on subsequent ratoo...

  19. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2012-04-01 2011-04-01 true Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  20. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2013-04-01 2013-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  1. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2011-04-01 2011-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND TUBES, AND PROCESSED TOBACCO Bonds...

  2. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2010-04-01 2010-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  3. 27 CFR 40.134 - Amount of blanket bond.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 2 2014-04-01 2014-04-01 false Amount of blanket bond. 40.134 Section 40.134 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY (CONTINUED) TOBACCO MANUFACTURE OF TOBACCO PRODUCTS, CIGARETTE PAPERS AND...

  4. Technical issues for beryllium use in fusion blanket applications

    SciTech Connect

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented.

  5. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... United States, the authorization is conditioned on the holding company, consistent with 18 CFR 385.2005(b... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Applicability, definitions, and blanket authorizations. 33.1 Section 33.1 Conservation of Power and Water Resources...

  6. Neutronics and activation analysis of lithium-based ternary alloys in IFE blankets

    DOE PAGESBeta

    Jolodosky, Alejandra; Kramer, Kevin; Meier, Wayne; DeMuth, James; Reyes, Susana; Fratoni, Massimiliano

    2016-04-09

    Here we report that an attractive feature of using liquid lithium as the breeder and coolant in fusion blankets is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. The Lawrence Livermore National Laboratory is carrying an effort to develop a lithium-based alloy that maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) and at the same time reduces overall flammability concerns. This study evaluates the neutronics performance of lithium-based alloys inmore » the blanket of an inertial fusion energy chamber in order to inform such development. 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and the fusion energy multiplication factor (EMF). It was found that elements that exhibit low absorption cross sections and higher q-values such as lead, tin, and strontium, perform well with those that have high neutron multiplication such as lead and bismuth. These elements meet TBR constrains ranging from 1.02 to 1.1. However, most alloys do not reach EMFs greater than 1.15. Additionally, it was found that enriching lithium significantly increases the TBR and decreases the minimum lithium concentration by more than 60%. The amount of enrichment depends on how much total lithium is in the alloy to begin with. Alloys that performed well in the TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). Some of the activation results for alloys with tin, zinc, and gallium were in

  7. Effective Thermal Property Estimation of Unitary Pebble Beds Based on a CFD-DEM Coupled Method for a Fusion Blanket

    NASA Astrophysics Data System (ADS)

    Chen, Lei; Chen, Youhua; Huang, Kai; Liu, Songlin

    2015-12-01

    Lithium ceramic pebble beds have been considered in the solid blanket design for fusion reactors. To characterize the fusion solid blanket thermal performance, studies of the effective thermal properties, i.e. the effective thermal conductivity and heat transfer coefficient, of the pebble beds are necessary. In this paper, a 3D computational fluid dynamics discrete element method (CFD-DEM) coupled numerical model was proposed to simulate heat transfer and thereby estimate the effective thermal properties. The DEM was applied to produce a geometric topology of a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. Based on this geometric topology, a CFD model was built to analyze the temperature distribution and obtain the effective thermal properties. The current numerical model was shown to be in good agreement with the existing experimental data for effective thermal conductivity available in the literature. supported by National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2015GB108002, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  8. Free-vibration characteristics and correlation of a space station split-blanket solar array

    NASA Technical Reports Server (NTRS)

    Carney, Kelly S.; Shaker, Francis J.

    1989-01-01

    Two methods for studying the free-vibration characteristics of a large split-blanket solar array in a zero-g cantilevered configuration are presented. The zero-g configuration corrresponds to an on-orbit configuration of the Space Station solar array. The first method applies the equations of continuum mechanics to determine the natural frequencies of the array; the second uses the finite element method program, MSC/NASTRAN. The stiffness matrix from the NASTRAN solution was found to be erroneously grounded. The results from the two methods are compared. It is concluded that the grounding does not seriously compromise the solution to the elastic modes of the solar array. However, the correct rigid body modes need to be included to obtain the correct dynamic model.

  9. Free-vibration characteristics and correlation of a Space Station split-blanket solar array

    NASA Technical Reports Server (NTRS)

    Carney, Kelly S.; Shaker, Francis J.

    1989-01-01

    Two methods for studying the free-vibration characteristics of a large split-blanket solar array in a zero-g cantilevered configuration are presented. The zero-g configuration corresponds to an on-orbit configuration of the Space Station solar array. The first method applies the equations of continuum mechanics to determine the natural frequencies of the array; the second uses the finite element method program, MSC/NASTRAN. The stiffness matrix from the NASTRAN solution was found to be erroneously grounded. The results from the two methods are compared. It is concluded that the grounding does not seriously compromise the solution to the elastic modes of the solar array. However, the correct rigid body modes need to be icluded to obtain the correct dynamic model.

  10. New Monte Carlo results for the TFTR/Lithium Blanket Module system

    SciTech Connect

    Engholm, B.A.

    1985-07-01

    Neutronics analysis results from Phase II of the TFTR Lithium Blanket Module (LBM) program are reported. Principal activities were analyses of new coverplate and protective plate designs; updating of the MCNP Monte Carlo model of TFTR/LBM; and performing new reference calculations for D-D and D-T plasmas. The new protective plate was found to reduce LBM responses by 20%. Updating the model included a new tally structure in which the LBM is divided into 92 volume elements corresponding to foil locations. A new version of the MCNP surface-source routine was used, along with the latest pointwise cross sections. All flux, tritium and foil responses are stored at NMFECC and are available for comparison with measurements, when the experimental program gets underway.

  11. Polonium aspects associated with the use of lead-lithium blankets in fusion applications

    SciTech Connect

    Hoffman, N.J.; Blink, J.A.; Meier, W.R.; Murray, K.A.; Vogelsang, W.F.

    1985-07-01

    Polonium, an alpha-emitting sulfur-like element, is formed by neutron irradiation of lead or bismuth impurity in lead. Design studies of both the Pulse*Star inertial confinement fusion (ICF) reactor and the MARS mirror fusion reactor postulated use of 83Pb-17Li melt as the tritium breeding blanket and coolant. Comparison of the amounts of polonium in the melt at plant shutdown indicated that Pulse*Star would have a far higher level of polonium in the melt. Neutronic considerations and the polonium distribution between the vacuum cleanup system and 83Pb-17Li melt for the two reactors are explored in this paper. Sample neutronics runs showed that the codes used by each design team were not the source of the difference in polonium content.

  12. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  13. Line Blanketing in Vega and Sirus

    NASA Technical Reports Server (NTRS)

    Kurucz, R. L.

    1976-01-01

    A theoretical model and spectrum calculation for Vega is discussed. The abundance of carbon is approximately -3.8, which is 0.3 lower than the old solar value and supports Mount and Linsky's newer value. The oxygen abundance is approximately -3.5. Assuming that Vega has solar abundances, the solar oxygen abundance appears to have been overestimated by 0.3 in the log. Other abundances appear to be solar. For Sirius the calculations do not agree with the observed spectrum. Line opacity is considerably underestimated, notably in third-spectrum iron group lines. Carbon is underabundant relative to Vega by 0.2 in the log. Nitrogen is unchanged. Oxygen is enhanced by 0.3. Heavier elements are enhanced by 1.0 in the log. Calibration yields 1.3E-10 ergs/sq cm/s/nm for each U1 Copernicus count at 130 nm.

  14. Surface heating blanket for soil remediation

    SciTech Connect

    Van Egmond, C.F.; Carl, F.G. Jr.; Stegemeier, G.L.; Vinegar, H.J.

    1993-07-20

    A heater assembly is described for use in soil remediation comprising: a plurality of metallic support rods spaced parallel to each other; a continuous metallic strand spirally encircling adjacent ones of said support rods and forming rungs therearound, said rungs extending the length of said support rods, making low resistance contact therewith but being frictionally movable with respect thereto; an electric beater element located between and parallel to a selected pair of said support rods and between said rungs encircling said selected support rods, said heater being in low resistance frictional contact with said rungs along its length; a layer of insulation on top of said assembly; and an impermeable sheet placed on top of said insulation.

  15. APT {sup 3}He target/blanket. Topical report

    SciTech Connect

    1995-03-01

    The {sup 3}He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D{sub 2}O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process.

  16. Development of insulating coatings for liquid metal blankets

    SciTech Connect

    Malang, S.; Borgstedt, H.U.; Farnum, E.H.; Natesan, K.; Vitkovski, I.V.

    1994-07-01

    It is shown that self-cooled liquid metal blankets are feasible only with electrically insulating coatings at the duct walls. The requirements on the insulation properties are estimated by simple analytical models. Candidate insulator materials are selected based on insulating properties and thermodynamic consideration. Different fabrication technologies for insulating coatings are described. The status of the knowledge on the most crucial feasibility issue, the degradation of the resisivity under irradiation, is reviewed.

  17. Blanket comparison and selection study. Final report. Volume 3

    SciTech Connect

    Not Available

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  18. Blanket comparison and selection study. Final report. Volume 2

    SciTech Connect

    Not Available

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  19. Blanket comparison and selection study. Final report. Volume 1

    SciTech Connect

    Not Available

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  20. MHD considerations for a self-cooled liquid lithium blanket

    NASA Astrophysics Data System (ADS)

    Sze, D. K.; Mattas, R. F.; Hull, A. B.; Picologlou, B. F.; Smith, D. L.

    1992-03-01

    The magnetohydrodynamic (MHD) effects can present a feasibility issue for a self-cooled liquid metal blanket of magnetically confined fusion reactors, especially the inboard regime of a tokamak. This pressure drop can be significantly reduced by using an insulated wall structure. A self-healing insulating coating has been identified, which will reduce the pressure drop by more than a factor of 10. The future research direction to further quantify the performance of this coating is also outlined.

  1. MFTF-B Upgrade for blanket-technology testing

    SciTech Connect

    Thomassen, K.I.; Doggett, J.N.; Logan, B.G.

    1982-10-22

    Based on preliminary studies at Lawrence Livermore National Laboratory (LLNL), we believe the Mirror Fusion Test Facility (MFTF-B) could be upgraded for operation in a hot-ion Kelley mode in a portion of the central cell to provide fusion nuclear engineering data, particularly blanket technology information, by the end of the decade. Cost of this mode of operation would be modest compared with that of the other fusion devices considered in the last few years for such purposes.

  2. Helium-3 blankets for tritium breeding in fusion reactors

    NASA Technical Reports Server (NTRS)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  3. Evaluation of US demo helium-cooled blanket options

    SciTech Connect

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed.

  4. Enhanced plasma current collection from weakly conducting solar array blankets

    NASA Technical Reports Server (NTRS)

    Hillard, G. Barry

    1993-01-01

    Among the solar cell technologies to be tested in space as part of the Solar Array Module Plasma Interactions Experiment (SAMPIE) will be the Advanced Photovoltaic Solar Array (APSA). Several prototype twelve cell coupons were built for NASA using different blanket materials and mounting techniques. The first conforms to the baseline design for APSA which calls for the cells to be mounted on a carbon loaded Kapton blanket to control charging in GEO. When deployed, this design has a flexible blanket supported around the edges. A second coupon was built with the cells mounted on Kapton-H, which was in turn cemented to a solid aluminum substrate. A final coupon was identical to the latter but used germanium coated Kapton to control atomic oxygen attack in LEO. Ground testing of these coupons in a plasma chamber showed considerable differences in plasma current collection. The Kapton-H coupon demonstrated current collection consistent with exposed interconnects and some degree of cell snapover. The other two coupons experienced anomalously large collection currents. This behavior is believed to be a consequence of enhanced plasma sheaths supported by the weakly conducting carbon and germanium used in these coupons. The results reported here are the first experimental evidence that the use of such materials can result in power losses to high voltage space power systems.

  5. NOEL: a no-leak fusion blanket concept

    SciTech Connect

    Powell, J R; Yu, W S; Fillo, J A; Horn, F L; Makowitz, H

    1980-01-01

    Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb/sub 2/, LiPb, Pb) and fused salt choices for material A.

  6. Analysis of LBM (lithium blanket module) experiments at LOTUS

    SciTech Connect

    Stepanek, J.; Davidson, J.W.; Dudziak, D.J.; Haldy, P.A.; Pelloni, S.

    1986-01-01

    A Lithium Blanket Module (LBM) has been designed at General Atomic Company for testing on the Tokamak Fusion Test Reactor (TFTR). The LBM has both realistic fusion blanket materials and configuration and has been designed for detailed experimental analyses of tritium breeding and neutron flux spatial/spectral distributions. It is {approximately}80 cm{sup 3} and the breeding material is Li{sub 2}O. The main objective of the LBM program was to perform a series of experiments by irradiating it using the toroidal neutron source of the TFTR. The comparison between measured and calculated neutron spectra and reaction rates would indicate the accuracy of presently available nuclear data and calculational methods and show the need for future data evaluation and methods development. With a delay in undertaking the deuterium-tritium operation of the Princeton TFTR, it was clear that the LOTUS facility could now provide an extremely valuable resolution of basic technological uncertainties in fusion reactor blanket physics. This resolution has been a major force behind EPRI's fusion program. 6 refs.

  7. Enhanced plasma current collection from weakly conducting solar array blankets

    NASA Astrophysics Data System (ADS)

    Hillard, G. Barry

    1993-05-01

    Among the solar cell technologies to be tested in space as part of the Solar Array Module Plasma Interactions Experiment (SAMPIE) will be the Advanced Photovoltaic Solar Array (APSA). Several prototype twelve cell coupons were built for NASA using different blanket materials and mounting techniques. The first conforms to the baseline design for APSA which calls for the cells to be mounted on a carbon loaded Kapton blanket to control charging in GEO. When deployed, this design has a flexible blanket supported around the edges. A second coupon was built with the cells mounted on Kapton-H, which was in turn cemented to a solid aluminum substrate. A final coupon was identical to the latter but used germanium coated Kapton to control atomic oxygen attack in LEO. Ground testing of these coupons in a plasma chamber showed considerable differences in plasma current collection. The Kapton-H coupon demonstrated current collection consistent with exposed interconnects and some degree of cell snapover. The other two coupons experienced anomalously large collection currents. This behavior is believed to be a consequence of enhanced plasma sheaths supported by the weakly conducting carbon and germanium used in these coupons. The results reported here are the first experimental evidence that the use of such materials can result in power losses to high voltage space power systems.

  8. Continuous fine pattern formation by screen-offset printing using a silicone blanket

    NASA Astrophysics Data System (ADS)

    Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

    2014-09-01

    Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes.

  9. Growth rate reduction of the curvature-driven flute instability by plasma blanket line tying

    SciTech Connect

    Segal, D.

    1983-09-01

    The effect of an annular, line-tied blanket, on the curvature-driven flute in a magnetic mirror is considered. The blanket is assumed to be line tied to a thermoionically emitting annular end plate. Reduction of the flute growth rate is computed as function of Larmor radius, blanket radius, and axial plasma conductance through either an external plasma or mirror sheath. It is found that significant reduction in growth rate can be achieved.

  10. Surge current and electron swarm tunnel tests of thermal blanket and ground strap materials

    NASA Technical Reports Server (NTRS)

    Hoffmaster, D. K.; Inouye, G. T.; Sellen, J. M., Jr.

    1977-01-01

    The results are described of a series of current conduction tests with a thermal control blanket to which grounding straps have been attached. The material and the ground strap attachment procedure are described. The current conduction tests consisted of a surge current examination of the ground strap and a dilute flow, energetic electron deposition and transport through the bulk of the insulating film of this thermal blanket material. Both of these test procedures were used previously with thermal control blanket materials.

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

    SciTech Connect

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

    1986-09-01

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

  12. Helium-cooled, FLiBe-breeder, beryllium-multiplier blanket for MINIMARS

    SciTech Connect

    Moir, R.W.; Lee, J.D.

    1986-06-01

    We adapted the helium-cooled, FLiBe-breeder blanket to the commercial tandem-mirror fusion-reactor design, MINIMARS. Vanadium was used to achieve high performance from the high-energy-release neutron-capture reactions and from the high-temperature operation permitted by the refractory property of the material, which increases the conversion efficiency and decreases the helium-pumping power. Although this blanket had the highest performance among the MINIMARS blankets designs, measured by Mn/sub th/ (blanket energy multiplication times thermal conversion efficiency), it had a cost of electricity (COE) 18% higher than the University of Wisconsin (UW) blanket design (42.5 vs 35.9 mills/kW.h). This increased cost was due to using higher-cost blanket materials (beryllium and vanadium) and a thicker blanket, which resulted in higher-cost central-cell magnets and the need for more blanket materials. Apparently, the high efficiency does not substantially affect the COE. Therefore, in the future, we recommend lowering the helium temperature so that ferritic steel can be used. This will result in a lower-cost blanket, which may compensate for the lower performance resulting from lower efficiency.

  13. Innovative anaerobic/upflow sludge blanket filtration bioreactor for phosphorus removal from wastewater.

    PubMed

    Khorsandi, H; Movahedyan, H; Bina, B; Farrokhzadeh, H

    2011-04-01

    Phosphorus is the key element to remove from aquatic environments to limit the growth of aquatic plants and algae and, thus, to control eutrophication. Because the upflow sludge blanket filtratio' (USBF) process, without addition of metal salts, entails low efficiency for phosphorus removal, we added an anaerobic reactor to the USBF bioreactor in order to promote the simultaneous removal of phosphorus and nitrogen from wastewater. The results revealed that the anaerobic/USBF bioreactor had a phosphorus removal efficiency up to 86%, with a sludge retention time (SRT) of 10 days, a hydraulic retention time (HRT) of 24 hours and an optimum COD/N/P ratio of 100/5/1. This ratio also improved the compaction quality of the sludge blanket in the USBF clarifier. The average specific phosphate uptake rate in the aerobic zone and the average specific phosphate release rate in the anaerobic reactor were 0.014 mg PO4-P removed/(g VSS x min) and 0.0525 mg PO4-P released/(g VSS x min), respectively. Secondary phosphorus release in the USBF clarifier was heightened with increasing HRT. Hence, the optimum total HRT can be selected between 16 and 24 hours based on effluent quality. Effluent phosphorus of about 1 mg/L was provided for wastewater with the COD/N/P ratio of 100/5/1 at the sludge age of 10 days and total HRT of 16 hours. This study illustrated that the anaerobic/USBF bioreactor at the optimum operational conditions can be an effective process for phosphorus removal from municipal wastewater. PMID:21877530

  14. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    SciTech Connect

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods.

  15. Impact of prescribed burning on blanket peat hydrology

    NASA Astrophysics Data System (ADS)

    Holden, Joseph; Palmer, Sheila M.; Johnston, Kerrylyn; Wearing, Catherine; Irvine, Brian; Brown, Lee E.

    2015-08-01

    Fire is known to impact soil properties and hydrological flow paths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied 10 blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments, we studied plots where the last burn occurred ˜2 (B2), 4 (B4), 7 (B7), or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at similar topographic wetness index locations in the control catchments. Plots subject to prescribed vegetation burning had significantly deeper water tables (difference in means = 5.3 cm) and greater water table variability than unburnt plots. Water table depths were significantly different between burn age classes (B2 > B4 > B7 > B10+) while B10+ water tables were not significantly different to the unburnt controls. Overland flow was less common on burnt peat than on unburnt peat, recorded in 9% and 17% of all runoff trap visits, respectively. Storm lag times and hydrograph recession limb periods were significantly greater (by ˜1 and 13 h on average, respectively) in the burnt catchments overall, but for the largest 20% of storms sampled, there was no significant difference in storm lag times between burnt and unburnt catchments. For the largest 20% of storms, the hydrograph intensity of burnt catchments was significantly greater than those of unburnt catchments (means of 4.2 × 10-5 and 3.4 × 10-5 s-1, respectively), thereby indicating a nonlinear streamflow response to prescribed burning. Together, these results from plots to whole river catchments indicate that prescribed vegetation burning has important effects on blanket peatland hydrology at a range of spatial scales.

  16. Accelerator-driven molten-salt blankets: Physics issues

    SciTech Connect

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-10-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m{sup 3} per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics.

  17. ORFEUS-SPAS MAIN TELESCOPE THERMAL BLANKETS ARE INSTALLED

    NASA Technical Reports Server (NTRS)

    1996-01-01

    German technicians in the Multi-Payload Processing Facility at KSC are installing thermal blankets on the main telescope of the Orbiting and Retrievable Far and Extreme Ultraviolet Spectrograph-Shuttle Pallet Satellite (ORFEUS- SPAS) II, mounted atop the Astronomy Shuttle Pallet Satellite (ASTRO-SPAS) platform. The main telescope houses two spectrographs, the Extreme Ultraviolet Spectrograph (EUV) and the Far Ultraviolet Spectrograph (FUV), which will gather data about the life cycle of stars during the flight of ORFEUS-SPAS II on Space Shuttle Mission STS-80 this fall. A third spectrograph, the Interstellar Medium Profile Spectrometer (IMAPS), will be installed on the side of the ASTRO-SPAS platform.

  18. 77 FR 38622 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-28

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on June 4, 2012, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...) regulations under the Natural Gas Act as amended and Southern Star's blanket certificate issued in Docket...

  19. 18 CFR 284.403 - Code of conduct for persons holding blanket marketing certificates.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... persons holding blanket marketing certificates. 284.403 Section 284.403 Conservation of Power and Water... Pipelines § 284.403 Code of conduct for persons holding blanket marketing certificates. (a) To the extent... must provide accurate and factual information, and not knowingly submit false or misleading...

  20. THERMAL INSULATION PROPERTIES OF NONWOVEN SEMI-DISPOSABLE BLANKETS FROM RECYCLED POLYESTER/COTTON FIBERS

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Recycled polyester fibers and cotton fibers that require no chemical processing were used to produce a low-cost, semi-durable, nonwoven thermal blanket. Thermal blankets were given carboxylic acid finish to improve structural stability during use and laundering. A Steady-State Heat Flow meter FOX ...

  1. Design analysis and optimization of self-cooled lithium blankets and shields

    SciTech Connect

    Gohar, Y.

    1988-02-01

    A study of self-cooled lithium blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main design parameters considered during the course of the study were the tritium breeding ratio, the blanket energy multiplication factor, the energy fraction lost to the shield, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Another study was carried out to determine materials, compositions, arrangements, and thickness of the shield zone for the reference blanket. Helium and water-cooled shields were optimized for the inboard and outboard sections of the reactor. Based on the above two studies, the reference blanket and shield configurations were developed for the ANL Tokamak Power Systems Study. The helium-cooled shield was selected for use with liquid metal blankets to reduce safety concerns related to lithium-water reactivity. This helium-cooled shield provides shielding characteristics similar to a conventional water-cooled shield. The analyses and results from these studies are the subject of this paper. 12 refs., 4 figs., 2 tabs.

  2. 78 FR 13663 - Equitrans, L.P. Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-28

    ... Energy Regulatory Commission Equitrans, L.P. Notice of Request Under Blanket Authorization Take notice that on February 12, 2013, Equitrans, L.P. (Equitrans), pursuant to the blanket certificate... open to public inspection. \\1\\ Equitrans, L.P., 85 FERC ] 61,089 (1998). Equitrans proposes to...

  3. 78 FR 30911 - Texas Eastern Transmission, LP; Prior Notice Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-23

    ... Energy Regulatory Commission Texas Eastern Transmission, LP; Prior Notice Activity Under Blanket Certificate On May 8, 2013, Texas Eastern Transmission, LP (Texas Eastern), filed a prior notice request... Act, and Texas Eastern's blanket certificate issued in Docket No. CP82-535-000. Texas Eastern...

  4. Cotton-based hydromulches versus conventional hydromulches and blankets: Erosion and grass establishment

    Technology Transfer Automated Retrieval System (TEKTRAN)

    One commonly used means of reducing the impact of erosion from steep slopes while vegetation is being established is with erosion control products such as roll-out blankets and/or hydromulches. Roll-out blankets are commonly made of wheat straw, coconut husks, or fiberized wood, while the most preva...

  5. 77 FR 52713 - PetroLogistics Natural Gas Storage, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-30

    ... Energy Regulatory Commission PetroLogistics Natural Gas Storage, LLC; Notice of Request Under Blanket Authorization Take notice that on August 17, 2012, PetroLogistics Natural Gas Storage, LLC (PetroLogistics... Iberville Parish, Louisiana, under PetroLogistics' blanket certificate issued in Docket No. CP07-427-000,...

  6. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign delivery... 48 Federal Acquisition Regulations System 4 2014-10-01 2014-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND...

  7. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign delivery... 48 Federal Acquisition Regulations System 4 2013-10-01 2013-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND...

  8. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign delivery... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND...

  9. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign delivery... 48 Federal Acquisition Regulations System 4 2012-10-01 2012-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND...

  10. 76 FR 44324 - Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-25

    ... Energy Regulatory Commission Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization Take notice that on July 14, 2011, Tennessee Gas Pipeline Company (Tennessee), 1001 Louisiana... located in San Jacinto and Liberty Counties, Texas, under Tennessee's blanket certificate issued in...

  11. Inhibition of Frying Oil Oxidation by Carbon Dioxide Blanketing.

    PubMed

    Totani, Nagao; Inoue, Ryota; Yawata, Miho

    2016-06-01

    The oxidation of oil starts, in general, from the penetration of atmospheric oxygen into oil. Inhibition of the vigorous oxidation of oil at deep-frying temperature under carbon dioxide flow, by disrupting the contact between oil and air, was first demonstrated using oil in a round bottom flask. Next, the minimum carbon dioxide flow rate necessary to blanket 4 L of frying oil in an electric fryer (surface area 690 cm(2)) installed with nonwoven fabric cover, was found to be 40 L/h. Then deep-frying of potato was done accordingly; immediately after deep-frying, an aluminum cover was placed on top of the nonwoven fabric cover to prevent the loss of carbon dioxide and the carbon dioxide flow was shut off. In conclusion, the oxidation of oil both at deep-frying temperature and during standing was remarkably inhibited by carbon dioxide blanketing at a practical flow rate and volume. Under the deep-frying conditions employed in this study, the increase in polar compound content was reduced to half of that of the control. PMID:27181248

  12. First wall/blanket/shield design and optimization system

    SciTech Connect

    Gohar, Y.; Baker, C.; Attaya, H.; Cha, Y.; Majumdar, S.; Scandora, T.

    1988-02-01

    First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit under several design constraints, (4) permitting the use of different reactor parameters in the evaluation and optimization analyses, (5) allowing the use of improved eingineering data bases to study the impact on the design performance for planning future research and development, and (6) evaluating the effect of the data base uncertainties on the design performance. BSDOS is the first design and optimization system to couple the highly interacting neutronics, heat transfer, thermal hydraulics, stress analysis, radioactivity and decay-heat analyses, tritium balance, and capital cost. A brief description of the main features of BSDOS is given in this paper. Also, results from using BSDOS to perform design analysis for several reactor components are presented. 17 refs., 1 fig., 2 tabs.

  13. Forced-air patient warming blankets disrupt unidirectional airflow.

    PubMed

    Legg, A J; Hamer, A J

    2013-03-01

    We have recently shown that waste heat from forced-air warming blankets can increase the temperature and concentration of airborne particles over the surgical site. The mechanism for the increased concentration of particles and their site of origin remained unclear. We therefore attempted to visualise the airflow in theatre over a simulated total knee replacement using neutral-buoyancy helium bubbles. Particles were created using a Rocket PS23 smoke machine positioned below the operating table, a potential area of contamination. The same theatre set-up, warming devices and controls were used as in our previous study. This demonstrated that waste heat from the poorly insulated forced-air warming blanket increased the air temperature on the surgical side of the drape by > 5°C. This created convection currents that rose against the downward unidirectional airflow, causing turbulence over the patient. The convection currents increased the particle concentration 1000-fold (2 174 000 particles/m(3) for forced-air warming vs 1000 particles/m(3) for radiant warming and 2000 particles/m(3) for the control) by drawing potentially contaminated particles from below the operating table into the surgical site. Cite this article: Bone Joint J 2013;95-B:407-10. PMID:23450029

  14. Modelling of tritium transport in a pin-type solid breeder blanket

    SciTech Connect

    Martin, R.; Ghoniem, N.M.

    1986-02-01

    This study supplements a larger study of a solid breeder blanket design featuring lithium ceramic pins. This aspect of the study looks at tritium transport, release, and inventory within this blanket design. Li/sub 2/O and ..gamma..-LiAlO/sub 2/ are the two primary candidates for ceramic solid breeders. ..gamma..-LiAlO/sub 2/ was chosen for this blanket design due to its higher structural stability. Analysis of tritium behavior in solid breeder blankets is of great importance due to its impact on several critical issues: the generation of an adequate amount of fusion fuel, the safety-related issue of keeping radioactive blanket inventories as low as possible, and the release, purge, and economical processing of the bred tritium without undue contamination of the coolant and other reactor structures.

  15. Tritium processing system for the ITER Li/V blanket test module

    SciTech Connect

    Sze, D.K.; Hua, T.Q.; Abdou, M.A.; Dagher, M.A.; Waganer, L.M.

    1997-04-01

    The purpose of the ITER Blanket Testing Module is to test the operating and performance of candidate blanket concepts under a real fusion environment. To assure fuel self-sufficiency the tritium breeding, recovery and processing have to be demonstrated. The tritium produced in the blanket has to be processed to a purity which can be used for refueling. All these functions need to be accomplished so that the tritium system can be scaled to a commercial fusion power plant from a safety and reliability point of view. This paper summarizes the tritium processing steps, the size of the equipment, power requirements, space requirements, etc. for a self-cooled lithium blanket. This information is needed for the design and layout of the test blanket ancillary system and to assure that the ITER guidelines for remote handling of ancillary equipment can be met.

  16. 46 CFR 151.50-50 - Elemental phosphorus in water.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 5 2014-10-01 2014-10-01 false Elemental phosphorus in water. 151.50-50 Section 151.50... phosphorus in water. (a) Tanks shall be designed and tested for a head equivalent to the design lading of phosphorus and its water blanket extended to 8 feet above the tank top. In addition, tank design...

  17. 46 CFR 151.50-50 - Elemental phosphorus in water.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 false Elemental phosphorus in water. 151.50-50 Section 151.50... phosphorus in water. (a) Tanks shall be designed and tested for a head equivalent to the design lading of phosphorus and its water blanket extended to 8 feet above the tank top. In addition, tank design...

  18. 46 CFR 151.50-50 - Elemental phosphorus in water.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 5 2013-10-01 2013-10-01 false Elemental phosphorus in water. 151.50-50 Section 151.50... phosphorus in water. (a) Tanks shall be designed and tested for a head equivalent to the design lading of phosphorus and its water blanket extended to 8 feet above the tank top. In addition, tank design...

  19. 46 CFR 151.50-50 - Elemental phosphorus in water.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 false Elemental phosphorus in water. 151.50-50 Section 151.50... phosphorus in water. (a) Tanks shall be designed and tested for a head equivalent to the design lading of phosphorus and its water blanket extended to 8 feet above the tank top. In addition, tank design...

  20. Neutronic analysis of alternative structural materials for fusion reactor blankets

    NASA Astrophysics Data System (ADS)

    Santos, Raul dos

    1988-07-01

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calculated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316.

  1. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    SciTech Connect

    Choi, B. William; Chiu, Ing L.

    2015-10-26

    Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

  2. Tailorable advanced blanket insulation using aluminoborosilicate and alumina batting

    NASA Technical Reports Server (NTRS)

    Calamito, Dominic P.

    1989-01-01

    Two types of Tailorable Advanced Blanket Insulation (TABI) flat panels for Advanced Space Transportation Systems were produced. Both types consisted of integrally woven, 3-D fluted core having parallel faces and connecting ribs of Nicalon yarns. The triangular cross section flutes of one type was filled with mandrels of processed Ultrafiber (aluminoborosilicate) stitchbonded Nextel 440 fibrous felt, and the second type wall filled with Saffil alumina fibrous felt insulation. Weaving problems were minimal. Insertion of the fragile insulation mandrels into the fabric flutes was improved by using a special insertion tool. An attempt was made to weave fluted core fabrics from Nextel 440 yarns but was unsuccessful because of the yarn's fragility. A small sample was eventually produced by an unorthodox weaving process and then filled with Saffil insulation. The procedures for setting up and weaving the fabrics and preparing and inserting insulation mandrels are discussed. Characterizations of the panels produced are also presented.

  3. Heating performances of a IC in-blanket ring array

    SciTech Connect

    Bosia, G.; Ragona, R.

    2015-12-10

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  4. Heating performances of a IC in-blanket ring array

    NASA Astrophysics Data System (ADS)

    Bosia, G.; Ragona, R.

    2015-12-01

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  5. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    SciTech Connect

    DeMuth, J. A.; Meier, W. R.; Jolodosky, A.; Frantoni, M.; Reyes, S.

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  6. Thermal-hydraulic design of the target/blanket for the accelerator production of tritium conceptual design

    SciTech Connect

    Willcutt, G.J.E. Jr.; Kapernick, R.J.

    1997-11-01

    A conceptual design was developed for the target/blanket system of an accelerator-based system to produce tritium. The target/blanket system uses clad tungsten rods for a spallation target and clad lead rods as a neutron multiplier in a blanket surrounding the target. The neutrons produce tritium in {sup 3}He, which is contained in aluminum tubes located in the decoupler and blanket regions. This paper presents the thermal-hydraulic design of the target, decoupler, and blanket developed for the conceptual design of the Accelerator Production of Tritium Project, and demonstrates there is adequate margin in the design at full power operation.

  7. Neutronics experiments for DEMO blanket at JAERI/FNS

    NASA Astrophysics Data System (ADS)

    Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-07-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

  8. Pressure drop considerations of a lithium cooled fusion breeder tokamak reactor blanket

    SciTech Connect

    Wong, C.P.C.

    1983-12-06

    Liquid lithium was selected as one of the coolants for the 1983 fusion breeder blanket used on the magnetically confined tokamak fusion reactor, and as a result, the thermal-hydraulic calculations were dominated by magnetohydrodynamic (MHD) considerations. The applicable sets of MHD equations for the engineering thermal-hydraulic design were reviewed and compared. Special attention was given to the MHD calculations for the fertile material zone, a packed bed of composite beryllium and thorium balls, since this region can dominate the thermal-hydraulic behavior of this blanket module. To keep the pressure drops acceptable, fertile fuel balls were omitted in the inboard blanket.

  9. First wall and blanket module safety enhancement by material selection and design decision

    SciTech Connect

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems.

  10. Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

    NASA Astrophysics Data System (ADS)

    Wang, Hong-yan; Wu, Yi-can; He, Xiao-xong

    2002-10-01

    Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

  11. APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break

    SciTech Connect

    Shadday, M.A.

    1999-11-19

    The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks.

  12. Molten salt cooling//sup 17/Li-/sup 83/Pb breeding blanket concept

    SciTech Connect

    Sze, D.K.; Cheng, E.T.

    1985-02-01

    A description of a fusion breeding blanket concept using draw salt coolant and static /sup 17/Li-/sup 83/Pb is presented. /sup 17/Li-/sup 83/Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator.

  13. In plain sight: the Chesapeake Bay crater ejecta blanket

    NASA Astrophysics Data System (ADS)

    Griscom, D. L.

    2012-02-01

    The discovery nearly two decades ago of a 90 km-diameter impact crater below the lower Chesapeake Bay has gone unnoted by the general public because to date all published literature on the subject has described it as "buried". To the contrary, evidence is presented here that the so-called "upland deposits" that blanket ∼5000 km2 of the U.S. Middle-Atlantic Coastal Plain (M-ACP) display morphologic, lithologic, and stratigraphic features consistent with their being ejecta from the 35.4 Ma Chesapeake Bay Impact Structure (CBIS) and absolutely inconsistent with the prevailing belief that they are of fluvial origin. Specifically supporting impact origin are the facts that (i) a 95 %-pure iron ore endemic to the upland deposits of southern Maryland, eastern Virginia, and the District of Columbia has previously been proven to be impactoclastic in origin, (ii) this iron ore welds together a small percentage of well-rounded quartzite pebbles and cobbles of the upland deposits into brittle sheets interpretable as "spall plates" created in the interference-zone of the CBIS impact, (iii) the predominantly non-welded upland gravels have long ago been shown to be size sorted with an extreme crater-centric gradient far too large to have been the work of rivers, but well explained as atmospheric size-sorted interference-zone ejecta, (iv) new evidence is provided here that ~60 % of the non-welded quartzite pebbles and cobbles of the (lower lying) gravel member of the upland deposits display planar fractures attributable to interference-zone tensile waves, (v) the (overlying) loam member of the upland deposits is attributable to base-surge-type deposition, (vi) several exotic clasts found in a debris flow topographically below the upland deposits can only be explained as jetting-phase crater ejecta, and (vii) an allogenic granite boulder found among the upland deposits is deduced to have been launched into space and sculpted by hypervelocity air friction during reentry. An

  14. 77 FR 73652 - Honeoye Storage Corporation: Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-11

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Honeoye Storage Corporation: Notice of Request Under Blanket Authorization Take notice that on November 16, 2012, Honeoye Storage Corporation (Honeoye) as supplemented...

  15. ITER fast ion confinement in the presence of the European test blanket module

    NASA Astrophysics Data System (ADS)

    Äkäslompolo, Simppa; Kurki-Suonio, Taina; Asunta, Otto; Cavinato, Mario; Gagliardi, Mario; Hirvijoki, Eero; Saibene, Gabriella; Sipilä, Seppo; Snicker, Antti; Särkimäki, Konsta; Varje, Jari

    2015-09-01

    This paper addresses the confinement of thermonuclear alpha particles and neutral beam injected deuterons in the 15 MA Q = 10 inductive scenario in the presence of the magnetic perturbation caused by the helium cooled pebble bed test blanket module using the vacuum approximation. Both the flat top phase and plasma ramp-up are studied. The transport of fast ions is calculated using the Monte Carlo guiding center orbit-following code ASCOT. A detailed three-dimensional wall, derived from the ITER blanket module CAD data, is used for evaluating the fast ion wall loads. The effect of the test blanket module is studied for both overall confinement and possible hot spots. The study indicates that the test blanket modules do not significantly deteriorate the fast ion confinement.

  16. Feasibility study of a fission supressed blanket for a tandem-mirror hybrid reactor

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Barr, W.L.

    1981-10-05

    A study of fission suppressed blankets for the tandem mirror not only showed such blankets to be feasible but also to be safer than fissioning blankets. Such hybrids could produce enough fissile material to support up to 17 light water reactors of the same nuclear power rating. Beryllium was compared to /sup 7/Li for neutron multiplication; both were considered feasible but the blanket with Li produced 20% less fissile fuel per unit of nuclear power in the reactor. The beryllium resource, while possibly being too small for extensive pure fusion application, would be adequate (with carefully planned industrial expansion) for the hybrid because of the large support ratio, and hence few hybrids required. Radiation damage and coatings for beryllium remain issues to be resolved by further study and experimentation.

  17. Role of Fabrication on Materials Compatibility in APT Target/Blanket

    SciTech Connect

    Iyer, N.; Louthan, M.R. Jr.; Dunn, K.; Fisher, D.L.

    1998-09-01

    This paper summarizes several of the options associated with the fabrication of selected target/blanket components. In addition, the materials characterization technologies required to validate these components performance is presented.

  18. 77 FR 26544 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-04

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on April 18, 2012 Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe... authorization to construct and operate certain natural gas transmission facilities in Chesterfield...

  19. 78 FR 20315 - Columbia Gas Transmission, LLC; Notice of Request under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-04

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request under Blanket Authorization Take notice that on Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe, Suite 2500... the application should be directed to Fredric J. George, Senior Counsel, Columbia Gas...

  20. Thin blanket design for MINIMARS - A compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-11-01

    Recent fusion power reactor designs have shown a trend toward lower power, lower cost, higher mass utilization compact configurations with inherent safety, in order to improve the economic aspects of fusion and make them more competitive with other energy sources. Since the blanket thickness directly impacts the size and mass of the remaining reactor components, it is prudent to minimize its thickness while ensuring adequate neutronic and thermal performance. This paper describes the blanket for the MINI-MARS compact tandem mirror fusion power reactor. The blanket which utilizes HT-9 ferritic steel structure, LiPb breeder, Be multiplier/moderator and He gas cooling is only 17 cm thick and is backed up by a steel reflector. Helium gas cools the blanket and reflector in series and the outlet temperature of 575/sup 0/C gives a gross thermal power cycle efficiency of 42.7%.

  1. Climate-driven expansion of blanket bogs in Britain during the Holocene

    NASA Astrophysics Data System (ADS)

    Gallego-Sala, A. V.; Charman, D. J.; Harrison, S. P.; Li, G.; Prentice, I. C.

    2016-01-01

    Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash) based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka) and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts that large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales, and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic) blanket-bog initiation at over half of the sites in the core areas of Scotland and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later expansion of

  2. Climate-driven expansion of blanket bogs in Britain during the Holocene

    NASA Astrophysics Data System (ADS)

    Gallego-Sala, A. V.; Charman, D. J.; Harrison, S. P.; Li, G.; Prentice, I. C.

    2015-10-01

    Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash) based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka) and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic) blanket-bog initiation at over half of the sites in the core areas of Scotland, and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later expansion of blanket

  3. 32 CFR Appendix C to Part 310 - DoD Blanket Routine Uses

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... management inspections conducted under authority of 44 U.S.C. 2904 and 2906. M. Routine Use—Disclosure to the... 32 National Defense 2 2011-07-01 2011-07-01 false DoD Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DOD PRIVACY PROGRAM Pt. 310, App. C Appendix C to Part 310—DoD Blanket Routine Uses...

  4. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    SciTech Connect

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  5. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    SciTech Connect

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  6. 32 CFR Appendix C to Part 806b - DoD ‘Blanket Routine Uses’

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 6 2013-07-01 2013-07-01 false DoD âBlanket Routine Usesâ C Appendix C to Part... PRIVACY ACT PROGRAM Pt. 806b, App. C Appendix C to Part 806b—DoD ‘Blanket Routine Uses’ Certain DoD... the issuance of a license, grant, or other benefit. c. Disclosure of Requested Information Routine...

  7. 32 CFR Appendix C to Part 310 - DoD Blanket Routine Uses

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 2 2010-07-01 2010-07-01 false DoD Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DOD PRIVACY PROGRAM Pt. 310, App. C Appendix C to Part 310—DoD Blanket Routine Uses (See paragraph (c) of § 310.22 of subpart E) A. Routine Use—Law Enforcement If a system of records maintained...

  8. 32 CFR Appendix C to Part 806b - DoD ‘Blanket Routine Uses’

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 6 2011-07-01 2011-07-01 false DoD âBlanket Routine Usesâ C Appendix C to Part... PRIVACY ACT PROGRAM Pt. 806b, App. C Appendix C to Part 806b—DoD ‘Blanket Routine Uses’ Certain DoD... the issuance of a license, grant, or other benefit. c. Disclosure of Requested Information Routine...

  9. First wall and blanket design for a high wall loading compact tokamak power reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Abdel-Khalik, S.I.; Corradini, M.L.; El-Afify, M.; Huh, K.Y.; Kuleinski, G.L.; Wittenberg, L.J.

    1985-07-01

    Among the specific limitations which tend to complicate a compact high wall loading (HWL) tokamak reactor design are high surface and nuclear heating, compactness leading to crowded components, unlikely breeding on the inboard side and frequent first wall/blanket replacement. This paper describes the mechanical, thermal hydraulic and tritium aspects of an improved blanket design for a high ..beta.. (20%), high wall loading (R 10 MW/m/sup 2/) compact fusion power reactor of 1000 MW /sub th/ power output.

  10. Multiplier, moderator, and reflector materials for lithium-vanadium fusion blankets.

    SciTech Connect

    Gohar, Y.; Smith, D. L.

    1999-10-07

    The self-cooled lithium-vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium-vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at high loading conditions of 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.

  11. Liquid immersion blanket design for use in a compact modular fusion reactor

    NASA Astrophysics Data System (ADS)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  12. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    NASA Astrophysics Data System (ADS)

    Huang, Qun-Ying; Li, Jian-Gang; Chen, Yi-Xue

    2004-12-01

    Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  13. Immobilization effect of air-injected blanket (AIB) for abdomen fixation

    SciTech Connect

    Ko, Young Eun; Suh, Yelin; Ahn, Seung Do; Lee, Sang-wook; Shin, Seong Soo; Kim, Jong Hoon; Choi, Eun Kyung; Yi, Byong Yong

    2005-11-15

    A new device for reducing the amplitude of breathing motion by pressing a patient's abdomen using an air-injected blanket (AIB) for external beam radiation treatments has been designed and tested. The blanket has two layers sealed in all four sides similar to an empty pillow made of urethane. The blanket is spread over the patient's abdomen with both ends of the blanket fixed to the sides of the treatment couch or a baseboard. The inner side, or patient side, of the blanket is thinner and expands more than the outer side. When inflated, the blanket balloons and effectively puts an even pressure on the patient's abdomen. Fluoroscopic observation was performed to verify the usefulness of AIB for patients with lung, breast cancer, or abdominal cancers. Internal organ movement due to breathing was monitored and measured with and without AIB. With the help of AIB, the average range of diaphragm motion was reduced from 2.6 to 0.7 cm in the anterior-to-posterior direction and from 2.7 to 1.3 cm in the superior-to-inferior direction. The motion range in the right-to-left direction was negligible, for it was less than 0.5 cm. These initial testing demonstrated that AIB is useful for reducing patients' breathing motion in the thoracic and abdominal regions comfortably and consistently.

  14. (Deuterium-deuterium)-driven experimental hybrid blankets and their neutronic analyses

    SciTech Connect

    Kumar, A.; Sahin, S.

    1984-09-01

    The impressive progress made so far toward the achievement of the physics goal of ignited fusion fuel of deuterium-tritium (D-T) is stirring the scientific community to look back and work for the earliest possible introduction of advanced fusion fuel based reactors with the ultimate objective of very clean, safe, and limitless fusion power. As the introduction of advanced fuel fusion drivers is expected to be in phases due to energetics considerations, it is quite instructive to examine the neutronic aspects of deuterium-deuterium (D-D) neutron driven hybrid blankets. The neutronics investigations of some compact hybrid blankets that could be tested experimentally are presented. The blanket designs are selected to conform to a rather small experimental chamber of the LOTUS fusionfission hybrid facility. The parallelepiped-shaped blankets are driven by a (D-D) neutron source from one side. The fertile fuel is either ThO/sub 2/, natural UO/sub 2/, or LOTUS UO/sub 2/. The tritium breeders are chosen from lithium, LiAlO/sub 2/, or Li/sub 2/O. The relative performances of different fertile fuels and tritium breeders are compared. The performance characteristics of ThO/sub 2/ blankets driven by (D-T) and (D-D) neutrons are compared. The improvement in performance characteristics obtained by the introduction of actinides as multipliers with ThO/sub 2/ hybrid blankets is also investigated.

  15. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    SciTech Connect

    Humrickhouse, Paul Weston; Merrill, Brad Johnson

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  16. A code for line blanketing without local thermodynamic equilibrium

    NASA Astrophysics Data System (ADS)

    Anderson, L. S.

    A numerical code has been written which is designed to calculate radiation transport and atmospheric structure under the constraints of statistical equilibrium in atomic transitions and radiative and hydrostatic equilibrium in the medium. In addition to the complete linearization and variable Eddington factor techniques of Auer and Mihalas, it uses a multi-frequency/multi-gray algorithm which admits the inclusion of many spectral lines in full statistical equilibrium. The program can comfortably accept up to about 300 specific lines arising from about 30 lower states and any number of continua. Cleverly constructed artificial model atoms can extend the number of lines to 3000 or more, where opacity sampling techniques can begin to approximate the blanketing accomplished by Kurucz in LTE. By way of example, a model of a stellar atmosphere is presented with effective temperature 35,000 K and surface gravity 10 to the 4 cm per second squared. The calculation includes 98 bound-free transitions and 93 bound-bound transitions (57 with radiative rates) between 91 states in 36 ions of nine cosmically abundant species.

  17. Smoke from Canadian Fires Blankets Eastern U.S.

    NASA Technical Reports Server (NTRS)

    2002-01-01

    Smoke from multiple large wildfires in Quebec is blanketing the southern portion of the Canadian province and extending southward over the Great Lakes and eastern United States. This image was acquired by the Moderate Resolution Imaging Spectroradiometer (MODIS) on the Terra satellite on July 7, 2002, and shows dozens of active fire detections (red dots) east of James Bay at upper left. The enormous smoke plume is almost 200 miles wide where it enters the United States over the New York and Vermont state lines. The thick pall is affecting air quality in places well to the south, including New York, Baltimore, and Washington, D.C. The image shows the smoke drifting out over the Atlantic Ocean, and then curling back in over North Carolina (bottom right). On Sunday, July 7, the Canadian Interagency Forest Fire Center reported 15 new fires in Quebec in the preceding 24 hours, bringing the total to more than 40 fires in the region, at least 7 which were burning out of control. Most of the fires are believed to have been caused by lightning, more of which is expected on Monday. According to news reports, several hundred people remain evacuated from their homes. Image by Jesse Allen, NASA Earth Observatory, based upon data provided by the MODIS Land Rapid Response Team at NASA GSFC

  18. Axial blanket fuel design and demonstration. First semi-annual progress report, January-September 1980. [PWR

    SciTech Connect

    Not Available

    1980-11-01

    The axial blanket fuel design in this program, which is retrofittable in operating pressurized water reactors, involves replacing the top and bottom of the enriched fuel column with low-enriched (less than or equal to 1.0 wt % /sup 235/U) fertile uranium. This repositioning of the fissile inventory in the fuel rod leads to decreased axial leakage and increased discharge burnups in the enriched fuel. Various axial blanket fuel designs, with blanket thicknesses from 0 to 10 inches and blanket enrichments from 0.2 to 1.0 wt % /sup 235/U, were investigated to determine the relationship between uranium utilization and power peaking. Analyses were preformed to assess the nuclear, mechanical, and thermal-hydraulic effects arising from the use of axial blankets. Four axial blanket lead test assemblies are being fabricated for scheduled irradiation in cycle 5 of Sacramento Municipal Utility District's Rancho Seco pressurized water reactor. Analyses to support licensing cycle 5 are in progress.

  19. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    SciTech Connect

    Jolodosky, A.; Fratoni, M.

    2015-09-22

    Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

  20. A passively-safe fusion reactor blanket with helium coolant and steel structure

    SciTech Connect

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  1. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    SciTech Connect

    Patterson-Hine, F.A.

    1984-05-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown.

  2. Sensitivity of blanket peat vegetation and hydrochemistry to local disturbances.

    PubMed

    Robroek, Bjorn J M; Smart, Richard P; Holden, Joseph

    2010-10-01

    At the ecosystem scale, peatlands can be extremely resilient to perturbations. Yet, they are very sensitive to local disturbances, especially mechanical perturbations (e.g. trampling). The effects of these disturbances on vegetation, and potential effects on hydrochemical conditions along the peat surface, however, are largely unknown. We used three research tracks (paths researchers use to access their study sites) differing in time of abandonment to investigate the impact of local disturbance (trampling) on the vegetation and its short-term (< or = 2 year) recovery in a flagship research blanket peatland. Additionally, we examined the effects of local disturbance on fluvial runoff events and the concentrations of dissolved organic carbon (DOC) and particulate organic carbon (POC) in runoff water. Local disturbance heavily impacted peat vegetation, resulting in large areas of scarred and churned peat. Recovery of vascular plants along abandoned tracks was slow, but a functional Sphagnum layer re-established after just one year. The absence of vegetation elicited an increase in the number of runoff events along the tracks, by which POC runoff from the tracks increased. POC concentrations were highest in the surface water from the recently abandoned track, while they were low in the runoff water from the track abandoned longest and the undisturbed control track. We attribute this to the relatively fast recovery of the Sphagnum vegetation. DOC concentrations did not differ significantly either spatially or temporally in surface runoff or soil solution waters. While at an ecosystem scale local disturbances may be negligible in terms of carbon loss, our data points to the need for further research on the potential long-term effects of local disturbance on the vegetation, and significant effects on local scale carbon fluxes. Moreover, the effects of disturbances could be long-lasting and their role on ecosystem processes should not be underestimated. PMID:20692016

  3. ITER Test Blanket Module Error Field Simulation Experiments

    NASA Astrophysics Data System (ADS)

    Schaffer, M. J.

    2010-11-01

    Recent experiments at DIII-D used an active-coil mock-up to investigate effects of magnetic error fields similar to those expected from two ferromagnetic Test Blanket Modules (TBMs) in one ITER equatorial port. The largest and most prevalent observed effect was plasma toroidal rotation slowing across the entire radial profile, up to 60% in H-mode when the mock-up local ripple at the plasma was ˜4 times the local ripple expected in front of ITER TBMs. Analysis showed the slowing to be consistent with non-resonant braking by the mock-up field. There was no evidence of strong electromagnetic braking by resonant harmonics. These results are consistent with the near absence of resonant helical harmonics in the TBM field. Global particle and energy confinement in H-mode decreased by <20% for the maximum mock-up ripple, but <5% at the local ripple expected in ITER. These confinement reductions may be linked with the large velocity reductions. TBM field effects were small in L-mode but increased with plasma beta. The L-H power threshold was unaffected within error bars. The mock-up field increased plasma sensitivity to mode locking by a known n=1 test field (n = toroidal harmonic number). In H-mode the increased locking sensitivity was from TBM torque slowing plasma rotation. At low beta, locked mode tolerance was fully recovered by re-optimizing the conventional DIII-D ``I-coils'' empirical compensation of n=1 errors in the presence of the TBM mock-up field. Empirical error compensation in H-mode should be addressed in future experiments. Global loss of injected neutral beam fast ions was within error bars, but 1 MeV fusion triton loss may have increased. The many DIII-D mock-up results provide important benchmarks for models needed to predict effects of TBMs in ITER.

  4. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    NASA Astrophysics Data System (ADS)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  5. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

  6. Liquid metal blanket module testing and design for ITER/TIBER II

    SciTech Connect

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs.

  7. Thin blanket design for MINIMARS - a compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-01-01

    A primary goal in the MINIMARS fusion power reactor design is to achieve the lowest possible cost of electricity and highest mass utilization while maintaining credibility and passive safety. Because the blanket impacts many components, reducing its thickness-while achieving adequate breeding and a high energy multiplication-was of prime importance. The MINIMARS blanket is a helium-gas-cooled design using 17Li-83Pb (LiPb) breeder, HT-9 structure, and beryllium moderator/multiplier. The helium gas is contained in small tubes that are immersed in a close-packed matrix of beryllium balls and LiPb. The result is a compact blanket only 18 cm thick in which only the tubes are operated in a stressed condition, but the blanket structure is designed to withstand a helium gas leak in one of the tubes. By circulating the helium gas from the blanket into the reflector, the reflector energy is recovered at a high temperature giving a gross power cycle efficiency of 42.7% while maintaining a low interface temperature between the breeding material and structure.

  8. Tritium processing for the European test blanket systems: current status of the design and development strategy

    SciTech Connect

    Ricapito, I.; Calderoni, P.; Poitevin, Y.; Aiello, A.; Utili, M.; Demange, D.

    2015-03-15

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  9. Electrically insulating coatings for V-Li self-cooled blanket in a fusion system

    SciTech Connect

    Natesan, K.; Reed, C. B.; Uz, M.; Park, J. H.; Smith, D. L.

    2000-05-17

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The liquid-metal blanket concept requires an electrically insulating coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions between the coating and the liquid lithium on one side and the structural V-base alloy on the other side, several coating candidates are being examined to perform the insulating function over a wide range of temperatures and lithium chemistries.

  10. Analysis of MHD Pressure Drop in the Packed Pebble Bed-Based Blanket for the Fds

    NASA Astrophysics Data System (ADS)

    Wang, Hongyan; Wu, Yican; He, Xiaoxiong

    2003-06-01

    The Fusion-Driven Sub-critical System as a multifunctional hybrid reactor has been investigated in ASIPP. The liquid metal LiPb flow through a packed pebble bed-based blanket is considered to be one of the blanket candidates. In this contribution, the MHD pressure drop of liquid metal flow through the packed pebble bed has been calculated and analyzed under various conditions including (a) the size of the packed pebbles; (b) the ratio of occupied room by the packed pebbles to that of liquid metal; and (c) whether the pebbles surface is insulated or not Furthermore, asymptotic techniques to analyze large Hartmann parameter flow and interaction parameter flow are employed and an analytical model has been developed for the calculations of MHD pressure drop of liquid metal flow in a packed pebble bed. The appropriate method for calculating the MHD effects on the pressure drop through the packed pebble bed-based blanket for the FDS has been presented.

  11. Neutronic optimization of a LiAlO/sub 2/ solid breeder blanket

    SciTech Connect

    Levin, P.; Ghoniem, N.M.

    1986-02-01

    In this report, a pressurized lobular blanket configuration is neutronically optimized. Among the features of this blanket configuration are the use of beryllium and LiAlO/sub 2/ solid breeder pins in a cross-flow configuration in a helium coolant. One-dimensional neutronic optimization calculations are performed to maximize the tritium breeding ratio (TER). The procedure involves spatial allocations of Be, LiAlO/sub 2/, 9-C (ferritic steel), and He; in such a way as to maximize the TBR subject to several material, engineering and geometrical constraints. A TBR of 1.17 is achieved for a relatively thin blanket (approx. = 43 cm depth), and consistency with all imposed constraints.

  12. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    SciTech Connect

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250/sup 0/C).

  13. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    NASA Astrophysics Data System (ADS)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  14. Lithium Blanket Module (LBM) dosimetry measurements at the LOTUS 14-MeV neutron source facility

    SciTech Connect

    Tsang, F.Y.; Leo, W.; Sahraoui, C.; Wuthrich, S.; Shaer, M.

    1986-11-01

    A series of passive dosimetry irradiation experiments were performed inside the Lithium Blanket Module (LBM) with the 14-MeV neutron source at the Ecole Polytechnique Federale de Lausane (EPFL). Sets of passive dosimetry foils were utilized to measure fusion-reactor-blanket neutronic environments. The dosimeter reaction data are analyzed and compared with calculational models. These experimental results demonstrate the ability to simulate low power deuterium-tritium (D-T) plasma shots by measuring the neutron field in a reactor-representative fusion blanket environment. The dosimeter results can determine the entire neutron spectrum along the full length of the LBM test rod. The set of selected dosimetry materials meets the requirements of neutronic characterization in future LBM-TFTR D-T and high power deuterium-deuterium (D-D) plasma experiments.

  15. Warm Ocean Temperatures Blanket the Far-Western Pacific

    NASA Technical Reports Server (NTRS)

    2001-01-01

    These data, taken during a 10-day collection cycle ending March 9, 2001, show that above-normal sea-surface heights and warmer ocean temperatures(indicated by the red and white areas) still blanket the far-western tropical Pacific and much of the north (and south) mid-Pacific. Red areas are about 10centimeters (4 inches) above normal; white areas show the sea-surface height is between 14 and 32 centimeters (6 to 13 inches) above normal.

    This build-up of heat dominating the Western Pacific was first noted by TOPEX/Poseidon oceanographers more than two years ago and has outlasted the El Nino and La Nina events of the past few years. See: http://www.jpl.nasa.gov/elnino/990127.html . This warmth contrasts with the Bering Sea, Gulf of Alaska and tropical Pacific where lower-than-normal sea levels and cool ocean temperatures continue (indicated by blue areas). The blue areas are between 5 and 13centimeters (2 and 5 inches) below normal, whereas the purple areas range from 14 to 18 centimeters (6 to 7 inches) below normal. Actually, the near-equatorial ocean cooled through the fall of 2000 and into mid-winter and continues almost La Nina-like.

    Looking at the entire Pacific basin, the Pacific Decadal Oscillation's warm horseshoe and cool wedge pattern still dominates this sea-level height image. Most recent National Oceanic and Atmospheric Administration (NOAA) sea-surface temperature data also clearly illustrate the persistence of this basin-wide pattern. They are available at http://psbsgi1.nesdis.noaa.gov:8080/PSB/EPS/SST/climo.html

    The U.S.-French TOPEX/Poseidon mission is managed by JPL for NASA's Earth Science Enterprise, Washington, D.C. JPL is a division of the California Institute of Technology in Pasadena. For more information on the TOPEX/Poseidon project, see: http://topex-www.jpl.nasa.gov

  16. APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling.

  17. 78 FR 11867 - CenterPoint Energy Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-20

    ... Energy Regulatory Commission CenterPoint Energy Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization Take notice that on January 31, 2013, CenterPoint Energy Gas Transmission Company... Commission's regulations under the Natural Gas Act (NGA), and CenterPoint's blanket certificate authorized...

  18. 76 FR 2093 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-12

    ... Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... November 30, 2010, by Eni USA Gas Marketing LLC (Eni USA), requesting blanket authorization to export... purchasing and marketing supplies of natural gas and LNG. Eni USA is a customer of the Cameron Terminal...

  19. 75 FR 38459 - Certain Woven Electric Blankets From the People's Republic of China: Final Determination of Sales...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-02

    ...The Department of Commerce (``the Department'') has determined that certain woven electric blankets (``woven electric blankets'') from the People's Republic of China (``PRC'') are being, or are likely to be, sold in the United States at less than fair value (``LTFV'') as provided in section 735 of the Tariff Act of 1930, as amended (``the Act''). The final dumping margins for this......

  20. Concept for testing fusion first wall/blanket systems in existing nuclear facilities

    SciTech Connect

    Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

    1980-10-01

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment (except the 14 MeV neutron component) employing an existing nuclear facility is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of a test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module.

  1. Development of tritium breeding blankets for DT-burning fusion reactors

    SciTech Connect

    Clemmer, R.G.

    1980-01-01

    This study examines the status of understanding of blanket tritium recovery and the performance of potentially viable tritium breeding materials under conditions anticipated in a DT-fueled fusion reactor environment. The existing physicochemical, thermophysical, and ceramographic data for candidate liquid and solid breeders are reviewed and appropriate operating conditions defined. It is shown that selection of a breeding material and an appropriate tritium recovery method can impose significant constraints upon blanket design, particularly when considerations of breeder/coolant/structure compatibility and temperature limitations are taken into account.

  2. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    SciTech Connect

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper.

  3. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  4. 77 FR 64982 - WBI Energy Transmission, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-24

    ... this Application should be directed to Keith A. Tiggelaar, Director of Regulatory Affairs, WBI Energy... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission WBI Energy Transmission, Inc.; Notice of Request Under Blanket...

  5. 78 FR 30918 - Perryville Gas Storage LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Perryville Gas Storage LLC; Notice of Request Under Blanket Authorization Take notice that on May 3, 2013, Perryville Gas Storage LLC (Perryville), Three Riverway, Suite...

  6. 76 FR 13612 - Freebird Gas Storage, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-14

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Freebird Gas Storage, LLC; Notice of Request Under Blanket Authorization Take notice that on March 1, 2011, Freebird Gas Storage, LLC (Freebird) filed a Prior Notice...

  7. Effects of the LDEF environment on the Ag/FEP thermal blankets

    NASA Technical Reports Server (NTRS)

    Levadou, Francois; Pippin, H. Gary

    1992-01-01

    This presentation was made by Francois Levadou at the NASA Langley Research Center LDEF materials workshop, November 19-22, 1991. It represents the results to date on the examination of silvered teflon thermal blankets primarily from the Ultra-heavy Cosmic Ray Experiment and also from the blanket from the Park Seed Company experiment. ESA/ESTEC and Boeing conducted a number of independent measurements on the blankets and in particular on the exposed fluorinated ethylene-propylene (FEP) layer of the blankets. Mass loss, thickness, and thickness profile measurements have been used by ESA, Boeing, and NASA LeRC to determine recession and average erosion yield under atomic oxygen exposure. Tensile strength and percent elongation to failure data, surface characterization by ESCA, and SEM images are presented. The Jet Propulsion Laboratory analysis of vacuum radiation effects is also presented. The results obtained by the laboratories mentioned and additional results from the Aerospace Corporation on samples provided by Boeing are quite similar and give confidence in the validity of the data.

  8. 78 FR 76827 - Midwestern Gas Transmission Company; Prior Notice of Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Midwestern Gas Transmission Company; Prior Notice of Activity Under Blanket Certificate On December 4, 2013, Midwestern Gas Transmission Company (Midwestern) filed with the Federal Energy Regulatory Commission...

  9. Neutronics investigation of advanced self-cooled liquid blanket systems in the helical reactor

    NASA Astrophysics Data System (ADS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M. Z.

    2008-03-01

    Neutronics investigations have been conducted in the design activity of the helical-type reactor Force Free Helical Reactor (FFHR2) adopting Flibe-cooled and Li-cooled advanced liquid blanket systems. In this study, comprehensive investigations and geometry modifications related to the tritium breeding ratios (TBRs), neutron shielding performance and neutron wall loading on the first walls in FFHR2 have been performed by improving the three-dimensional (3D) neutronics calculation system developed for non-axisymmetric helical designs. The total TBRs obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. However, it appeared that the most important neutronics issue in the present helical blanket configuration was suppression of neutron streaming through the divertor pumping areas and reflection from support structures for protection of poloidal and helical coils. Evaluation of neutron wall loading on the first walls indicated that the peaking factor would be moderated as low as 1.2 by the toroidal and helical effect of the helical-shaped plasma distribution in the helical reactor.

  10. 78 FR 63179 - Notice of Request Under Blanket Authorization; Petal Gas Storage, LLC.

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Notice of Request Under Blanket Authorization; Petal Gas Storage, LLC. Take notice that on October 9, 2013, Petal Gas Storage, L.L.C. (Petal), 9 Greenway Plaza, Suite 2800, Houston, Texas 77046, filed in Docket No....

  11. 29 CFR 2580.412-10 - Individual or schedule or blanket form of bonds.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...) Individual bond. Covers a named individual in a stated penalty. (b) Name schedule bond. Covers a number of... 1974 TEMPORARY BONDING RULES Scope and Form of the Bond § 2580.412-10 Individual or schedule or blanket form of bonds. Section 13 provides that “any bond shall be in a form or of a type approved by...

  12. Prevention of perioperative hypothermia with forced-air warming systems and upper-body blankets.

    PubMed

    Perl, Thorsten; Bräuer, Anselm; Quintel, Michael

    2006-01-01

    Forced-air warming is known as an effective procedure in prevention and treatment of perioperative hypothermia. Hypothermia is associated with disturbances of coagulation, raises postoperative oxygen consumption by shivering, increases cardiac morbidity, leads to a higher incidence of wound infection, and prolongs hospital stay. Additionally, preoperative local warming reduces the incidence of wound infection after clean surgery. In an animal experiment it has been demonstrated that even during large abdominal operations the major source of heat loss was the skin. Although evaporation accounted for the largest heat loss from the abdominal cavity, it was a minor source due to the smaller heat losing area. As a consequence, reduction of heat loss from the skin is the most promising approach to avoid hypothermia. During abdominal surgery and lower-limb surgery, the use of upper blankets is favourable. The use of upper-body blankets implies a reduction of heat loss in a relevant area and, furthermore, a heat gain. The covered area is approximately 0.35 m2, or approximately 15%-20% of body surface. The heat balance in this area can be changed by 46.1W to 55.0W by forced-air warming systems with upper body blankets. Depending on the surgical procedure and resulting fluid demand, forced-air warming with upper-body blankets-in combination with insulation and fluid warming-is an effective method to prevent perioperative hypothermia. PMID:17029156

  13. Depth of Blanket. Operational Control Tests for Wastewater Treatment Facilities. Instructor's Manual [and] Student Workbook.

    ERIC Educational Resources Information Center

    Arasmith, E. E.

    The determination of the thickness of a sludge blanket in primary and secondary clarifiers and in gravity thickness is important in making operational control decisions. Knowing the thickness and concentration will allow the operator to determine sludge volume and detention time. Designed for individuals who have completed National Pollutant…

  14. 77 FR 58125 - Trunkline Gas Company, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Trunkline Gas Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on August 30, 2012, Trunkline Gas Company, L.L.C. (Trunkline), P.O. Box 4967, Houston, Texas, 77210, filed in Docket No....

  15. 75 FR 32460 - Natural Gas Pipeline Company of America LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-08

    ... Energy Regulatory Commission Natural Gas Pipeline Company of America LLC; Notice of Request Under Blanket Authorization May 28, 2010. Take notice that on May 14, 2010, Natural Gas Pipeline Company of America LLC... Products and Services, Natural Gas Pipeline Company of America LLC, 3250 Lacey Road, 7th Floor,...

  16. 76 FR 48853 - Natural Gas Pipeline Company of America LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-08-09

    ... Energy Regulatory Commission Natural Gas Pipeline Company of America LLC; Notice of Request Under Blanket Authorization Take notice that on July 19, 2011, Natural Gas Pipeline Company of America LLC (Natural), 3250... Products and Services, Natural Gas Pipeline Company of America LLC, 3250 Lacey Road, 7th Floor,...

  17. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... authority of 44 U.S.C. 2904 and 2906. (m) Routine Use—Disclosure to the Merit Systems Protection Board. A... 32 National Defense 2 2011-07-01 2011-07-01 false DeCA Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DEFENSE COMMISSARY AGENCY PRIVACY ACT PROGRAM Pt. 327, App. C Appendix C to Part...

  18. Copyright Center Will Let Colleges Pay Blanket Fees to Reuse Print Material

    ERIC Educational Resources Information Center

    Read, Brock

    2007-01-01

    This article reports on an annual copyright license for colleges created by the Copyright Clearance Center, a nonprofit group that manages licenses for the reuse of published material, that will allow institutions to pay a blanket fee to use copyrighted material instead of securing the rights to such content on a case-by-case basis. The blanket…

  19. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... date of the certificate. The obligation to sell at the cost-based rate expires one year after the... for unbundled sales services. 284.284 Section 284.284 Conservation of Power and Water Resources... Sales by Interstate Pipelines § 284.284 Blanket certificates for unbundled sales services....

  20. 78 FR 66915 - Notice of Request Under Blanket Authorization; Southern Star Central Gas Pipeline, Inc.

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-07

    ... Energy Regulatory Commission Notice of Request Under Blanket Authorization; Southern Star Central Gas Pipeline, Inc. Take notice that on October 21, 2013 Southern Star Central Gas Pipeline, Inc. (Southern Star... in Johnson and Pettis Counties, Missouri, under authorization issued to Southern Star in Docket...

  1. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-28

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 11, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star... is on file with the Commission and open for public inspection. Specifically, Southern Star...

  2. 78 FR 53746 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-30

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on August 13, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star..., Chautauqua and Montgomery Counties, Kansas. Southern Star's prior notice request is more fully set forth...

  3. 77 FR 14517 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-12

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 21, 2012 Southern Star Central Gas Pipeline, Inc. (Southern Star... City, Missouri. Specifically, Southern Star proposes to replace 3 miles of 12-inch diameter XT...

  4. 75 FR 8053 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-23

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization February 16, 2010. Take notice that on January 29, 2010, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 State Highway 56, Owensboro, Kentucky 42301, filed in Docket No. CP10-48-000, a...

  5. 78 FR 68835 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-15

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on October 31, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star...). Southern Star seeks authorization to increase the Maximum Operating Pressure (MOP) of its Waynoka...

  6. 78 FR 25264 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-30

    ... Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on April 16, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...) 208-3676 (toll free). For TTY, call (202) 502-8659. Specifically, Southern Star proposes to abandon...

  7. Performance testing of the 5 kW EOS AM-1 flexible solar array blanket

    SciTech Connect

    Schurig, H.H.; Kruer, M.A.; Levesque, M.N.; Gaddy, E.M.

    1997-12-31

    A GaAs/Ge flexible solar array blanket has been developed for use on the NASA/GSFC remote sensing EOS AM-1 spacecraft. This single wing array has been designed to provide a 5 kW of power after five years in a low earth polar orbit. The blanket configuration includes design features such as thin GaAs/Ge cell stacks mounted on a large flexible, hinged substrate, parallel connected solar cell strings providing high voltage output, a printed circuit harness, and a multi-layer jumper bus providing electrical continuity between the cell strings and the printed circuit harness. This work was contracted to TRW Space and Electronics Group in 1993 by Lockheed Martin Missiles and Space (LMMS). This paper presents the essential design of the EOS AM-1 solar array blanket, and summarizes the results of a qualification test program designed to demonstrate adequate design margins and to assess the performance of the mechanical and electrical components after exposure to a simulated mission space environment. It also reviews the complexities of performing electrical output on a 8.9 m x 5.0 m deployed solar array blanket under AM0 conditions.

  8. 76 FR 28972 - Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on April 28, 2011, Eastern Shore Natural Gas Company (Eastern Shore), 1110 Forrest Avenue, Dover, Delaware,...

  9. 76 FR 20659 - Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-13

    ... From the Federal Register Online via the Government Publishing Office ] DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on April 1, 2011, Eastern Shore Natural Gas Company (Eastern Shore), 1110 Forrest Avenue, Dover, Delaware,...

  10. 76 FR 28972 - Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-19

    ... From the Federal Register Online via the Government Publishing Office ] DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Eastern Shore Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on April 28, 2011, Eastern Shore Natural Gas Company (Eastern Shore), 1110 Forrest Avenue, Dover, Delaware...

  11. 29 CFR 2580.412-10 - Individual or schedule or blanket form of bonds.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... ADMINISTRATION, DEPARTMENT OF LABOR TEMPORARY BONDING RULES UNDER THE EMPLOYEE RETIREMENT INCOME SECURITY ACT OF 1974 TEMPORARY BONDING RULES Scope and Form of the Bond § 2580.412-10 Individual or schedule or blanket.... Bonding, to the extent required, of persons indirectly employed, or otherwise delegated, to...

  12. 77 FR 47629 - Dominion Transmission, Inc.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-09

    ... Federal Energy Regulatory Commission's regulations under the Natural Gas Act (NGA), and Dominion's blanket... the Commission's staff may, pursuant to section 157.205 of the Commission's Regulations under the NGA... section 7 of the NGA. Persons who wish to comment only on the environmental review of this project...

  13. Materials data base and design equations for the UCLA solid breeder blanket

    SciTech Connect

    Sharafat, S.; Amodeo, R.; Ghoniem, N.M.

    1986-02-01

    The materials and properties investigated for this blanket study are listed. The phenomenological equations and mathematical fits for all materials and properties considered are given. Efforts to develop a swelling equation based on the few experimental data points available for breeder materials are described. The sintering phenomena for ceramics is investigated.

  14. Study of the effects of corrugated wall structures due to blanket modules around ICRH antennas

    SciTech Connect

    Dumortier, Pierre; Louche, Fabrice; Messiaen, André; Vervier, Michel

    2014-02-12

    In future fusion reactors, and in ITER, the first wall will be covered by blanket modules. These blanket modules, whose dimensions are of the order of the ICRF wavelengths, together with the clearance gaps between them will constitute a corrugated structure which will interact with the electromagnetic waves launched by ICRF antennas. The conditions in which the grooves constituted by the clearance gaps between the blanket modules can become resonant are studied. Simple analytical models and numerical simulations show that mushroom type structures (with larger gaps at the back than at the front) can bring down the resonance frequencies, which could lead to large voltages in the gaps between the blanket modules and perturb the RF properties of the antenna if they are in the ICRF operating range. The effect on the wave propagation along the wall structure, which is acting as a spatially periodic (toroidally and poloidally) corrugated structure, and hence constitutes a slow wave structure modifying the wall boundary condition, is examined.

  15. 76 FR 58488 - Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-21

    ... Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported Liquefied Natural Gas AGENCY: Office of Fossil Energy, DOE. ACTION: Notice of application. SUMMARY: The Office of Fossil Energy (FE) of the Department of Energy (DOE) gives notice of receipt of an...

  16. 75 FR 8052 - Centre Lane Trading Ltd.; Rate Filing Includes Request for Blanket Section 204 Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Centre Lane Trading Ltd.; Rate Filing Includes Request for Blanket Section... of Centre Lane Trading Ltd.'s application for market-based rate authority, with an accompanying...

  17. 75 FR 31430 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-03

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization May 27, 2010. Take notice that on May 24, 2010, Transcontinental Gas Pipe Line Company, LLC (Transco), Post Office Box...

  18. 78 FR 38712 - Ryckman Creek Resources, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-27

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Ryckman Creek Resources, LLC; Notice of Request Under Blanket Authorization Take notice that on June 11, 2013, Ryckman Creek Resources, LLC (Ryckman), 3 Riverway, Suite 1110, Houston, TX 77056, filed in Docket No....

  19. Study of the effects of corrugated wall structures due to blanket modules around ICRH antennas

    NASA Astrophysics Data System (ADS)

    Dumortier, Pierre; Louche, Fabrice; Messiaen, André; Vervier, Michel

    2014-02-01

    In future fusion reactors, and in ITER, the first wall will be covered by blanket modules. These blanket modules, whose dimensions are of the order of the ICRF wavelengths, together with the clearance gaps between them will constitute a corrugated structure which will interact with the electromagnetic waves launched by ICRF antennas. The conditions in which the grooves constituted by the clearance gaps between the blanket modules can become resonant are studied. Simple analytical models and numerical simulations show that mushroom type structures (with larger gaps at the back than at the front) can bring down the resonance frequencies, which could lead to large voltages in the gaps between the blanket modules and perturb the RF properties of the antenna if they are in the ICRF operating range. The effect on the wave propagation along the wall structure, which is acting as a spatially periodic (toroidally and poloidally) corrugated structure, and hence constitutes a slow wave structure modifying the wall boundary condition, is examined.

  20. 75 FR 21290 - Caledonia Energy Partners, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-04-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Caledonia Energy Partners, L.L.C.; Notice of Request Under Blanket Authorization April 16, 2010. Take notice that on April 12, 2010, Caledonia Energy Partners, L.L.C....

  1. 78 FR 44558 - Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-24

    ... Energy Regulatory Commission Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on July 3, 2013, Stingray Pipeline Company, L.L.C. (Stingray), 1100 Louisiana... directed to Cynthia Hornstein Roney, Manager, Regulatory Compliance, Stingray Pipeline Company,...

  2. 77 FR 48149 - Columbia Gas Transmission, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-13

    ... Energy Regulatory Commission Columbia Gas Transmission, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on July 24, 2012 Columbia Gas Transmission, L.L.C. (Columbia), P.O. Box 1273... directed to Fredric J. George, Senior Counsel, Columbia Gas Transmission, L.L.C., P.O. Box 1273,...

  3. 76 FR 4651 - Venice Gathering System, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-26

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Venice Gathering System, L.L.C.; Notice of Request Under Blanket Authorization January 19, 2010. Take notice that on January 7, 2011, Venice Gathering System, L.L.C....

  4. 75 FR 62533 - Destin Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-12

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Destin Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization October 1, 2010. Take notice that on September 20, 2010, Destin Pipeline Company, L.L.C....

  5. 77 FR 42302 - Texas Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-18

    ... Energy Regulatory Commission Texas Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on July 2, 2012 Texas Gas Transmission, LLC (Texas Gas), 3800 Frederica Street... this application should be directed to Kathy D. Fort, Manager of Certificates and Tariffs, Texas...

  6. 75 FR 74713 - Panhandle Eastern Pipe Line Company, LP; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-01

    ... Energy Regulatory Commission Panhandle Eastern Pipe Line Company, LP; Notice of Request Under Blanket Authorization November 23, 2010. Take notice that on November 12, 2010, Panhandle Eastern Pipe Line Company, LP..., Missouri. Panhandle states that a portion of the pipe underlying Boller Lane will be grouted while...

  7. 77 FR 39699 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-05

    ... Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization Take notice that on June 13, 2012, Transcontinental Gas Pipe Line Company, LLC (Transco), Post... regarding this Application should be directed to Nan Miksovsky, Transcontinental Gas Pipe Line Company,...

  8. 75 FR 33803 - Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-15

    ... Energy Regulatory Commission Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization June 8, 2010. Take notice that on June 1, 2010, Sabine Pipe Line LLC (Sabine), 4800 Fournace Place, Bellaire... L. Kirk, Regulatory Specialist, Chevron Pipe Line Company, 4800 Fournace Place, Bellaire,...

  9. 76 FR 44903 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-27

    ... Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization Take notice that on June 30, 2011, Transcontinental Gas Pipe Line Company, LLC (Transco), Post... this application should be directed to Nan Miksovsky, Transcontinental Gas Pipe Line Company, LLC,...

  10. 77 FR 33213 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-05

    ... Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization Take notice that on May 17, 2012 Transcontinental Gas Pipe Line Company, LLC (Transco), Post... directed to Bela Patel, Transcontinental Gas Pipe Line Company, LLC, P.O. Box 1396, Houston, Texas...

  11. 75 FR 66751 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-29

    ... Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization October 22, 2010. Take notice that on October 14, 2010, Transcontinental Gas Pipe Line Company... directed to Nan Miksovsky, Transcontinental Gas Pipe Line Company, LLC, Post Office Box 1396,...

  12. 76 FR 29745 - Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-23

    ... Energy Regulatory Commission Transcontinental Gas Pipe Line Company, LLC; Notice of Request Under Blanket Authorization Take notice that on May 5, 2011 Transcontinental Gas Pipe Line Company, LLC (Transco), Post Office... prior notice should be directed to Nan Miksovsky, Transcontinental Gas Pipe Line Company, LLC, P.O....

  13. 75 FR 45111 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-02

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization July 26, 2010. Take notice that on July 20, 2010, Kinder Morgan Interstate Gas Transmission LLC (KMIGT), PO Box...

  14. Security Blanket or Crutch? Crib Card Usage Depends on Students' Abilities

    ERIC Educational Resources Information Center

    Burns, Kathleen C.

    2014-01-01

    This study investigated whether students use crib cards as a security blanket or a crutch by asking students to tally the number of times they used them during exams in a statistics class. There was a negative correlation between the number of times students used their crib cards and exam performance. High-achieving students did not utilize their…

  15. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY OTHER REGULATIONS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES CERTAIN SALES AND TRANSPORTATION OF NATURAL GAS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES Blanket Certificates Authorizing Certain Natural...

  16. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-01-20

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization January 8, 2010. Take notice that on December 30, 2009, Northern Natural Gas Company (Northern), 1111... sections 157.205 and 157.214 of the Commission's regulations under the Natural Gas Act for authorization...

  17. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-24

    ... Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite....210 of the Federal Energy Regulatory Commission's regulations under the Natural Gas Act (NGA),...

  18. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at...

  19. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY OTHER REGULATIONS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES CERTAIN SALES AND TRANSPORTATION OF NATURAL GAS UNDER THE NATURAL GAS POLICY ACT OF 1978 AND RELATED AUTHORITIES Blanket Certificates Authorizing Certain Natural...

  20. 76 FR 18216 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-01

    ... Federal Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on March 16, 2011, Southern Natural Gas Company (Southern), Post Office Box 2563... and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA) as amended, to abandon...

  1. 78 FR 63974 - Enable Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... Energy Regulatory Commission Enable Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate On October 11, 2013, Enable Gas Transmission, LLC (Enable) filed with the Federal Energy..., Enable Gas Transmission, LLC, P.O. Box 21734, Shreveport, Louisiana 71151 or by calling...

  2. 76 FR 34072 - Midwestern Gas Transmission Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-10

    ... Energy Regulatory Commission Midwestern Gas Transmission Company; Notice of Request Under Blanket Authorization Take notice that on May 19, 2011, Midwestern Gas Transmission Company (Midwestern) filed a prior... directed to Joseph Miller, Midwestern Gas Transmission Company, 100 West 5th Street, ONEOK Plaza,...

  3. 75 FR 33299 - Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-11

    ... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization June 4, 2010. Take notice that on May 28, 2010, Florida Gas Transmission Company, LLC (FGT), 5444... should be directed to Stephen Veatch, Senior Director of Certificates & Tariffs, Florida Gas...

  4. 75 FR 33298 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-11

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization June 4, 2010. Take notice that on May 20, 2010 Columbia Gas Transmission, LLC (Columbia), 5151 San... directed to Fredic J. George, Senior Counsel, Columbia Gas Transmission, LLC, P.O. Box 1273,...

  5. 78 FR 44111 - Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-23

    ... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization Take notice that on July 1, 2013, Florida Gas Transmission Company, LLC (Florida Gas), 1300 Main... directed Stephen Veatch, Senior Director of Certificates & Tariffs, Florida Gas Transmission Company,...

  6. 78 FR 62015 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-11

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on September 25, 2013, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe..., Senior Counsel, Columbia Gas Transmission, LLC, P.O. Box 1273, Charleston, West Virginia 25325-1273,...

  7. 78 FR 53742 - Columbia Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-30

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate On August 14, 2013, Columbia Gas Transmission, LLC (Columbia) filed with the Federal Energy... application may be directed to Fredric J. George, Senior Counsel, Columbia Gas Transmission, LLC, P.O....

  8. 76 FR 2371 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-13

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization January 5, 2011. Take notice that on December 22, 2010, Columbia Gas Transmission, LLC (Columbia... J. George, Senior Counsel, Columbia Gas Transmission, LLC, P.O. Box 1273, Charleston, West...

  9. 75 FR 26224 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-11

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization May 4, 2010. Take notice that on April 22, 2010, Columbia Gas Transmission, LLC (Columbia), 5151..., Regulatory Affairs, Columbia Gas Transmission, LLC, 5151 San Felipe, Suite 2500, Houston, Texas 77056, or...

  10. 76 FR 1429 - Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-10

    ... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Request Under Blanket Authorization December 29, 2010. Take notice that on December 16, 2010 Florida Gas Transmission Company, LLC... & Tariffs, Florida Gas Transmission Company, LLC, 5444 Westheimer Road, Houston, Texas 77056, or call...

  11. 78 FR 26772 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-08

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on April 24, 2013, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe..., and 157.216(b) of the Commission's regulations under the Natural Gas Act (NGA) for authorization...

  12. 78 FR 58303 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-23

    ... Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on September 6, 2013, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe... sections 157.205 and 157.216(b) of the Commission's Regulations under the Natural Gas Act (NGA) as...

  13. 78 FR 3893 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-17

    ... Federal Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on January 7, 2013, Columbia Gas Transmission, LLC (Columbia) 5151 San Felipe... directed to Fredric J. George, Senior Counsel, Columbia Gas Transmission, LLC, P.O. Box 1273,...

  14. 77 FR 36532 - Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-19

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on June 1, 2012, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe... Columbia Gas Transmission, LLC, P.O. Box 1273, Charleston, West Virginia 25325-1273, phone (304)...

  15. 76 FR 65720 - Kern River Gas Transmission Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-24

    ... Energy Regulatory Commission Kern River Gas Transmission Company; Notice of Request Under Blanket Authorization Take notice that on October 6, 2011, Kern River Gas Transmission Company (Kern River) filed a..., Kern River Gas Transmission Company, 1111 South 103 Street, Omaha, Nebraska 68124, at (402)...

  16. 75 FR 52519 - Columbia Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-26

    ... Energy Regulatory Commission Columbia Gas Transmission, LLC; Prior Notice of Activity Under Blanket Certificate August 19, 2010. On August 9, 2010, Columbia Gas Transmission, LLC (Columbia) filed with the... Fredric J. George, Senior Counsel, Columbia Gas Transmission, LLC, P.O. Box 1273, Charleston,...

  17. 76 FR 58263 - Notice of Request Under Blanket Authorization; Columbia Gas Transmission, LLC

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-20

    ... Energy Regulatory Commission Notice of Request Under Blanket Authorization; Columbia Gas Transmission, LLC Take notice that on July 26, 2011, Columbia Gas Transmission, LLC (Columbia), 5151 San Felipe... application should be directed to Fredric J. George, Senior Counsel, Columbia Gas Transmission Corporation,...

  18. 75 FR 35019 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-21

    ... Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization June 11, 2010. Take notice that on June 3, 2009, Kinder Morgan Interstate Gas..., Kinder Morgan Interstate Gas Transmission LLC, P.O. Box 281304, Lakewood, Colorado 80228-8304, or...

  19. 78 FR 48669 - Carolina Gas Transmission Corporation; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-09

    ... Energy Regulatory Commission Carolina Gas Transmission Corporation; Notice of Request Under Blanket Authorization Take notice that on July 19, 2013 Carolina Gas Transmission Corporation (CGT), 601 Old Taylor Road... application should be directed to Michael R. Ferguson, Manager-System Intergity, Carolina Gas...

  20. Upflow anaerobic sludge blanket reduces COD 75-85%, produces methane gas

    SciTech Connect

    Ruppel, W.; Biedron, M.; Thornton, B.; Swientek, R.J.

    1982-01-01

    The wastewater from a brewery at 3 million gallons/day is treated in an upflow anaerobic sludge blanket process with a COD removal efficiency of 75% and the CH/sub 4/ gas content of the 400 cubic metres/day biogas produced 74%.

  1. 78 FR 79691 - Trunkline Gas Company, LLC; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-31

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Trunkline Gas Company, LLC; Notice of Request Under Blanket Authorization Take notice that on December 13, 2013, Trunkline Gas Company, LLC (Trunkline), PO Box 4967, Houston, Texas 77210-4967, filed in Docket No....

  2. A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

    SciTech Connect

    Martinez-Frances, N.; Timm, W.; Rossbach, D.

    2012-07-01

    Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main design criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)

  3. Thermal hydraulic study of the ESPRESSO blanket for a Tandem Mirror Reactor

    SciTech Connect

    Raffray, A.R.; Hoffman, M.A.

    1986-02-01

    This paper deals primarily with the thermal-hydraulic design and some critical thermomechanical aspects of the proposed ESPRESSO blanket for the Tandem Mirror Fusion Reactor. This conceptual design was based on the same physics as used in the MARS study.

  4. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 2 2010-07-01 2010-07-01 false DeCA Blanket Routine Uses C Appendix C to Part...) PRIVACY PROGRAM DEFENSE COMMISSARY AGENCY PRIVACY ACT PROGRAM Pt. 327, App. C Appendix C to Part 327—DeCA... letting of a contract, or the issuance of a license, grant, or other benefit. (c) Routine...

  5. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 3 2013-07-01 2013-07-01 false Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  6. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 3 2012-07-01 2009-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  7. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 3 2014-07-01 2014-07-01 false Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  8. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  9. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 3 2011-07-01 2009-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

  10. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    NASA Astrophysics Data System (ADS)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  11. Swiss fusion blanket experiments: Final report, November 1, 1985-October 31, 1987

    SciTech Connect

    Woodruff, G.L.

    1987-12-13

    The major thrust of this project related to the effort to transfer the Lithium Blanket Module (LBM) to the Nuclear Engineering Laboratory of the Swiss Institute of Technology at Lausanne, and to the subsequent support with analytical calculations of a variety of experiments performed with the LBM. 12 refs.

  12. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 4 2011-10-01 2011-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES CONTRACTING METHODS AND CONTRACT TYPES SIMPLIFIED ACQUISITION PROCEDURES Simplified Acquisition Methods 313.303-5 Purchases under...

  13. Combined glyphosate-ripener and residue blanket stresses reduce ratoon yields in Louisiana

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Failure to remove the blanket of residue generated during green-cane harvesting and certain glyphosate ripener application regimes have independently been shown to reduce yields of the subsequent ratoon crop of Louisiana’s leading variety LCP 85-384. The objectives of this experiment were to determ...

  14. 76 FR 23808 - Colorado Interstate Gas Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-28

    ..., Regulatory Affairs Department, Colorado Interstate Gas Company, Post Office Box 1087, Colorado Springs... Energy Regulatory Commission Colorado Interstate Gas Company; Notice of Request Under Blanket Authorization Take notice that on April 7, 2011, Colorado Interstate Gas Company (CIG) filed a prior...

  15. European ceramic B.I.T. blanket for DEMO: Recent development for the zirconate version

    SciTech Connect

    Bielak, B.; Eid, M.; Fuetterer, M.

    1994-12-31

    Within the framework of the European test-blanket program, CEA and ENEA are jointly developing a DEMO-relevant, helium-cooled, Breeder-Inside-Tube (BIT) ceramic blanket. Two ceramics are possible breeder material candidate: LiAlO{sub 2} and Li{sub 2}ZrO{sub 3}. Despite the design has been originally developed for aluminate, the CEA has recently focused its work on zirconate. This concept blanket segments are made by a directly-cooled vacuum-tight steel box which contains banana-shaped poloidal breeder modules arranged in rows (6 rows in an outboard segment and 4 rows in an inboard one). A breeder module consists of a pressure vessel containing a bundle of breeder rods surrounded by baffles. Each one of the rods is made-up of a steel tube containing a stack of annular pellets of sintered lithium-zirconate, through which flows helium (the tritium purge gas). The thermo-mechanical analysis has shown that the thermal gradient in the ceramics can be kept at acceptable levels despite the poorer out-of-pile thermo-mechanical properties of zirconate compared to aluminate. Moreover, the neutronic analysis has shown that, besides the maintained tritium-breeding self-sufficiency capability of this blanket, the lower lithium burn-up could be an indication that the zirconate characteristics remains more stable after long term irradiation (i.e., close to the end-of-life fluence of 5 MWa/m{sup 2}).

  16. 75 FR 39681 - Tennessee Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-12

    ... Energy Regulatory Commission Tennessee Pipeline Company; Notice of Request Under Blanket Authorization July 1, 2010. Take notice that on June 18, 2010, Tennessee Pipeline Company (Tennessee), 1001 Louisiana... TTY, (202) 502-8659. Specifically, Tennessee proposes to abandon an inactive offshore supply...

  17. 76 FR 45253 - Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-28

    ... Energy Regulatory Commission Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization July 20, 2011. Take notice that on July 11, 2011, Tennessee Gas Pipeline Company (Tennessee Gas... up to the higher MAOP. Specifically, Tennessee Gas proposes to increase the MAOP of Line 2B-100...

  18. 76 FR 60016 - Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-28

    ... Energy Regulatory Commission Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization Take notice that on September 9, 2011, Tennessee Gas Pipeline Company (Tennessee), 1001 Louisiana... Tennessee's authorization in Docket No. CP82-413-000, to abandon in place and by removal an inactive...

  19. 75 FR 5317 - Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-02

    ... Federal Energy Regulatory Commission Tennessee Gas Pipeline Company; Notice of Request Under Blanket Authorization January 26, 2010. Take notice that on January 25, 2010, Tennessee Gas Pipeline Company (Tennessee... TTY, (202) 502-8659. Specifically, Tennessee proposes to abandon in place Line 509A-3600...

  20. RELAP5 Model of the First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

    2010-06-01

    ITER inductive power operation is modeled and simulated using a system level computer code to evaluate the behavior of the Primary Heat Transfer System (PHTS) and predict parameter operational ranges. The control algorithm strategy and derivation are summarized in this report as well. A major feature of ITER is pulsed operation. The plasma does not burn continuously, but the power is pulsed with large periods of zero power between pulses. This feature requires active temperature control to maintain a constant blanket inlet temperature and requires accommodation of coolant thermal expansion during the pulse. In view of the transient nature of the power (plasma) operation state a transient system thermal-hydraulics code was selected: RELAP5. The code has a well-documented history for nuclear reactor transient analyses, it has been benchmarked against numerous experiments, and a large user database of commonly accepted modeling practices exists. The process of heat deposition and transfer in the blanket modules is multi-dimensional and cannot be accurately captured by a one-dimensional code such as RELAP5. To resolve this, a separate CFD calculation of blanket thermal power evolution was performed using the 3-D SC/Tetra thermofluid code. A 1D-3D co-simulation more realistically models FW/blanket internal time-dependent thermal inertia while eliminating uncertainties in the time constant assumed in a 1-D system code. Blanket water outlet temperature and heat release histories for any given ITER pulse operation scenario are calculated. These results provide the basis for developing time dependent power forcing functions which are used as input in the RELAP5 calculations.

  1. Pre-clinic study of uniformity of light blanket for intraoperative photodynamic therapy

    NASA Astrophysics Data System (ADS)

    Hu, Yida; Wang, Ken; Zhu, Timothy C.

    2010-02-01

    A large-size blanket composed of the parallel catheters and silica core side glowing fiber is designed to substitute the hand-held point source in the photodynamic therapy treatment (PDT) of the malignant pleural or intraperitoneal diseases. It produces a reasonably uniform field for effective light coverage and is flexible to conform to anatomic structures in intraoperative PDT. The size of the blanket is 30cm×20cm. The light blanket composed of several PVC layers and a series of parallel catheters attached on both sides of the intralipid layer of 0.2% concentration. On one side of the intralipid layer, the parallel fiber catheters were attached using thermal sealing technique. On the other side, the parallel detect catheters were attached along the perpendicular direction. 0.1mm aluminum foil was used to construct the reflection layer to enhance the efficiency of light delivery. The long single side-glowing fiber goes through the fiber catheters according to the specific fiber pattern design. Compared with the prototype of the first generation, the new design is more cost-efficient and more applicable for clinical applications. The light distribution of the blanket was characterized by scanning experiments which were performed in flatness and on the curved surface of tissue body phantom. The fluence rate generated by the blanket can meet requirements for the light delivery in pleural or intraperitoneal (IP) PDT. Taking the advantage of large coverage and flexible conformity, it has great value to increase the reliability and consistency of PDT.

  2. Tritium self-sufficiency time and inventory evolution for solid-type breeding blanket materials for DEMO

    NASA Astrophysics Data System (ADS)

    Packer, L. W.; Pampin, R.; Zheng, S.

    2011-10-01

    One of the primary functions of a fusion blanket is to generate enough tritium to make a fusion power plant (FPP) self-sufficient. To ensure that there is satisfactory tritium production in a real plant the tritium breeding ratio (TBR) in the blanket must be greater than 1 + M, where M is the breeding margin. For solid-type blanket designs, the initial TBR must be significantly higher than 1 + M, since the blanket TBR will be reduced over time as the lithium fuel is consumed. The rate of TBR reduction will impact on the overall blanket self-sufficiency time, the time in which the net tritium inventory of the system is positive. DEMO relevant blanket materials, Li 4SiO 4 and Li 2TiO 3, are investigated by computational simulation using radiation transport tools coupled with time-dependent inventory calculations. The results include tritium inventory assessments and depletion of breeding materials over time, which enable self-sufficiency times and maximum surplus tritium inventories to be evaluated, which are essential quantities to determine to allow one to design a credible FPP using solid-type breeding material concepts. The blanket concepts investigated show self-sufficiency times of several years in some cases and maximum surplus inventories of up to a few tens of kg.

  3. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  4. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    NASA Astrophysics Data System (ADS)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  5. Non-LTE, line-blanketed model atmospheres for late O- and early B-type stars

    NASA Technical Reports Server (NTRS)

    Grigsby, James A.; Morrison, Nancy D.; Anderson, Lawrence S.

    1992-01-01

    The use of non-LTE line-blanketed model atmospheres to analyze the spectra of hot stars is reported. The stars analyzed are members of clusters and associations, have spectral types in the range O9-B2 and luminosity classes in the range III-IV, have slow to moderate rotation, and are photometrically constant. Sampled line opacities of iron-group elements were incorporated in the radiative transfer solution; solar abundances were assumed. Good to excellent agreement is obtained between the computed profiles and essentially all the line profiles used to fix the model, and reliable stellar parameters are derived. The synthetic M II 5581 equivalent widths agree well with the observed ones at the low end of the temperature range studied, but, above 25,000 K, the synthetic line is generally stronger than the observed line. The behavior of the observed equivalent widths of N II, N III, C II and C III lines as a function of Teff is studied. Most of the lines show much scatter, with no consistent trend that could indicate abundance differences from star to star.

  6. It's elemental

    NASA Astrophysics Data System (ADS)

    The Periodic Table of the elements will now have to be updated. An international team of researchers has added element 110 to the Earth's armory of elements. Though short-lived—of the order of microseconds, element 110 bottoms out the list as the heaviest known element on the planet. Scientists at the Heavy Ion Research Center in Darmstadt, Germany, made the 110-proton element by colliding a lead isotope with nickel atoms. The element, which is yet to be named, has an atomic mass of 269.

  7. Exploring climatic controls on blanket bog litter decomposition across an altitudinal gradient

    NASA Astrophysics Data System (ADS)

    Bell, Michael; Ritson, Jonathan P.; Clark, Joanna M.; Verhoef, Anne; Brazier, Richard E.

    2016-04-01

    The hydrological and ecological functioning of blanket bogs is strongly coupled, involving multiple ecohydrological feedbacks which can affect carbon cycling. Cool and wet conditions inhibit decomposition, and favour the growth of Sphagnum mosses which produce highly recalcitrant litter. A small but persistent imbalance between production and decomposition has led to blanket bogs in the UK accumulating large amounts of carbon. Additionally, healthy bogs provide a suite of other ecosystems services including water regulation and drinking water provision. However, there is concern that climate change could increase rates of litter decomposition and disrupt this carbon sink. Furthermore, it has been argued that the response of these ecosystems in the warmer south west and west of the UK may provide an early analogue for later changes in the more extensive northern peatlands. In order to investigate the effects of climate change on blanket bog litter decomposition, we set-up a litter bag experiment across an altitudinal gradient spanning 200 m of elevation (including a transition from moorland to healthy blanket bog) on Dartmoor, an area of hitherto unstudied, climatically marginal blanket bog in the south west of the UK. At seven sites, water table depth and soil and surface temperature were recorded continuously. Litter bags filled with the litter of three vegetation species dominant on Dartmoor were incubated just below the bog surface and retrieved over a period of 12 months. We found significant differences in the rate of decomposition between species. At all sites, decomposition progressed in the order Calluna vulgaris (dwarf shrub) > Molinia caerulea (graminoid) > Sphagnum (bryophyte). However, while soil temperature did decrease along the altitudinal gradient, being warmer in the lower altitudes, a hypothesised accompanying decrease in decomposition rates did not occur. This could be explained by greater N deposition at the higher elevation sites (estimated

  8. 76 FR 57731 - Supplemental Notice That Initial Market-Based Rate Filing Includes Request for Blanket Section...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-16

    ... Request for Blanket Section 204 Authorization; Rockland Wind Farm, LLC This is a supplemental notice in the above-referenced proceeding of Rockland Wind Farm, LLC's application for market-based...

  9. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    SciTech Connect

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  10. Rapid thermal cycling of solar array blanket coupons for Space Station Freedom

    NASA Technical Reports Server (NTRS)

    Scheiman, David A.; Smith, Bryan K.

    1991-01-01

    The NASA Lewis Research Center has been conducting rapid thermal cycling on blanket coupons for Space Station Freedom. This testing includes two designs (8 coupons total) of the solar array. Four coupons were fabricated as part of the Photovoltaic Array Environmental Protection Program (PAEP), NAS3-25079, at Lockheed Missiles and Space Company. These coupons began cycling in early 1989 and have completed 172,000 thermal cycles. Four other coupons were fabricated a year later and included several design changes; cycling of these began in early 1990 and has reached 90,000 cycles. The objective of this testing is to demonstrate the durability or operational lifetime (15 yrs.) of the welded interconnects within a low earth orbit (LEO) thermal cycling environment. The blanket coupons, design changes, test description, status to date including performance and observed anomalies, and any insights related to the testing of these coupons are described. The description of a third design is included.

  11. Development of welding technologies for the manufacturing of European Tritium Breeder blanket modules

    NASA Astrophysics Data System (ADS)

    Poitevin, Y.; Aubert, Ph.; Diegele, E.; de Dinechin, G.; Rey, J.; Rieth, M.; Rigal, E.; von der Weth, A.; Boutard, J.-L.; Tavassoli, F.

    2011-10-01

    Europe has developed two reference Tritium Breeder Blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both are using the reduced-activation ferritic-martensitic EUROFER-97 steel as structural material and will be tested in ITER under the form of test blanket modules. The fabrication of their EUROFER structures requires developing welding processes like laser, TIG, EB and diffusion welding often beyond the state-of-the-art. The status of European achievements in this area is reviewed, illustrating the variety of processes and key issues behind retained options, in particular with respect to metallurgical aspects and mechanical properties. Fabrication of mock-ups is highlighted and their characterization and performances with respect to design requirements are reviewed.

  12. Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

    NASA Astrophysics Data System (ADS)

    Fradera, J.; Sedano, L.; Mas de les Valls, E.; Batet, L.

    2011-10-01

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phenomena in liquid Pb17.5Li and EUROFER is proposed. In a first approximation, it has been considered that He bubbles can be represented as a passive scalar. The nucleation model is based on the classical theory and includes a simplified bubble growth model. The model captures the interaction of tritium with bubbles and tritium diffusion through walls. Results show the influence of helium cavitation on tritium inventory and the importance of simulating the system walls instead of imposing fixed boundary conditions.

  13. Numerical Simulation of MHD Effect in Liquid Metal Blankets with Flow Channel Insert

    NASA Astrophysics Data System (ADS)

    Mao, J.; Pan, H. C.

    2011-09-01

    The magnetohydrodynamic effect in liquid metal blankets with flow channel insert and pressure equalization slot for fusion liquid metal blanket is studied by numerical simulation based on two dimensional fully developed flow model. The code is verified by comparing analytical solution and numerical solution of Hunt Case II. The velocity field and MHD pressure drop varying with electric conductivity of the FCI is analyzed. The result shows that the average velocity in central area of the cross section decreases with the increase of the electric conductivity of FCI. While the average velocity in gap zone is reverse. Comparing with MHD duct flow without FCI, MHD pressure drop is reduced significantly when the FCI material is electrically insulating.

  14. A model for wind-extension of the Copernicus ejecta blanket

    NASA Technical Reports Server (NTRS)

    Rehfuss, D. E.; Michael, D.; Anselmo, J. C.; Kincheloe, N. K.

    1977-01-01

    The interaction between crater ejecta and the transient wind from impact-shock vaporization is discussed. Based partly on Shoemaker's (1962) ballistic model of the Copernicus ejecta and partly on Rehfuss' (1972) treatment of lunar winds, a simple model is developed which indicates that if Copernicus were formed by a basaltic meteorite impacting at 20 km/s, then 3% of the ejecta mass would be sent beyond the maximum range expected from purely ballistic trajectories. That 3% mass would, however, shift the position of the outer edge of the ejecta blanket more than 400% beyond the edge of the ballistic blanket. For planetary bodies lacking an intrinsic atmosphere, the present model indicates that this form of hyperballistic transport can be very significant for small (no more than about 1 kg) ejecta fragments.

  15. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    NASA Astrophysics Data System (ADS)

    Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

    2013-07-01

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ˜14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

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

    DOEpatents

    Peterson, Per F.

    2013-05-14

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

  17. Efficient COD removal and nitrification in an upflow microaerobic sludge blanket reactor for domestic wastewater.

    PubMed

    Zheng, Shaokui; Cui, Cancan

    2012-03-01

    The treatment performance of an upflow microaerobic sludge blanket reactor (UMSB) for synthetic domestic wastewater was investigated at two dissolved oxygen (DO) levels, 0.3-0.5 and 0.7-0.9 mg l(-1), focusing on nitrification performance. The higher DO level induced complete nitrification of ammonia nitrogen (NH(3)-N), achieving chemical oxygen demand and NH(3)-N removals of 97 and 92%, respectively. There were consistently significantly higher nitrate nitrogen (NO(3)-N) and nitrite nitrogen (NO(2)-N) levels in the effluent, with ~66% of newly-produced oxidised nitrogen as NO(2)-N. Despite the high nitrification efficiency, only about 23% of the removed NH(3)-N amount from the influent was ultimately transformed into oxidised nitrogen due to the simultaneous nitrification-denitrification. Sludge blanket development and granulation occurred simultaneously in the UMSB. PMID:22105554

  18. Evaluating the impacts of re-vegetation of bare peat on blanket peat water tables

    NASA Astrophysics Data System (ADS)

    Shuttleworth, Emma; Richards, Rebecca; Evans, Martin; Agnew, Clive; Pilkington, Mike; Maskill, Rachael; Allott, Tim

    2015-04-01

    Studies of the hydrological impacts of peat restoration in blanket peat systems have focused on the impacts of drain and gully blocking on water tables. However, in the South Pennines of the UK large areas of previously bare blanket peat have been restored by re-vegetation. The effects of this restoration treatment on water table behaviour have not been fully evaluated. Preliminary data from space-for-time studies indicate that re-vegetation leads to significant rises in water tables and decreases in water table variability. Here we present additional data from a before-after-control-intervention (BACI) study to validate these preliminary observations. We also present meteorological, net radiation and evapotranspiration data to test the hypothesis that water table changes associated with re-vegetation are driven by changing evapotranspiration rates as bare peat surfaces re-vegetate. The wider ecosystem service benefits of water table increases associated with re-vegetation of bare peat are discussed.

  19. Reduced activation martensitic steels as a structural material for ITER test blanket

    NASA Astrophysics Data System (ADS)

    Shiba, K.; Enoeda, M.; Jitsukawa, S.

    2004-08-01

    A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production.

  20. Thermal cycle testing of Space Station Freedom solar array blanket coupons

    NASA Technical Reports Server (NTRS)

    Scheiman, David A.; Schieman, David A.

    1991-01-01

    Lewis Research Center is presently conducting thermal cycle testing of solar array blanket coupons that represent the baseline design for Space Station Freedom. Four coupons were fabricated as part of the Photovoltaic Array Environment Protection (PAEP) Program, NAS 3-25079, at Lockheed Missile and Space Company. The objective of the testing is to demonstrate the durability or operational lifetime of the solar array welded interconnect design within the durability or operational lifetime of the solar array welded interconnect design within a low earth orbit (LEO) thermal cycling environment. Secondary objectives include the observation and identification of potential failure modes and effects that may occur within the solar array blanket coupons as a result of thermal cycling. The objectives, test articles, test chamber, performance evaluation, test requirements, and test results are presented for the successful completion of 60,000 thermal cycles.

  1. Electromagnetic Launch Vehicle Fairing and Acoustic Blanket Model of Received Power Using FEKO

    NASA Technical Reports Server (NTRS)

    Trout, Dawn H.; Stanley, James E.; Wahid, Parveen F.

    2011-01-01

    Evaluating the impact of radio frequency transmission in vehicle fairings is important to electromagnetically sensitive spacecraft. This study employs the multilevel fast multipole method (MLFMM) from a commercial electromagnetic tool, FEKO, to model the fairing electromagnetic environment in the presence of an internal transmitter with improved accuracy over industry applied techniques. This fairing model includes material properties representative of acoustic blanketing commonly used in vehicles. Equivalent surface material models within FEKO were successfully applied to simulate the test case. Finally, a simplified model is presented using Nicholson Ross Weir derived blanket material properties. These properties are implemented with the coated metal option to reduce the model to one layer within the accuracy of the original three layer simulation.

  2. Preliminary lifetime predictions for 304 stainless steel as the LANL ABC blanket material

    SciTech Connect

    Park, J.J.; Buksa, J.J.; Houts, M.G.; Arthur, E.D.

    1997-11-01

    The prediction of materials lifetime in the preconceptual Los Alamos National Laboratory (LANL) Accelerator-Based Conversion of Plutonium (ABC) is of utmost interest. Because Hastelloy N showed good corrosion resistance to the Oak Ridge National Laboratory Molten Salt Reactor Experiment fuel salt that is similar to the LANL ABC fuel salt, Hastelloy N was originally proposed for the LANL ABC blanket material. In this paper, the possibility of using 304 stainless steel as a replacement for the Hastelloy N is investigated in terms of corrosion issues and fluence-limit considerations. An attempt is made, based on the previous Fast Flux Test Facility design data, to predict the preliminary lifetime estimate of the 304 stainless steel used in the blanket region of the LANL ABC.

  3. Radiation effects measurements on spacecraft electrostatic discharge tapes, thermal blankets and thermooptical coatings

    NASA Technical Reports Server (NTRS)

    Bouquet, F. L.; Hribar, V. F.; Metzler, E. C.; Russell, D. A.

    1984-01-01

    Selective results are presented of laboratory radiation tests of metallic foil tapes, thermal blankets, and thermooptical coatings undertaken as part of the development and qualification of materials for the Galileo spacecraft. Of the two metallic foil tapes used for electrical continuity, the adhesive used on the aluminum embossed foil was superior to the copper embossed foil when exposed to simulated Jovian electrons. Proton-irradiation tests performed on a number of thermal blanket samples showed that black polyester on Kapton proved to be a lower weight loss (i.e., outgassing) material than Fluorglas. In addition, preliminary results concerning the response of thermooptical coatings to simulated Jovian electrons show that the ITO-coated polyester over a Kapton surface gave the lowest absorptance.

  4. Neutron dosimetry qualification experiments for the Tokamak Fusion Test Reactor Lithium Blanket Module program

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderi, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment and the use of such data in neutron spectrum adjustment and in predicting integral performance parameters, e.g., tritium production.

  5. Prostate cancer in relation to the use of electric blanket or heated water bed.

    PubMed

    Zhu, K; Weiss, N S; Stanford, J L; Daling, J R; Stergachis, A; McKnight, B; Brawer, M K; Levine, R S

    1999-01-01

    Using data from a case-control study conducted in Group Health Cooperative (GHC) of Puget Sound, we examined the relation between the use of electric blankets or heated water beds and the risk of prostate cancer. Cases were 175 prostate cancer patients ages 40-69 years. Controls were 258 male GHC members frequency matched to cases. The odds ratio (OR) for prostate cancer associated with the use of an electric blanket or heated water bed was 1.4 (95% confidence interval (CI) 0.9-2.2). The risk, however, did not tend to be higher with increasing months per year or years of use. This study did not provide clear evidence on the hypothesized association. PMID:9888286

  6. A table of semiempirical gf values. Part 1: Wavelengths: 5.2682 nm to 272.3380 nm. [to calculate line-blanketed model atmospheres for solar and stellar spectra

    NASA Technical Reports Server (NTRS)

    Kurucz, R. L.; Peytremann, E.

    1975-01-01

    The gf values for 265,587 atomic lines selected from the line data used to calculate line-blanketed model atmospheres are tabulated. These data are especially useful for line identification and spectral synthesis in solar and stellar spectra. The gf values are calculated semiempirically by using scaled Thomas-Fermi-Dirac radial wavefunctions and eigenvectors found through least-squares fits to observed energy levels. Included in the calculation are the first five or six stages of ionization for sequences up through nickel. Published gf values are included for elements heavier than nickel. The tabulation is restricted to lines with wavelengths less than 10 micrometers.

  7. Lightweight solar array blanket tooling, laser welding and cover process technology

    NASA Technical Reports Server (NTRS)

    Dillard, P. A.

    1983-01-01

    A two phase technology investigation was performed to demonstrate effective methods for integrating 50 micrometer thin solar cells into ultralightweight module designs. During the first phase, innovative tooling was developed which allows lightweight blankets to be fabricated in a manufacturing environment with acceptable yields. During the second phase, the tooling was improved and the feasibility of laser processing of lightweight arrays was confirmed. The development of the cell/interconnect registration tool and interconnect bonding by laser welding is described.

  8. Space Photovoltaic Research and Technology 1983. High Efficiency, Radiation Damage, and Blanket Technology

    NASA Technical Reports Server (NTRS)

    1984-01-01

    This three day conference, sixth in a series that began in 1974, was held at the NASA Lewis Research Center on October 18-20, 1983. The conference provided a forum for the discussion of space photovoltaic systems, their research status, and program goals. Papers were presented and workshops were held in a variety of technology areas, including basic cell research, advanced blanket technology, and radiation damage.

  9. Palladium-catalyzed oxidative diffusion for tritium extraction from breeder-blanket fluids at low concentrations

    NASA Astrophysics Data System (ADS)

    Hsu, Cheazone; Buxbaum, Robert E.

    1986-11-01

    Oxidative diffusion can extract hydrogen from metal solutions at extremely low partial pressures. The hydrogen diffuses through a metal membrane and is oxidized to water. The oxidation reaction produces the very low downstream pressures that drive the flux. This method is attractive because the flux can be proportional to the square-root of upstream pressure. For fusion reactors with liquid lithium or lithium-lead alloy breeder blankets, permeation windows provide a simple, cheap tritium extraction method. Interdiffusion rates, separation flux, window size, helium contents, tritium holdup costs, and overall costs are calculated for membranes of palladium-coated zirconium, niobium, vanadium, nickel and stainless-steel. For extracting tritium from liquid lithium using the cheapest windows, Zr-Pd, the material and labor cost is 8.0 M at 1 wppm, and is inversely proportional to tritium concentration in the lithium. The tritium holdup cost for the windows is 4.8 M, and for the blanket it is proportional to the blanket volume and concentration. An overall economic optimization suggests that 1 to 1.5 wppm in lithium is optimal. For extracting tritium from 17Li83Pb at 0.26 wppb, the cheapest window is V-Pd; the cost is 2.6 M$, and the tritium holdup is negligible.

  10. Recent progress in blanket materials development in the Broader Approach activities

    NASA Astrophysics Data System (ADS)

    Nishitani, T.; Tanigawa, H.; Nozawa, T.; Jitsukawa, S.; Nakamichi, M.; Hoshino, T.; Yamanishi, T.; Baluc, N.; Möslang, A.; Lindou, R.; Tosti, S.; Hodgson, E. R.; Clement Lorenzo, S.; Kohyama, A.; Kimura, A.; Shikama, T.; Hayashi, K.; Araki, M.

    2011-10-01

    As a part of the Broader Approach activities, R&D on blanket related materials, reduced-activation ferritic martensitic (RAFM) steels as a structural material, SiC f/SiC composites for flow channel insert in the liquid blanket and/or use as advanced structural material, advanced tritium breeders and neutron multiplier, has been initiated directed at DEMO. As part of the RAFM steel mass production development, a 5 ton heat of RAFM steel (F82H) was procured by Electro Slag Re-melting as the secondary melting method, which was effective in controlling unwanted impurities. An 11 ton heat of EUROFER was also produced. For the SiC f/SiC composite development, NITE- and CVI-SiC f/SiC composites were prepared as reference materials and preliminary mechanical and physical properties were measured. Also compatibility tests between SiC and Pb-17Li have been prepared, related to the He-cooled Li-Pb blanket concept. For the beryllide neutron multiplayer Be-Ti alloy development, large size rods of about 30 mm diameter were fabricated successfully in EU.

  11. US-DOE Fusion-Breeder Program: blanket design and system performance

    SciTech Connect

    Lee, J.D.

    1983-01-01

    Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of /sup 233/U per year. The /sup 233/U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U/sub 3/O/sub 8/ at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program.

  12. Development of electrically insulating coatings on vanadium alloys for lithium-cooled blankets

    SciTech Connect

    Smith, D.L.; Natesan, K.; Park, J.H.; Mattas, R.; Reed, C.

    1997-10-01

    The self-cooled lithium blanket concept with a vanadium structure offers a potential for high performance with attractive safety and environmental features. Based on blanket design studies, it became apparent that electrically insulating duct walls would be required to reduce the magnetohydrodynamic (MHD) pressure drop for liquid metal-cooled blankets for high magnetic field fusion devices. As a result, development of insulator coatings was recommended as the most appropriate approach for resolving this issue. Oxides such as CaO, Y{sub 2}O{sub 3}, BeO, MgO, MgAl{sub 2}O{sub 4}, and Y{sub 3}Al{sub 2}O{sub 12} and nitrides such as AlN, BN and Si{sub 3}N{sub 2} were initially considered potential candidate coating materials. Based on results of scoping studies, CaO and AlN have been selected as primary candidates for further development. Progress on the development of CaO and AlN coatings, including in-situ formation and electrical properties measurements, are summarized in this paper.

  13. Beta cloth durability assessment for Space Station Freedom (SSF) Multi-Layer Insulation (MLI) blanket covers

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L.; Jacobs, Stephen; Le, Julie

    1993-01-01

    MLI blankets for the Space Station Freedom (SSF) must comply with general program requirements and recommendations for long life and durability in the low-Earth orbit (LEO) environment. Atomic oxygen and solar ultraviolet/vacuum ultraviolet are the most important factors in the SSF natural environment which affect materials life. Two types of Beta cloth (Teflon coated woven glass fabric), which had been proposed as MLI blanket covers, were tested for long-term durability in the LEO environment. General resistance to atomic oxygen attack and permeation were evaluated in the high velocity atomic oxygen beam system at Los Alamos National Laboratories. Long-term exposure to the LEO environment was simulated in the laboratory using a radio frequency oxygen plasma asher. The plasma asher treated Beta cloth specimens were tested for thermo-optical properties and mechanical durability. Space exposure data from the Long Duration Exposure Facility and the Intelsat Solar Array Coupon were also used in the durability assessment. Beta cloth fabricated to Rockwell specification MBO 135-027 (Chemglas 250) was shown to have acceptable durability for general use as an MLI blanket cover material in the LEO environment while Sheldahl G414500 should be used only in locations which are protected from direct Ram atomic oxygen.

  14. Modeling the liquid flow in up-flow anaerobic sludge blanket reactors

    SciTech Connect

    Bolle, W.L.; Van Breugel, J.; Van Eybergen, G.C.; Kossen, N.W.F.; Zoetemeyer, R.J.

    1986-11-01

    By means of stimulus-response experiments and Li+ tracer, models for the fluid flow in a 30-cubic m UASB reactor, used for the anaerobic treatment of wastewater, were tested. From the model with the best fit it could be derived that both the sludge bed and the sludge blanket can be described as perfectly mixed tank reactors with short-circuiting flows; the settler volume acts like a plug-flow region. Apart from the volumes of the different flow regions, two parameters are necessary and sufficient to describe the fluid flow in a well functioning UASB reactor, i.e., the short-circuiting flow over the sludge bed and the short-circuiting flow over the sludge blanket. The volumes could be measured accurately. The short-circuiting flow over the sludge bed is a linear function of the sludge bed height. When the optimal height of the sludge bed is defined as the height for which the short-ciruiting flows are as small as possible, a bed-height of 3.5-4 m is sufficient (for superficial gas velocities between 1 and 1.5 m/h). This is in contradiction to the results of other authors. The short-circuiting flows over the sludge bed and the sludge blanket were also influenced by the superficial gas velocity. 7 references.

  15. Blanketing effect of expansion foam on liquefied natural gas (LNG) spillage pool.

    PubMed

    Zhang, Bin; Liu, Yi; Olewski, Tomasz; Vechot, Luc; Mannan, M Sam

    2014-09-15

    With increasing consumption of natural gas, the safety of liquefied natural gas (LNG) utilization has become an issue that requires a comprehensive study on the risk of LNG spillage in facilities with mitigation measures. The immediate hazard associated with an LNG spill is the vapor hazard, i.e., a flammable vapor cloud at the ground level, due to rapid vaporization and dense gas behavior. It was believed that high expansion foam mitigated LNG vapor hazard through warming effect (raising vapor buoyancy), but the boil-off effect increased vaporization rate due to the heat from water drainage of foam. This work reveals the existence of blocking effect (blocking convection and radiation to the pool) to reduce vaporization rate. The blanketing effect on source term (vaporization rate) is a combination of boil-off and blocking effect, which was quantitatively studied through seven tests conducted in a wind tunnel with liquid nitrogen. Since the blocking effect reduces more heat to the pool than the boil-off effect adds, the blanketing effect contributes to the net reduction of heat convection and radiation to the pool by 70%. Water drainage rate of high expansion foam is essential to determine the effectiveness of blanketing effect, since water provides the boil-off effect. PMID:25194555

  16. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    NASA Astrophysics Data System (ADS)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  17. Effect of thick blanket modules on neoclassical tearing mode locking in ITER

    NASA Astrophysics Data System (ADS)

    La Haye, R. J.; Paz-Soldan, C.; Liu, Y.

    2015-11-01

    m/n=2/1 tearing modes can be slowed and stop rotating (lock) by eddy currents induced in resistive walls, a particular issue in ITER with large inertia and low applied torque. Previous estimates of tolerable 2/1 island widths in ITER, based on a forecast of initial island rotation, the n=1 resistive penetration time of the innervacuumvessel wall and benchmarked to DIII-D high-torque plasmas, found that the ITER ECCD system could catch and subdue such islands before they persisted long and grew large enough to lock. However, rotating tearing modes in ITER will also induce eddy currents in the blanket as the effective first wall that can shield the inner vessel. The closer fitting blanket wall has a much shorter time constant and will allow several times smaller islands to lock several times faster in ITER. Recent DIII-D ITER baseline scenario plasmas with low-applied torque allow better modeling and scaling to ITER with the blanket as the first resistive wall. This motivates using the ITER ECCD system in a CW preemptive operation of as little as a well-aligned 3 MW to avoid destabilizing the 2/1 NTM. Work supported by the US DOE under DE-FC02-04ER54698.

  18. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  19. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    SciTech Connect

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop.

  20. Material problems and requirements related to the development of fusion blankets: The designer point of view

    NASA Astrophysics Data System (ADS)

    Donne, M. Dalle; Harries, D. R.; Kalinin, G.; Mattas, R.; Mori, S.

    1994-09-01

    The structural materials considered for solid and liquid metal breeder blankets are the austenitic and martensitic steels and vanadium alloys. The principal concerns with these materials are: (a) the high-temperature-induced swelling of the austenitic steels, (b) the low temperature irradiation embrittlement of martensitic steels, and (c) the exact specification of the preferred alloy composition(s), properties during and following irradiation, and technological aspects (fabrication and welding) for the vanadium alloys. Solid breeder blankets are based on the use of lithiated ceramics such as Li 2O, LiAlO 2, Li 4SiO 4 and Li 2ZrO 3 and beryllium as a neutron multiplier. The main uncertainty with these materials is their behaviour under irradiation, particularly at higher burnups and fluences than have been achieved hitherto. Liquid metal blankets, utilising pure Li or the LiPb eutectic as the tritium breeding material, can be either self- or separately-cooled; separate coolants include water (with LiPb) and helium. The important materials issues with the LiPb are the development of permeation barriers to contain the tritium and, for the self-cooled option, electrical insulators to reduce the MHD pressure drop to acceptable levels.

  1. Tritium permeation through steam generator tubing of helium-cooled ceramic breeder blankets

    SciTech Connect

    Fuetterer, M.; Raepsaet, X.; Proust, E.

    1994-12-31

    The potential sources of tritium contamination of the helium-coolant of ceramic breeder blankets have been evaluated in a previous paper for the specific case of the European BIT DEMO blanket. This evaluation associated with a rough assessment of the permeability to tritium of the tubing of helium-heated steam generators confirmed that the control of tritium losses to the steam circuit is a critical issue for this class of blanket requiring developments in three areas: (1) permeation barriers, (2) tritium recovery processes maintaining a very low concentration in tritiated species in the coolant, and (3) methods for controlling the chemistry of the coolant. Consequently, in order to define the specifications of these developments, a detailed evaluation of the permeability to tritium of helium-heated steam generators (SGs) was performed, which will be reported in this paper. This study includes the definition of the thermal-hydraulic operating conditions of the SGs through thermodynamic cycle calculations, and its thermal-hydraulic design. The obtained geometry, area and temperature profiles along the tubes are then used to estimate, based on relevant permeability data, the tritium permeation through the SG as a function of the composition in tritiated species of the coolant. The implications of these results, in terms of requirements for the considered tritium control methods, will also be discussed on the basis of expected limits in tritium release to the steam circuit.

  2. Test Strategy for the European HCPB Test Blanket Module in ITER

    SciTech Connect

    Boccaccini, L.V.; Meyder, R.; Fischer, U.

    2005-05-15

    According to the European Blanket Programme two blanket concepts, the Helium Cooled Pebble Bed (HCPB) and a Helium Cooled Lithium Lead (HCLL) will be tested in ITER. During 2004 the test blanket modules (TBM) of both concepts were redesigned with the goal to use as much as possible similar design options and fabrication techniques for both types in order to reduce the European effort for TBM development. The result is a robust TBM box being able to withstand 8 MPa internal pressure in case of in-box LOCA; the TBM box consists of First wall (FW), caps, stiffening grid and manifolds. The box is filled with typically 18 and 24 breeding units (BU), for HCPB and HCLL respectively. A breeding unit has about 200 mm in poloidal and toroidal direction and about 400 mm in radial direction; the design is adapted to contain and cooling ceramic breeder/beryllium pebble beds for the HCPB and eutectic Lithium-Lead for the HCLL.The use of a new material, EUROFER, and the innovative design of these Helium Cooled components call for a large qualification programme before the installation in ITER; availability and safety of ITER should not be jeopardised by a failure of these components. Fabrication technologies especially in the welding processes (diffusion welding, EB, TIG, LASER) need to be tested in the manufacturing of large mock-ups; an extensive out-of-pile programme in Helium facility should be foreseen for the verification of the concept from basic helium cooling functions (uniformity of flow in parallel channels, heat transfer coefficient in FW, etc.) up to the verification of large portions of the TBM design under relevant ITER loading.In ITER the TBM will have the main objective to collect information that will contribute to the final design of DEMO blankets. A strategy has been proposed in 2001 that leads to the tests in ITER 4 different Test Blanket Modules (TBM's) type during the first 10 years of ITER operation. For the new HCPB design this strategy is confirmed with

  3. Reflective Blankets Do Not Effect Cooling Rates after Running in Hot, Humid Conditions

    PubMed Central

    REYNOLDS, KORY A.; EVANICH, JOHN J.; EBERMAN, LINDSEY E.

    2015-01-01

    Reflective blankets (RB) are often provided at the conclusion of endurance events, even in extreme environments. The implications could be dangerous if increased core body temperature (CBT) is exacerbated by RB. To evaluate the effect of RB on cooling rate for individuals walking or sitting after intense running. Pilot, randomized control trial experimental design. Environmental chamber. Recreational runners (age=25±5y; mass=76.8±16.7kg; height=177±9cm) completed an 8km (actual mean distance=7.5±1.1km). We randomly assigned participants into one of four groups: walking with blanket (WB=5), walking without blanket (WNB=5), sitting with blanket (SB=5), or sitting without blanket (SNB=4). Participants ran on a treadmill at their own pace until volitional exhaustion, achieving the 8km distance, or experiencing CBT=40°C. Every three minutes during the running (time determined by pace) and cooling protocol (62 min in chamber), we measured CBT, HR, and Borg scale, and environmental conditions. We evaluated cooling rate, peak physiological variables, pace, and environment by condition using a Kruskal-Wallis non-parametric one-way ANOVAs. We identified similar exercise sessions (df=3; CBT χ2=0.921, p=0.82; HR χ2=7.446, p=0.06; Borg χ2= 5.732, p=0.13; pace χ2=0.747, p=0.86) and similar environmental characteristics between conditions (df=3; Wet Bulb Globe Temperature=26.18±2.78°C, χ2=1.552, p=0.67). No significant differences between conditions on cooling rate (df=3, χ2=2.301, p=0.512) were found, suggesting RBs neither cool nor heat the body, whether seated (SB=0.021±0.011deg/min; SNB=0.029±0.002deg/min) or walking (WB=0.015±0.025deg/min; WNB=0.021±0.011deg/min) in a hot, humid environment. CBT in distance runners is not altered by the use of a RB during a seated or walking cool down after a strenuous run. PMID:27182414

  4. Water movement through blanket peat is dominated by a complicated pattern of near-surface flows

    NASA Astrophysics Data System (ADS)

    Turner, Ed; Baird, Andy; Billett, Mike; Chapman, Pippa; Dinsmore, Kerry; Holden, Joseph

    2015-04-01

    Blanket peatland formation and functioning depend strongly on hydrology. Omitting the potential for pipe flow, the acrotelm-catotelm model is still widely held to apply to blanket peatlands. In the model, water flow through the peat profile is dominated by near-surface flow in the acrotelm, whereas water movement below the level of (near) permanent saturation (the catotelm) is characterised by very low hydraulic conductivity (K). Whilst some work has been done on characterising Kat different depths in blanket peatlands, very little is known about near-surface K, particularly with respect to how it varies between microforms and over fine spatial scales. We undertook a detailed investigation of near-surface (0 - 12 cm) and deeper (30 and 50 cm) K at a blanket peatland site in the Flow Country in Scotland (UK). Near-surface Kof peat samples taken across a range of microforms was measured vertically (Kv) and horizontally (Kh) in the laboratory using a new 'split cylinder' method (n = 48 excluding repeat tests). K30 (n = 20) andK50 (n = 20) were estimated in situ using the piezometer or seepage-tube method. To help our interpretation of the near-surface K measurements we recorded the vegetation cover from where the peat samples were taken and characterised each peat sample in terms of its plant macrofossil assemblage and dry bulk density. We found that Kvand Khwere highly variable between microforms in the near-surface samples, ranging over two orders of magnitude (0.489 - 0.003 cm s-1). Kernel density plots show that Kvwas most commonly in the region of ~0.03 cm s-1 at 0 - 6 cm, and ~0.015 cm s-1 at 6 - 12 cm, whereas Kh was ~0.05 and ~0.001 cm s-1 respectively. These data reveal a high degree of absolute variability and anisotropy in K over small scales. The deeper K30and K50 values were typically an order of magnitude or more lower than the near-surface K, and were less variable between test locations with the exception of poorly humified Sphagnum-dominated peat

  5. Fusion option to dispose of spent nuclear fuel and transuranic elements

    SciTech Connect

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  6. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    SciTech Connect

    Greenspan, Ehud

    2015-11-04

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and

  7. Strain rate dependence of the tensile properties of V-(4--5%)Cr-(4--5%)Ti irradiated in EBR-II and HFBR

    SciTech Connect

    Zinkle, S.J.; Snead, L.L.; Robertson, J.P.; Rowcliffe, A.F.

    1998-03-01

    Elevated temperature tensile tests performed on V-(405)Cr-(4-5)Ti indicate that the yield stress increases with increasing strain rate for irradiation and test temperatures near 200 C, and decreases with increasing strain rate for irradiation and test temperatures near 400 C. This observation is in qualitative agreement with the temperature-dependent strain rate effects observed on unirradiated specimens, and implies that some interstitial solute remains free to migrate in irradiated specimens. Additional strain rate data at different temperatures are needed.

  8. VAPORIZATION OF ELEMENTAL MERCURY FROM POOLS OF MOLTEN LEAD AT LOW CONCENTRATIONS.

    SciTech Connect

    GREENE,G.A.; FINFROCK,C.C.

    2000-10-01

    Should coolant accidentally be lost to the APT (Accelerator Production of Tritium) blanket and target, and the decay heat in the target be deposited in the surrounding blanket by thermal radiation, temperatures in the blanket modules could exceed structural limits and cause a physical collapse of the blanket modules into a non-coolable geometry. Such a sequence of unmitigated events could result in some melting of the APT blanket and create the potential for the release of mercury into the target-blanket cavity air space. Experiments were conducted which simulate such hypothetical accident conditions in order to measure the rate of vaporization of elemental mercury from pools of molten lead to quantify the possible severe accident source term for the APT blanket region. Molten pools of from 0.01% to 0.10% mercury in lead were prepared under inert conditions. Experiments were conducted, which varied in duration from several hours to as long as a month, to measure the mercury vaporization from the lead pools. The melt pools and gas atmospheres were held fixed at 340 C during the tests. Parameters which were varied in the tests included the mercury concentration, gas flow rate over the melt and agitation of the melt, gas atmosphere composition and the addition of aluminum to the melt. The vaporization of mercury was found to scale roughly linearly with the concentration of mercury in the pool. Variations in the gas flow rates were not found to have any effect on the mass transfer, however agitation of the melt by a submerged stirrer did enhance the mercury vaporization rate. The rate of mercury vaporization with an argon (inert) atmosphere was found to exceed that for an air (oxidizing) atmosphere by as much as a factor of from ten to 20; the causal factor in this variation was the formation of an oxide layer over the melt pool with the air atmosphere which served to retard mass transfer across the melt-atmosphere interface. Aluminum was introduced into the melt to

  9. Elemental health

    SciTech Connect

    Tonneson, L.C.

    1997-01-01

    Trace elements used in nutritional supplements and vitamins are discussed in the article. Relevant studies are briefly cited regarding the health effects of selenium, chromium, germanium, silicon, zinc, magnesium, silver, manganese, ruthenium, lithium, and vanadium. The toxicity and food sources are listed for some of the elements. A brief summary is also provided of the nutritional supplements market.

  10. The Practical Challenges of Evaluating a Blanket Emergency Feeding Programme in Northern Kenya

    PubMed Central

    Hall, Andrew; Oirere, Moragwa; Thurstans, Susan; Ndumi, Assumpta; Sibson, Victoria

    2011-01-01

    A blanket supplementary feeding programme for young children was implemented for four months in five northern districts of Kenya from January 2010 because of fears of food insecurity exacerbated by drought. An attempt to evaluate the impact of the food on children's anthropometric status was put in place in three districts. The main aim of the analysis was to assess the quality of the data on the cohort of children studied in the evaluation and to propose methods by which it could be improved to evaluate future blanket feeding programmes. Data on the name, age, sex, weight and height of a systematic sample of children recruited at 61 food distribution sites were collected at the first, second and third rounds and again at an extra, fifth food distribution, offered only to the evaluation subjects. Of the 3,544 children enrolled, 483 (13.63%) did not collect a fifth ration. Of the 2,640 children who were considered by their name to be the same at the first and fifth food distribution (13% were different), data on only 902 children (34.17%) were considered acceptable based on their age (an arbitrary ±3 months different) and their length or height (between >−1 or ≤4 cm different) at the two instances they were seen. Data on nearly two thirds of children were of questionable quality. The main reasons for the poor quality data were inconsistencies in estimating age or because caretakers may have brought different children. Recommendations are made about how to improve data quality including ensuring that entry to a blanket feeding programme is clearly based on height, not age, to avoid misreporting age; careful identification of subjects at all contacts; and using well-trained, specialist evaluation staff. PMID:22073119

  11. EOS AM-1 GaAs/Ge flexible blanket solar array

    SciTech Connect

    Herriage, M.J.; Kurland, R.M.; Faust, C.D.; Gaddy, E.M.; Keys, D.J.

    1996-12-31

    The Earth Observing System (EOS) AM-1 solar array will be the first operational flexible blanket array to use gallium arsenide on germanium (GaAs/Ge) solar cells. It will be one of the first arrays of any kind to use thin GaAs/Ge solar cells and to produce power at high voltage. The array thereby represents a significant advance in spacecraft array technology. This paper describes the progress made and the problems encountered since Kimber et al (1993) reported on the tradeoffs that led to the array`s configuration. The EOS AM-1 spacecraft, due to launch in 1998, is the first in a series of Goddard Space Flight Center remote sensing spacecraft. It requires a single wing solar array with an end of life (EOL) power requirement of 5 kW at 127 volts after a five year mission in a low earth polar orbit. Spacecraft system level design issues dealing with packaging volume, attitude control, weight and array size dictated that a low aspect ratio (2:1) flexible blanket solar array with 0.14 mm thick GaAs/Ge cells be used. Design of the array was contracted to TRW Space and Electronics Group in 1992 by Lockheed Martin Corporation. The flexible blanket array configuration proposed was based on the Advanced Photovoltaic Solar Array (APSA) flat-pack foldout concept demonstrated by TRW under an earlier Jet Propulsion Laboratory exploratory development contract, (Stella and Kurland, 1992). The EOS AM-1 solar array assembly (SAA) program is in its post critical design review phase, with design details completed, flight hardware in production, and the flight hardware verification test program about to begin. The protoflight solar array wing is scheduled for delivery in 1997.

  12. The practical challenges of evaluating a blanket emergency feeding programme in northern Kenya.

    PubMed

    Hall, Andrew; Oirere, Moragwa; Thurstans, Susan; Ndumi, Assumpta; Sibson, Victoria

    2011-01-01

    A blanket supplementary feeding programme for young children was implemented for four months in five northern districts of Kenya from January 2010 because of fears of food insecurity exacerbated by drought. An attempt to evaluate the impact of the food on children's anthropometric status was put in place in three districts. The main aim of the analysis was to assess the quality of the data on the cohort of children studied in the evaluation and to propose methods by which it could be improved to evaluate future blanket feeding programmes. Data on the name, age, sex, weight and height of a systematic sample of children recruited at 61 food distribution sites were collected at the first, second and third rounds and again at an extra, fifth food distribution, offered only to the evaluation subjects. Of the 3,544 children enrolled, 483 (13.63%) did not collect a fifth ration. Of the 2,640 children who were considered by their name to be the same at the first and fifth food distribution (13% were different), data on only 902 children (34.17%) were considered acceptable based on their age (an arbitrary ±3 months different) and their length or height (between >-1 or ≤4 cm different) at the two instances they were seen. Data on nearly two thirds of children were of questionable quality. The main reasons for the poor quality data were inconsistencies in estimating age or because caretakers may have brought different children. Recommendations are made about how to improve data quality including ensuring that entry to a blanket feeding programme is clearly based on height, not age, to avoid misreporting age; careful identification of subjects at all contacts; and using well-trained, specialist evaluation staff. PMID:22073119

  13. Spatial variation in concentrations of dissolved nitrogen species in an upland blanket peat catchment.

    PubMed

    Cundill, A P; Chapman, P J; Adamson, J K

    2007-02-01

    The concentration of nitrogen (N), particularly as nitrate (NO3-N), in upland streams, lakes and rivers is frequently used as a diagnostic of the vulnerability of upland ecosystems to increased atmospheric nitrogen deposition and N saturation. The N content of running waters, however, is generally assessed on the basis of sampling at a limited number of points in space and time within the catchment under investigation. The current study was conducted at Trout Beck, an 11.5 km2 blanket peat-dominated catchment in the North Pennine uplands of the UK. Results from sampling at 33 sites within this catchment demonstrated that the concentrations of all dissolved N species were highly variable, even over short distances. Statistical relationships between the concentrations of NO3-N and dissolved organic nitrogen (DON) and percentage catchment cover of Calluna/Eriophorum and Eriophorum vegetation were found. However, it was also noted that in catchments containing limestone outcrops, NO3-N concentration was much higher than in catchments where runoff was sourced directly from the blanket peat surface. It is possible that NH4-N and DON leached from the blanket peat are mineralised and nitrified, providing a source for the NO3-N found in the river channels. Overall, the current study suggests that interpretations of N-saturation based on river water chemistry measurements at a single point must be treated cautiously, and that the influence of catchment-scale physical factors, such as vegetation and geology cover on the concentration of dissolved N species in upland river waters should not be ignored. PMID:17182088

  14. Current status of final design and R&D for ITER blanket shield blocks in Korea

    NASA Astrophysics Data System (ADS)

    Ha, M. S.; Kim, S. W.; Jung, H. C.; Hwang, H. S.; Heo, Y. G.; Kim, D. H.; Ahn, H. J.; Lee, H. G.; Jung, K. J.

    2015-07-01

    The main function of the ITER blanket shield block (SB) is to provide nuclear shielding and support the first wall (FW) panel. It needs to accommodate all the components located on the vacuum vessel (in particular the in-vessel coils, blanket manifolds and the diagnostics). The conceptual, preliminary and final design reviews have been completed in the framework of the Blanket Integrated Product Team. The Korean Domestic Agency has successfully completed not only the final design activities, including thermo-hydraulic and thermo-mechanical analyses for SBs #2, #6, #8 and #16, but also the SB full scale prototype (FSP) pre-qualification program prior to issuing of the procurement agreement. SBs #2 and #6 are located at the in-board region of the tokamak. The pressure drop was less than 0.3 MPa and fully satisfied the design criteria. The thermo-mechanical stresses were also allowable even though the peak stresses occurred at nearby radial slit end holes, and their fatigue lives were evaluated over many more than 30 000 cycles. SB #8 is one of the most difficult modules to design, since this module will endure severe thermal loading not only from nuclear heating but also from plasma heat flux at uncovered regions by the FW. In order to resolve this design issue, the neutral beam shine-through module concept was applied to the FW uncovered region and it has been successfully verified as a possible design solution. SB #16 is located at the out-board central region of the tokamak. This module is under much higher nuclear loading than other modules and is covered by an enhanced heat flux FW panel. In the early design stage, many cooling headers on the front region were inserted to mitigate peak stresses near the access hole and radial slit end hole. However, the cooling headers on the front region needed to be removed in order to reduce the risk from cover welding during manufacturing. A few cooling headers now remain after efforts through several iterations to remove

  15. Analysis of the thorium axial blanket experiments in the PROTEUS reactor

    SciTech Connect

    White, J. R.; Ingersoll, D. T.; Schmocker, U.

    1980-01-01

    An extensive program of reactor physics experiments in GCFR fuel pin lattices has been completed recently at the PROTEUS critical facility located at EIR laboratory in Switzerland. The PROTEUS reactor consists of a central test zone surrounded by a uranium buffer and thermal driver region. The test lattices included a PuO/sub 2//UO/sub 2/ fuel region with internal and axial blankets of UO/sub 2/, ThO/sub 2/, and thorium metal. Detailed analysis of the thorium-bearing lattices has been performed at EIR and at ORNL in order to validate nuclear data and methods used for reactor physics analysis of advanced GCFR designs.

  16. Design of the waveguide for microwave heating of solid lithium ceramic blankets

    SciTech Connect

    Kustom, R.L.; Fendley, P.; Tidona, J.

    1985-01-01

    A description is given of the design of a dielectric-loaded waveguide for thermohydraulic testing of solid ceramic tritium breeder material in a non-nuclear environment. The dielectric-loaded waveguide provides uniform heating over module surfaces that would face a fusion reactor plasma and simulates the exponential power decay characteristic of the neutron flux over the high power region of the blankets. A 200-MHz design suitable for modules with cross section of up to 20 x 40 cm is presented.

  17. Electromagnetic Launch Vehicle Fairing and Acoustic Blanket Model of Received Power Using FEKO

    NASA Technical Reports Server (NTRS)

    Trout, Dawn H.; Stanley, James E.; Wahid, Parveen F.

    2011-01-01

    Evaluating the impact of radio frequency transmission in vehicle fairings is important to sensitive spacecraft. This paper employees the Multilevel Fast Multipole Method (MLFMM) feature of a commercial electromagnetic tool to model the fairing electromagnetic environment in the presence of an internal transmitter. This work is an extension of the perfect electric conductor model that was used to represent the bare aluminum internal fairing cavity. This fairing model includes typical acoustic blanketing commonly used in vehicle fairings. Representative material models within FEKO were successfully used to simulate the test case.

  18. Energy and mass distributions of impact ejecta blankets on the moon and Mercury

    NASA Technical Reports Server (NTRS)

    Ahrens, T. J.; Okeefe, J. D.

    1978-01-01

    The paper applies previously calculated impact-induced flow fields (O'Keefe and Ahrens, 1977) resulting from interaction of 5-cm radius gabbroic anorthosite impactor with a half-space of the same material, at various velocities, to obtain mass and energy ejecta distributions. Whereas earlier results described the ejecta distribution from a 15 km/s impact of an iron object on the moon in terms of mass vs. distance, the present results describe, at a given distance from the impact, the energy content as a function of depth, i.e., the thermal structure of ejecta blankets. Pertinent computational methods are included, and several tables and plots supplement the text.

  19. Analysis of in-situ tritium recovery from solid fusion-reactor blankets

    SciTech Connect

    Smith, D.L.; Clemmer, R.G.; Jankus, V.Z.; Rest, J.

    1980-01-01

    The proposed concept for in-situ tritium recovery from the STARFIRE blanket involves circulation of a low pressure (approx. 0.05 MPa) helium through formed channels in the highly porous solid breeding material. Tritium generated within the grains must diffuse to the grain boundaries, migrate through the grain boundaries to the particle surface and then percolate through the packed bed to the helium purge channel. Highly porous ..cap alpha..-LiAlO/sub 2/ with a bimodal pore distribution is proposed for the breeding material to facilitate the tritium release.

  20. Thermal blanket metallic film groundstrap and second surface mirror vulnerability to arc discharges

    NASA Technical Reports Server (NTRS)

    Inouye, G. T.; Sanders, N. L.; Komatsu, G. K.; Valles, J. R.; Sellen, J. M., Jr.

    1979-01-01

    Available data on the geosynchronous orbit energetic plasma environment were examined, and a crude model was generated to permit an estimation to be made of the number of arc discharges per year to which a thermal blanket groundstrap would be subjected. Laboratory experiments and a survey of the literature on arc discharge characteristics were performed to define typical and worst case arc discharge current waveforms. In-air tests of different groundstrap configurations to a standardized test pulse were performed and a wide variability of durability values were found. A groundstrap technique, not used thus far, was found to be far superior than the others.

  1. Lithium Blanket Module dosimetry measurements at the LOTUS 14-MeV neutron source facility

    SciTech Connect

    Tsang, F.Y.; Leo, W.R.; Sahraoui, C.; Wuthrich, S.; Harker, Y.D.

    1986-01-01

    This paper describes the measurements and results of the dosimeter material reaction rates inside the Lithium Blanket Module (LBM) after irradiation by the LOTUS 14-MeV neutron source at the Ecole Polytechnique Federale de Lausanne. The measurement program has been designed to utilize sets of passive dosimeter materials in the form of foils and wires. The dosimetry materials reaction thresholds and interaction response ranges chosen for this series of measurements encompass the entire neutron spectra along the full length of the LBM fuel rods.

  2. Progress of Integral Experiments in Benchmark Fission Assemblies for a Blanket of Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Liu, R.; Zhu, T. H.; Yan, X. S.; Lu, X. X.; Jiang, L.; Wang, M.; Han, Z. J.; Wen, Z. W.; Lin, J. F.; Yang, Y. W.

    2014-04-01

    This article describes recent progress in integral neutronics experiments in benchmark fission assemblies for the blanket design in a hybrid reactor. The spherical assemblies consist of three layers of depleted uranium shells and several layers of polyethylene shells, separately. In the assemblies with centralizing the D-T neutron source, the plutonium production rates, uranium fission rates and leakage neutron spectra are measured. The measured results are compared to the calculated ones with the MCNP-4B code and ENDF/B-VI library data, available.

  3. An analysis of LDEF-exposed silvered FEP teflon thermal blanket material

    NASA Technical Reports Server (NTRS)

    Young, Philip R.; Slemp, Wayne S.

    1991-01-01

    The characterization of selected silvered fluorinated ethylene propylene (FEP) teflon thermal blanket material which received 5 years and 9 months of exposure to the LEO environment on the Long Duration Exposure Facility is reported. X-ray photoelectron spectroscopy, infrared, and thermal analyses did not detect a significant change at the molecular level as the result of this exposure. However, various microscopic analyses revealed a roughening of the coating surface due to atomic oxygen erosion which resulted in some materials changing from specular reflectors of visible radiation to diffuse reflectors. The potential effect of silicon-containing molecular contamination on these materials is addressed.

  4. A helium-cooled blanket design of the low aspect ratio reactor

    SciTech Connect

    Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

    1998-03-01

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

  6. Elemental Education.

    ERIC Educational Resources Information Center

    Daniel, Esther Gnanamalar Sarojini; Saat, Rohaida Mohd.

    2001-01-01

    Introduces a learning module integrating three disciplines--physics, chemistry, and biology--and based on four elements: carbon, oxygen, hydrogen, and silicon. Includes atomic model and silicon-based life activities. (YDS)

  7. Superheavy Elements

    ERIC Educational Resources Information Center

    Tsang, Chin Fu

    1975-01-01

    Discusses the possibility of creating elements with an atomic number of around 114. Describes the underlying physics responsible for the limited extent of the periodic table and enumerates problems that must be overcome in creating a superheavy nucleus. (GS)

  8. Treatment of high-strength ethylene glycol waste water in an expanded granular sludge blanket reactor: use of PVA-gel beads as a biocarrier.

    PubMed

    Jin, Yue; Wang, Dunqiu; Zhang, Wenjie

    2016-01-01

    Industrial-scale use of polyvinyl alcohol (PVA)-gel beads as biocarriers is still not being implemented due to the lack of understanding regarding the optimal operational parameters. In this study, the parameters for organic loading rate (OLR), alkalinity, recycle rate, and addition of trace elements were investigated in an expanded granular sludge blanket reactor (EGSB) treating high-strength ethylene glycol wastewater (EG) with PVA-gel beads as biocarrier. Stable chemical oxygen demand (COD) removal efficiencies of 95 % or greater were achieved, and continuous treatment was demonstrated with appropriate parameters being an OLR of 15 kg COD/m(3)/day, NaHCO3 added at 400 mg/L, a recycle rate of 15 L/h, and no addition of trace elements addition. A biogas production yield rate of 0.24 m(3)/kg COD was achieved in this study. A large number of long rod-shaped bacteria (Methanosaeta), were found with low acetate concentration in the EGSB reactor. PMID:27386305

  9. Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

    SciTech Connect

    L. C. Cadwallader; C. P. C. Wong; M. Abdou; B. B. Morely; B.J Merrill

    2014-10-01

    A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

  10. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

  11. Ferritic-Martensitic steel Test Blanket Modules: Status and future needs for design criteria requirements and fabrication validation

    NASA Astrophysics Data System (ADS)

    Salavy, J.-F.; Aiello, G.; Aubert, P.; Boccaccini, L. V.; Daichendt, M.; De Dinechin, G.; Diegele, E.; Giancarli, L. M.; Lässer, R.; Neuberger, H.; Poitevin, Y.; Stephan, Y.; Rampal, G.; Rigal, E.

    2009-04-01

    The Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble Bed are the two breeding blankets concepts for the DEMO reactor which have been selected by EU to be tested in ITER in the framework of the Test Blanket Module projects. They both use a 9%CrWVTa Reduced Activation Ferritic-Martensitic steel, called EUROFER, as structural material and helium as coolant. This paper gives an overview of the status of the EUROFER qualification program and discusses the future needs for design criteria requirements and fabrication validation.

  12. Occurrence of the blanketing sporadic E layer during the recovery phase of the October 2003 superstorm

    NASA Astrophysics Data System (ADS)

    Denardini, Clezio Marcos; Resende, Laysa Cristina Araújo; Moro, Juliano; Chen, Sony Su

    2016-05-01

    We have routinely monitored the total frequency ( ftEs) and the blanketing frequency ( fbEs) of sporadic E layers with the digital sounder under the magnetic equator in the Brazilian sector. Sporadic layers appear in the equatorial region (Esq) at heights between 90 and 130 km, mainly due to irregularities in the equatorial electrojet current. However, during the recovery phase of the October 2003 superstorm, an anomalous intensification of the ionospheric density that exceeded the normal ambient background values for local time and location was observed. The parameter fbEs rose to almost 7.5 MHz during this event, due to a type "c" blanketing sporadic layer (Esc), which is driven by wind shear. This result is discussed in terms of the atmosphere dynamics based on magnetic signature of the equatorial electrojet current using magnetometer data. Also, using data measured by sensors onboard the Geostationary Operational Environmental Satellite (GOES) 10 we analyze the possible influence of the solar flare-associated X-ray flux as an additional source of ionization.

  13. Analysis of the thorium axial blanket experiments in the proteus reactor

    SciTech Connect

    White, J.R.; Ingersoll, D.T.

    1980-12-01

    Detailed analysis has been completed for the ThO/sub 2/ and Th-metal axial blanket experiments performed at the Swiss PROTEUS critical facility in order to compare reaction rates and neutron spectra measured in prototypic GCFR configurations with calculated results. The PROTEUS configurations allowed the analysis of infinitely dilute thorium data in a PuO/sub 2//UO/sub 2/ fast lattice spectrum at core center as well as the analysis of resonance self-shielding effects in the thorium-bearing axial blankets. These comparisons indicate that significant deficiencies still exist in the latest evaluated infinitely dilute thorium data file. Specifically, the analysis showed that the /sup 232/Th capture is underpredicted by ENDF/B-IV data, and the discrepancies are further exaggerated by ENDF/B-V data. On the other hand, ENDF/B-V /sup 232/Th fission data appear to be significantly improved relative to ENDF/B-IV data, while discrepancies are extremely large for the (n,2n) process in both data files. Finally, the (n,n') cross sections for thorium also appear improved in ENDF/B-V, except for a small energy range just above the 50 keV threshold. Therefore, these combined data deficiencies suggest that relatively large uncertainties should be associated with many of the results obtained from recent fast reactor alternate fuel cycle analyses. 38 figures, 12 tables.

  14. Terra Flexible Blanket Solar Array Deployment, On-Orbit Performance and Future Applications

    NASA Technical Reports Server (NTRS)

    Kurland, Richard; Schurig, Hans; Rosenfeld, Mark; Herriage, Michael; Gaddy, Edward; Keys, Denney; Faust, Carl; Andiario, William; Kurtz, Michelle; Moyer, Eric; Day, John H. (Technical Monitor)

    2000-01-01

    The Terra spacecraft (formerly identified as EOS AM1) is the flagship in a planned series of NASA/GSFC (Goddard Space Flight Center) Earth observing system satellites designed to provide information on the health of the Earth's land, oceans, air, ice, and life as a total ecological global system. It has been successfully performing its mission since a late-December 1999 launch into a 705 km polar orbit. The spacecraft is powered by a single wing, flexible blanket array using single junction (SJ) gallium arsenide/germanium (GaAs/Ge) solar cells sized to provide five year end-of-life (EOL) power of greater than 5000 watts at 127 volts. It is currently the highest voltage and power operational flexible blanket array with GaAs/Ge cells. This paper briefly describes the wing design as a basis for discussing the operation of the electronics and mechanisms used to achieve successful on-orbit deployment. Its orbital electrical performance to date will be presented and compared to analytical predictions based on ground qualification testing. The paper concludes with a brief section on future applications and performance trends using advanced multi-junction cells and weight-efficient mechanical components. A viewgraph presentation is attached that outlines the same information as the paper and includes more images of the Terra Spacecraft and its components.

  15. Terra Flexible Blanket Solar Array Deployment, On-Orbit Performance and Future Applications

    NASA Technical Reports Server (NTRS)

    Kurland, Richard; Schurig, Hans; Rosenfeld, Mark; Herriage, Michael; Gaddy, Edward; Keys, Denney; Faust, Carl; Andiario, William; Kurtz, Michelle; Moyer, Eric; Day, John H. (Technical Monitor)

    2000-01-01

    The Terra spacecraft (formerly identified as EOS AM1) is the flagship in a planned series of NASA/GSFC (Goddard Space Flight Center) Earth observing system satellites designed to provide information on the health of the Earth's land, oceans, air, ice, and life as a total ecological global system. It has been successfully performing its mission since a late-December 1999 launch into a 705 km polar orbit. The spacecraft is powered by a single wing, flexible blanket array using single junction (SJ) gallium arsenide/germanium (GaAs/Ge) solar cells sized to provide five year end-of-life (EOL) power of greater than 5000 watts at 127 volts. It is currently the highest voltage and power operational flexible blanket array with GaAs/Ge cells. This paper briefly describes the wing design as a basis for discussing the operation of the electronics and mechanisms used to achieve successful on-orbit deployment. Its orbital electrical performance to date will be presented and compared to analytical predictions based on ground qualification testing. The paper concludes with a brief section on future applications and performance trends using advanced multi-junction cells and weight-efficient mechanical components.

  16. Geomorphic clues to the Martian volatile inventory. 1: Flow ejecta blankets

    NASA Technical Reports Server (NTRS)

    Pieri, D.; Baloga, S.; Norris, M.

    1984-01-01

    There are classes of landforms whose presence on Mars is strongly suggestive, if not confirmatory, of the participation of volatiles, presumably water, in its geomorphic development: (1) valley networks, (2) outflow channels, (3) landslides, and (4) flow-ejecta blankets. The first two may represent landforms generated by the movement of volatiles from sources, while the latter two probably represent the dissipation of energy generated by forcing inputs (e.g., kinetic energy and gravity) modulated by volatiles. In many areas on Mars, all four processes have acted on the same lithologic materials and were influenced by the composition of those units, and possibility by the climatic regime at the time of their formation. One of the approaches discussed to this specific problem of landform genesis, and to the general problem of the present and past states of martian volatiles, is to attempt to constrain the distribution, amount, and history of available volatiles by using possible evidence of volatile participation expressed in the morphology of other related landforms (e.g., flow-ejecta blankets and landslides) coupled with physical models for landform genesis.

  17. Summary report for ITER task - T68: MHD facility preparation for Li/V blanket option

    SciTech Connect

    Reed, C.B.; Haglund, R.C.; Miller, M.E.

    1995-08-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To enable experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, the room-temperature ALEX (Argonne`s Liquid Metal EXperiment) NaK facility was upgraded to a 300{degrees}C lithium system. The objective of this upgrade was to modify the existing facility to the minimum extent necessary, consistent with providing a safe, flexible, and easy to operate MHD test facility which uses lithium at ITER-relevant temperatures, Hartmann numbers, and interaction parameters. The facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups. The system design description for this lithium upgrade of the ALEX facility is given in this document.

  18. Status of fuel, blanket, and absorber testing in the Fast Flux Test Facility

    SciTech Connect

    Baker, R.B.; Bard, F.E.; Leggett, R.D.; Pitner, A.L.

    1992-11-01

    Over 67,000 fuel, blanket and absorber pins have been irradiated in the Fast Flux Test Facility (FFTF) during its first 12 years of operation. Tests are run in highly controlled and monitored environments with core components similar in size to those in commercial liquid metal reactor (LMR) designs. While primary emphasis was placed on mixed oxide fuels, significant development programs have included metallic fuels, UO[sub 2] blankets, B[sub 4]C absorbers, and other fuels and materials of interest. Irradiation programs for mixed oxides have included progressively lower swelling cladding and duct alloys (e.g., 316 SS, D9 SS, and the ferritic HT9), which also have application to other core components. In many instances the current exposure levels of the advanced FFTF tests are the highest attained and reported in the literature. This paper summarizes the status of irradiation experience at the facility, presents some general conclusions, and reviews the potential for obtaining additional significant data.

  19. Status of fuel, blanket, and absorber testing in the Fast Flux Test Facility

    SciTech Connect

    Baker, R.B.; Bard, F.E.; Leggett, R.D.; Pitner, A.L.

    1992-11-01

    Over 67,000 fuel, blanket and absorber pins have been irradiated in the Fast Flux Test Facility (FFTF) during its first 12 years of operation. Tests are run in highly controlled and monitored environments with core components similar in size to those in commercial liquid metal reactor (LMR) designs. While primary emphasis was placed on mixed oxide fuels, significant development programs have included metallic fuels, UO{sub 2} blankets, B{sub 4}C absorbers, and other fuels and materials of interest. Irradiation programs for mixed oxides have included progressively lower swelling cladding and duct alloys (e.g., 316 SS, D9 SS, and the ferritic HT9), which also have application to other core components. In many instances the current exposure levels of the advanced FFTF tests are the highest attained and reported in the literature. This paper summarizes the status of irradiation experience at the facility, presents some general conclusions, and reviews the potential for obtaining additional significant data.

  20. Fluidized-bed design for ICF reactor blankets using solid-lithium compounds

    SciTech Connect

    Sucov, E.W.; Malick, F.S.; Green, L.; Hall, B.O.

    1983-01-01

    A fluidized-bed concept for blankets of dry or wetted first-wall ICF reactors using solid-lithium compounds is described. The reaction chamber is a right cylinder, 32 m high and 20 m in diameter; the blanket is composed of 36 steel tanks, 32 m high, which carry the sintered Li/sub 2/O particles in the fluidizing helium gas. Each tank has a radial thickness of 2 m which generates a tritium breeding ration (TBR) of 1.27 and absorbs over 98% of the neutron energy; reducing the thickness to 1.2 m produces a TBR of 1.2 and energy absorption of 97% which satisfy the design goals. Calculations of tritium diffusion through the grains and heat removal from the grains showed that neither could be removed by the carrier gas; tritium and heat are therefore removed by removing the grains themselves by varying the helium flow rate. The particles are continuously fed into the bottom of the tanks at 300/sup 0/C and removed at the top at 475/sup 0/C. Tritium and heat extraction are easily and conveniently done outside the reactor.