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Sample records for ebr-ii integral measurements

  1. EBR-II: summary of operating experience

    International Nuclear Information System (INIS)

    Perry, W.H.; Leman, J.D.; Lentz, G.L.; Longua, K.J.; Olson, W.H.; Shields, J.A.; Wolz, G.C.

    1978-01-01

    Experimental Breeder Reactor II (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. The primary cooling system is a submerged-pool type. The early operation of the reactor successfully demonstrated the feasibility of a sodium-cooled fast breeder reactor operating as an integrated reactor, power plant, and fuel-processing facility. In 1967, the role of EBR-II was reoriented from a demonstration plant to an irradiation facility. Many changes have been made and are continuing to be made to increase the usefulness of EBR-II for irradiation and safety tests. A review of EBR-II's operating history reveals a plant that has demonstrated high availability, stable and safe operating characteristics, and excellent performance of sodium components. Levels of radiation exposure to the operating and maintenance workers have been low; and fission-gas releases to the atmosphere have been minimal. Driver-fuel performance has been excellent. The repairability of radioactive sodium components has been successfully demonstrated a number of times. Recent highlights include installation and successful operation of (1) the hydrogen-meter leak detectors for the steam generators, (2) the cover-gas-cleanup system and (3) the cesium trap in the primary sodium. Irradiations now being conducted in EBR-II include the run-beyond-cladding breach fuel tests for mixed-oxide and carbide elements. Studies are in progress to determine EBR-II's capability for conducting important ''operational safety'' tests. These tests would extend the need and usefulness of EBR-II into the 1980's

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

    International Nuclear Information System (INIS)

    Grimm, K. N.

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    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

  4. Tightly coupled transient analysis of EBR-II

    International Nuclear Information System (INIS)

    Makowitz, H.; Lehto, W.K.; Sackett, J.I.

    1988-01-01

    A Tightly Coupled transient analysis system for the Experimental Breeder Reactor-II (EBR-II) is currently being tested. The system consists of a faster than real time high fidelity reactor simulation, advanced graphics displays, expert system coupling, and real time data coupling via the EBR-II data acquisition system to and from the plant and the control system. The base, first generation software has been developed and is presently being tested. Various subsystem couplings and the total system integration are being checked out. This system should enhance the diagnostic and prognostic capability of EBR-II in the near term and provide automatic control during startup and power maneuvering in the future, as well as serve as a testbed for new control system development for advanced reactors

  5. Operating and test experience of EBR-II

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1991-01-01

    EBR-II has operated for 27 years, the longest for any Liquid Metal Reactor (LMR) power plant. During that time, much has been learned about successful LMR operation and design. The basic lesson is that conservatism in design can pay significant dividends in operating reliability. Furthermore, such conservatism need not mean high cost. The EBR-II system emphasizes simplicity, minimizing the number of valves in the heat transport system, for example, and simplifying the primary heat-transport-system layout. Another lesson is that emphasizing reliability of the steam generating system at the sodium-water interface (by using duplex tubes in the case of EBR-II) has been well worth the higher initial costs; no problems with leakage have been encountered in EBR-II's operating history. Locating spent fuel storage in the primary tank and providing for decay heat removal by natural connective flow have also been contributors to EBR-II's success. The ability to accommodate loss of forced cooling or loss of heat sink passively has resulted in benefits for simplification, primarily through less reliance on emergency power and in not requiring the secondary sodium or steam systems to be safety grade. Also, the 'piped-pool' arrangement minimizes thermal stress to the primary tank and enhances natural convective flow. These benefits have been realized through a history of operation that has seen EBR-II evolve through four major phases in its test programs, culminating in its present mission as the Integral Fast Reactor (IFR) prototype. (author)

  6. EBR-II Data Digitization

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su-Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sackett, John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  7. Evolution of thermal-hydraulics testing in EBR-II

    International Nuclear Information System (INIS)

    Golden, G.H.; Planchon, H.P.; Sackett, J.I.; Singer, R.M.

    1987-01-01

    A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink wihout scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies. (orig.)

  8. The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    The physics design oi EBR-II. Calculations of the static, dynamic and long-term reactivity behaviour of EBR-II are reported together with results and analysis of EBR-II dry critical and ZPR-III mock-up experiments. Particular emphasis is given to reactor-physics design problems which arise after the conceptual design is established and before the reactor is built or placed into operation. Reactor-safety analyses and hazards-evaluation considerations are described with their influence on the reactor design. The manner of utilizing the EBR-II mock-up on ZPR-III data and the EBR-II dry critical data is described. These experiments, their analysis and theoretical predictions are the basis for predetermining the physics behaviour of the reactor system. The limitations inherent in applying the experimental data to the performance of the power-reactor system are explored in some detail. This includes the specification of reactor core size and/or fuel-alloy enrichment, provisions for adequate operating and shut-down reactivity, determination of operative temperature and power coefficients of reactivity, and details of power- and flux-distribution as a function of position within the reactor structure. The overall problem of transferring information from simple idealized analytical or experimental geometry to actual hexagonal reactor geometry is described. Nuclear performance, including breeding, of the actual reactor system is compared with that of the idealized conceptual system. The long-term reactivity and power behaviour of the reactor blanket is described within the framework of the proposed cycling of the fuel and blanket alloy. Safety considerations, including normal and abnormal rates of reactivity-insertion, the implication of postulated reactivity effects based on the physical behaviour of the fuel alloy and reactor structure as well as extrapolation of TREAT experiments to the EBR-II system are analysed. The EBR-II core melt-down problem is reviewed. (author

  9. EBR-II: twenty years of operating experience

    International Nuclear Information System (INIS)

    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

  10. EBR-II: search for the lost subassembly

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.; Poloncsik, J.; Remsburg, J.S.; Sine, H.W.

    1983-01-01

    Experimental Breeder Reactor II (EBR-II) has been operating for nearly 20 years as part of the foundation of the US Department of Energy's LMFBR development program. During that time, the EBR-II fuel-handling system has performed extremely well, especially considering the conditions under which much of the system operates and the reliability required to maintain the high plant factor routinely demonstrated by EBR-II. Since EBR-II is a pool-type reactor, much of the fuel handling is done remotely within the sodium-filled primary tank at 371 0 C. Activities involved in locating a misplaced fuel subassembly in the primary tank are described

  11. The physics design of EBR-II

    International Nuclear Information System (INIS)

    Loewenstein, W.B.

    1962-01-01

    The physics design oi EBR-II. Calculations of the static, dynamic and long-term reactivity behaviour of EBR-II are reported together with results and analysis of EBR-II dry critical and ZPR-III mock-up experiments. Particular emphasis is given to reactor-physics design problems which arise after the conceptual design is established and before the reactor is built or placed into operation. Reactor-safety analyses and hazards-evaluation considerations are described with their influence on the reactor design. The manner of utilizing the EBR-II mock-up on ZPR-III data and the EBR-II dry critical data is described. These experiments, their analysis and theoretical predictions are the basis for predetermining the physics behaviour of the reactor system. The limitations inherent in applying the experimental data to the performance of the power-reactor system are explored in some detail. This includes the specification of reactor core size and/or fuel-alloy enrichment, provisions for adequate operating and shut-down reactivity, determination of operative temperature and power coefficients of reactivity, and details of power- and flux-distribution as a function of position within the reactor structure. The overall problem of transferring information from simple idealized analytical or experimental geometry to actual hexagonal reactor geometry is described. Nuclear performance, including breeding, of the actual reactor system is compared with that of the idealized conceptual system. The long-term reactivity and power behaviour of the reactor blanket is described within the framework of the proposed cycling of the fuel and blanket alloy. Safety considerations, including normal and abnormal rates of reactivity-insertion, the implication of postulated reactivity effects based on the physical behaviour of the fuel alloy and reactor structure as well as extrapolation of TREAT experiments to the EBR-II system are analysed. The EBR-II core melt-down problem is reviewed. (author

  12. Experience with advanced driver fuels in EBR-II

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

  13. The EBR-II fuel cycle story

    International Nuclear Information System (INIS)

    Stevenson, C.E.

    1987-01-01

    This volume on the history of the Experimental Breeder Reactor (EBR) program and the Fuel Cycle Facility (FCF) offers both the historical perspective and ''reasons why'' the project was so successful. The operation of the FCF in conjunction with the EBR-II was prepared because of the unique nature of the pyrmetallurgical processing system that was demonstrated at the time. Following brief descriptions and histories of the EBR-I and EBR-II reactors, the FCF and its process requirements are described. The seven principal process steps are presented, including for each one, the development, equipment used, operating procedures, results, problems and other data. Scrap and waste disposition, analytical control, safety, management, and cost of the FCF are also included

  14. EBR-II high-ramp transients under computer control

    International Nuclear Information System (INIS)

    Forrester, R.J.; Larson, H.A.; Christensen, L.J.; Booty, W.F.; Dean, E.M.

    1983-01-01

    During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients

  15. Experience with lifetime limits for EBR-II core components

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Smith, R.N.; Golden, G.H.

    1987-01-01

    The Experimental Breeder Reactor No. 2 (EBR-II) is operated for the US Department of Energy by Argonne National Laboratory and is located on the Idaho National Engineering Laboratory where most types of American reactor were originally tested. EBR-II is a complete electricity-producing power plant now in its twenty-fourth year of successful operation. During this long history the reactor has had several concurrent missions, such as demonstration of a closed Liquid-Metal Reactor (LMR) fuel cycle (1964-69); as a steady-state irradiation facility for fuels and materials (1970 onwards); for investigating effects of operational transients on fuel elements (from 1981); for research into the inherent safety aspects of metal-fueled LMR's (from 1983); and, most recently, for demonstration of the Integral Fast Reactor (IFR) concept using U-Pu-Zr fuels. This paper describes experience gained at EBR-II in defining lifetime limits for LMR core components, particularly fuel elements

  16. Deactivation of the EBR-II complex

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Earle, O.K.; Henslee, S.P.; Wells, P.B.; Zahn, T.P.

    1996-01-01

    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 and 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 and D plan has necessitated this current action

  17. Deactivation of the EBR-II complex

    Energy Technology Data Exchange (ETDEWEB)

    Michelbacher, J A; Earle, O K; Henslee, S P; Wells, P B; Zahn, T P

    1996-01-01

    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 and 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 and D plan has necessitated this current action.

  18. Operating experience of the EBR-II steam generating system

    International Nuclear Information System (INIS)

    Buschman, H.W.; Penney, W.H.; Quilici, M.D.; Radtke, W.H.

    1981-01-01

    The Experimental Breeder Reactor II (EBR-II) is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C (820 F) and 8.62 MPa (1250 psi). The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. Safety and reliability are maximized by using duplex tubes and tubesheets. The performance of the system has been excellent and essentially trouble free. The operating experience of EBR-II provides confidence that the technology can be applied to commercial LMFBR's for an abundant supply of energy for the future. 5 refs

  19. The EBR-II steam generating system - operation, maintenance, and inspection

    International Nuclear Information System (INIS)

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

    2002-01-01

    The Experimental Breeder Reactor II (EBR-II) has operated for 20 years at the Idaho National Engineering Laboratory near Idaho Falls. EBR-II is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. EBR-II has operated at a capacity factor over 70% in the past few years. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C and 8.62 MPa. The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. During the 20 years of operation, components of the steam generator have been subjected to a variety of inspections including visual, dimensional, and ultrasonic. One superheater was removed from service because of anomalous performance and was replaced with an evaporator which was removed, examined, and converted into a superheater. Overall operating experience of the system has been excellent and essentially trouble free. Inspections have not revealed any conditions that are performance or life limiting. (author)

  20. LMFBR operational safety: the EBR-II experience

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

  1. Deactivation of the EBR-II complex

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  2. Deactivation of the EBR-II complex

    International Nuclear Information System (INIS)

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

    1997-01-01

    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 ampersand 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 ampersand 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 solderclose quotes components in place, making future operation or removal unfeasible

  3. Characterization of spent EBR-II driver fuel

    International Nuclear Information System (INIS)

    McKnight, R. D.

    1998-01-01

    Operations and material control and accountancy requirements for the Fuel Conditioning Facility demand accurate prediction of the mass flow of spent EBR-II driver fuel into the facility. This requires validated calculational tools that can predict the burnup and isotopic distribution in irradiated Zr-alloy fueled driver assemblies. Detailed core-follow depletion calculations have been performed for an extensive series of EBR-II runs to produce a database of material inventories for the spent fuel to be processed. As this fuel is processed, comparison of calculated values with measured data obtained from samples of this fuel is producing a growing set of validation data. A more extensive set of samples and measurements from the initial processing of irradiated driver fuel has produced valuable estimates of the biases and uncertainties in both the measured and calculated values. Results of these comparisons are presented herein and indicate the calculated values adequately predict the mass flows

  4. Irradiation of microphones in the EBR-II core

    International Nuclear Information System (INIS)

    Gavin, A.P.; Anderson, T.T.; Bobis, J.P.

    1976-06-01

    Six ANL developed high temperature microphone (acoustic detectors) have been exposed in flowing sodium in the In-Core Instrument Test Facility (INCOT) in the Experimental Breeder Reactor-II (EBR-II) for seven months without any indications of serious degradation of signal output due to the exposure. The YY05 experiment (EBR-II INCOT experiment designation) was performed to obtain data which would be useful in evaluating the ability of the microphones whose active elements are lithium niobate to serve as sensors for acoustic surveillance of fast breeder reactors. The reactor was at full power for 136 days of the experiment exposure period. The microphone temperatures varied from 371 0 C (700 0 F) to 621 0 C (1150 0 F). Neutron exposure varied from 2.64 x 10 22 nvt for the microphone at the elevation of the bottom of the EBR-II core to 0.24 x 10 22 nvt for the microphone at the elevation of the top of an EBR-II fuel assembly. The maximum gamma dose was 5 x 10 12 rads

  5. System modeling and simulation at EBR-II

    International Nuclear Information System (INIS)

    Dean, E.M.; Lehto, W.K.; Larson, H.A.

    1986-01-01

    The codes being developed and verified using EBR-II data are the NATDEMO, DSNP and CSYRED. NATDEMO is a variation of the Westinghouse DEMO code coupled to the NATCON code previously used to simulate perturbations of reactor flow and inlet temperature and loss-of-flow transients leading to natural convection in EBR-II. CSYRED uses the Continuous System Modeling Program (CSMP) to simulate the EBR-II core, including power, temperature, control-rod movement reactivity effects and flow and is used primarily to model reactivity induced power transients. The Dynamic Simulator for Nuclear Power Plants (DSNP) allows a whole plant, thermal-hydraulic simulation using specific component and system models called from libraries. It has been used to simulate flow coastdown transients, reactivity insertion events and balance-of-plant perturbations

  6. Evidence of fast non-linear feedback in EBR-II rod-drop measurements

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1987-06-01

    Feedback reactivities determine the time dependence of a reactor during and after a transient initiating event. Recent analysis of control-rod drops in the Experimental Breeder Reactor II (EBR-II) Reactor has indicated that some relatively fast feedback may exist which cannot be accounted for by the linear feedback mechanisms. The linear and deduced non-linear feedback reactivities from a control-rod drop in EBR-II run 93A using detailed temperature coefficients of reactivity in the EROS kinetics code have been reported. The transient analyses have now been examined in more detail for times close to the drop to ascertain if additional positive reactivity is being built-in early in the drop which could be gradually released later in the drop

  7. Demonstration of passive safety features in EBR-II

    International Nuclear Information System (INIS)

    Planchon, H.P. Jr.; Golden, G.H.; Sackett, J.I.

    1987-01-01

    Two tests of great importance to the design of future commercial nuclear power plants were carried out in the Experimental Breeder Reactor-II on April 3, 1986. These tests, (viewed by about 60 visitors, including 13 foreign LMR specialists) were a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. In these tests, inherent feedback shut the reactor down without damage to the fuel or other reactor components. This resulted primarily from advantageous characteristics of the metal driver fuel used in EBR-II. Work is currently underway at EBR-II to develop a control strategy that promotes inherent safety characteristics, including survivability of transient overpower accidents. In parallel, work is underway at EBR-II on the development of state-of-the-art plant diagnostic techniques

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

    International Nuclear Information System (INIS)

    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. Planning for closure and deactivation of the EBR-II complex

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; Poland, H.F.; Wells, P.B.

    1997-01-01

    In January 1994, DOE terminated the Integral Fast Reactor (IFR) Program. Argonne National Laboratory-West (ANL-W) prepared a detailed plan to put Experimental Breeder Reactor-II (EBR-II) in a safe condition, including removal of irradiated fueled subassemblies from the plant, transfer of subassemblies, and removal and stabilization of primary and secondary sodium liquid heat transfer metal. The goal of deactivation is to stabilize the EBR-II complex until decontamination and decommissioning (D ampersand D) is implemented, thereby minimizing maintenance and surveillance. Deactivation of a sodium cooled reactor presents unique concerns. Residual sodium in the primary and secondary systems must be either reacted or inerted to preclude concerns with explosive sodium-air reactions. Also, residual sodium on components will effectively solder these items in place, making removal unfeasible. Several special cases reside in the primary system, including primary cold traps, a cesium trap, a cover gas condenser, and systems containing sodium-potassium alloy. The sodium or sodium-potassium alloy in these components must be reacted in place or the components removed. The Sodium Components Maintenance Shop at ANL-W provides the capability for washing primary components, removing residual quantities of sodium while providing some decontamination capacity. Considerations need to be given to component removal necessary for providing access to primary tank internals for D ampersand D activities, removal of hazardous materials, and removal of stored energy sources. ANL-W's plan for the deactivation of EBR-II addresses these issues, providing for an industrially and radiologically safe complex, requiring minimal surveillance during the interim period between deactivation and D ampersand D. Throughout the deactivation and closure of the EBR-II complex, federal environmental concerns will be addressed, including obtaining the proper permits for facility condition and waste processing

  10. Data handling at EBR-II [Experimental Breeder Reactor II] for advanced diagnostics and control work

    International Nuclear Information System (INIS)

    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

  11. Safety and operating experience at EBR-II: lessons for the future

    International Nuclear Information System (INIS)

    Sackett, J.I.; Golden, G.H.

    1981-01-01

    EBR-II is a small LMFBR power plant that has performed safely and reliably for 16 years. Much has been learned from operating it to facilitate the design, licensing, and operation of large commercial LMFBR power plants in the US. EBR-II has been found relatively easy to keep in conformity with evolving safety requirements, largely because of inherent safety features of the plant. Such features reduce dependence on active safety systems to protect against accidents. EBR-II has experienced a number of plant-transient incidents, some planned, others inadvertent; none has resulted in any significant plant damage. The operating experience with EBR-II has led to the formulation of an Operational Reliability Test Program (ORTP), aimed at showing inherently safe performance of fuel and plant systems

  12. Pattern-recognition system application to EBR-II plant-life extension

    International Nuclear Information System (INIS)

    King, R.W.; Radtke, W.H.; Mott, J.E.

    1988-01-01

    A computer-based pattern-recognition system, the System State Analyzer (SSA), is being used as part of the EBR-II plant-life extension program for detection of degradation and other abnormalities in plant systems. The SSA is used for surveillance of the EBR-II primary system instrumentation, primary sodium pumps, and plant heat balances. Early results of this surveillance indicate that the SSA can detect instrumentation degradation and system performance degradation over varying time intervals, and can provide derived signal values to replace signals from failed critical sensors. These results are being used in planning for extended-life operation of EBR-II

  13. Component configuration control system development at EBR-II

    International Nuclear Information System (INIS)

    Monson, L.R.; Stratton, R.C.

    1984-01-01

    One ofthe major programs being pursued by the EBR-II Division of Argonne National Laboratory is to improve the reliability of plant control and protection systems. This effort involves looking closely at the present state of the art and needs associated with plant diagnostic, control and protection systems. One of the areas of development at EBR-II involves a component configuration control system (CCCS). This system is a computerized control and planning aid for the nuclear power operator

  14. Current status of experimental breeder reactor-II [EBR-II] shutdown planning

    International Nuclear Information System (INIS)

    McDermott, M. D.; Griffin, C. D.; Michelbacher, J. A.; Earle, O. K.

    2000-01-01

    The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad

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

    International Nuclear Information System (INIS)

    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 -6 yr -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 -6 yr -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

  16. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  17. Recent operating experiences and programs at EBR-II

    International Nuclear Information System (INIS)

    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

  18. SASSYS validation with the EBR-II shutdown heat removal tests

    International Nuclear Information System (INIS)

    Herzog, J.P.

    1989-01-01

    SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

  19. Studies related to emergency decay heat removal in EBR-II

    International Nuclear Information System (INIS)

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

    1979-01-01

    Experimental and analytical studies related to emergency decay heat removal by natural circulation in the EBR-II heat transport circuits are described. Three general categories of natural circulation plant transients are discussed and the resultant reactor flow and temperature response to these events are presented. these categories include the following: (1) loss of forced flow from decay power and low initial flow rates; (2) reactor scram with a delayed loss of forced flow; and (3) loss of forced flow with a plant protective system activated scram. In all cases, the transition from forced to natural convective flow was smooth and the peak in-core temperature rises were small to moderate. Comparisons between experimental measurements in EBR-II and analytical predictions of the NATDEMO code are included

  20. Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

    International Nuclear Information System (INIS)

    Smith, M.S.; Ragland, W.A.

    1989-01-01

    The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

  1. EBR-II operating experience

    International Nuclear Information System (INIS)

    Smith, C.R.F.

    1978-07-01

    Operation of the EBR-2 reactor is presented concerning the performance of the heat removal system; reactor materials; fuel handling system; sodium purification and sampling system; cover-gas purification; plant diagnostics and instrumentation; recent improvements in identifying fission product sources in EBR-2; and EBR-2 safety

  2. Tightly coupled transient analysis of EBR-II: An INEL [Idaho National Engineering Laboratory] Engineering Simulation Center Project

    International Nuclear Information System (INIS)

    Makowitz, H.; Barber, D.G.; Dean, E.M.

    1989-01-01

    A ''Tightly Coupled'' transient analysis system for the Experimental Breeder Reactor-II (FBR-II) is presently under development. The system consists of a faster-than-real-time high fidelity reactor simulation, advanced graphics displays, expert system coupling, and real-time data coupling via the EBR-II data acquisition system to and from the plant and the control system. The first generation software has been developed and tested. Various subsystem couplings and the total system integration have been checked out. A ''Lightly Coupled'' EBR-II reactor startup was conducted in August of 1988 as a demonstration of the system. This system should enhance the diagnostic and prognostic capability of EBR-II in the near term and provide automatic control during startup and power maneuvering in the future, as well as serve as a testbed for new control system development for advanced reactors. 8 refs., 7 figs., 1 tab

  3. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

  4. Review process and quality assurance in the EBR-II probabilistic risk assessment

    International Nuclear Information System (INIS)

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

    1992-01-01

    A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A reactor, has recently been completed at Argonne National Laboratory (ANL). Within the scope of the ANL QA Programs, a QA Plan specifically for the EBR-II PRA was developed. The QA Plan covered all aspects of the PRA development, with emphasis on the procedures for document and software control, and the internal and external review process. The effort spent in the quality assurance tasks for the EBR-II PRA has reciprocated by providing acceptance of the work and confidence in the quality of the results

  5. Analysis of carbon transport in the EBR-II and FFTF primary sodium systems

    International Nuclear Information System (INIS)

    Snyder, R.B.; Natesan, K.; Kassner, T.F.

    1976-01-01

    An analysis of the carburization-decarburization behavior of austenitic stainless steels in the primary heat-transport systems of the EBR-II and FFTF has been made that is based upon a kinetic model for the diffusion process and the surface area of steel in contact with flowing sodium at various temperatures in the two systems. The analysis was performed for operating conditions that result in sodium outlet temperatures of 474 and 566 0 C in the FFTF and 470 0 C in the EBR-II. If there was no external source of carbon to the system, i.e., other than the carbon initially present in the steel and the sodium, the dynamic-equilibrium carbon concentrations calculated for the FFTF primary sodium were approximately 0.025 and approximately 0.065 ppm for the 474 and 566 0 C outlet temperatures, respectively, and approximately 0.018 ppm for the EBR-II primary system. The analysis indicated that a carbon-source rate of approximately 250 g/y would be required to increase the carbon concentration of the EBR-II sodium to the measured range of approximately 0.16--0.19 ppm. An evaluation of possible carbon sources and the amount of carbonaceous material introduced into the reactor cover gas and sodium suggests that the magnitude of the calculated contamination rate is reasonable. For a 566 0 C outlet temperature, carbonaceous material would have to be introduced into the FFTF primary system at a rate approximately 4--6 times higher than in EBR-II to achieve the same carbon concentration in the sodium in the two systems. Since contamination rates of approximately 1500 g/y are unlikely, high-temperature fuel cladding in the FFTF should exhibit decarburization similar to that observed in laboratory loop systems, in contrast to the minimal compositional changes that result after exposure of Type 316 stainless steel to EBR-II sodium at temperatures between approximately 625 and 650 0 C

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    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

  8. EBR-II experience with sodium cleaning and radioactivity decontamination

    International Nuclear Information System (INIS)

    Ruther, W.E.; Smith, C.R.F.

    1978-01-01

    The EBR-II is now in Its 13th year of operation. During that period more than 2400 subassemblies have been cleaned of sodium without a serious incident of any kind by a two-step process developed at Argonne. Sodium cleaning and decontamination of other reactor components has been performed only on the relatively few occasions in which a repair or replacement has been required. A summary of the EBR-II experience will be presented. A new facility will be described for the improved cleaning and maintenance of sodium-wetted primary components

  9. EBR-II experience with sodium cleaning and radioactivity decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Ruther, W E; Smith, C R.F. [Argonne National Laboratory, Argonne (United States)

    1978-08-01

    The EBR-II is now in Its 13th year of operation. During that period more than 2400 subassemblies have been cleaned of sodium without a serious incident of any kind by a two-step process developed at Argonne. Sodium cleaning and decontamination of other reactor components has been performed only on the relatively few occasions in which a repair or replacement has been required. A summary of the EBR-II experience will be presented. A new facility will be described for the improved cleaning and maintenance of sodium-wetted primary components.

  10. Experience with advanced driver fuels in EBR-II

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

    1992-01-01

    This paper discusses several metallic fuel element designs which have been tested and used as driver fuel in Experimental Breeder Reactor II (EBR-II). The most recent advanced designs have all performed acceptably in EBR-H and can provide reliable performance to high burnups. Fuel elements tested have included use of U-l0Zr metallic fuel with either D9, 316 or HT9 stainless steel cladding; the D9 and 316-clad designs have been used as standard driver fuel. Experimental data indicate that fuel performance characteristics are very similar for the various designs tested. Cladding materials can be selected that optimize performance based on reactor design and operational goals

  11. Nuclear instrumentation system operating experience and nuclear instrument testing in the EBR-II

    International Nuclear Information System (INIS)

    Yingling, G.E.; Curran, R.N.

    1980-01-01

    In March of 1972 three wide range nuclear channels were purchased from Gulf Atomics Corporation and installed in EBR-II as a test. The three channels were operated as a test until April 1975 when they became a permanent part of the reactor shutdown system. Also described are the activities involved in evaluating and qualifying neutron detectors for LMFBR applications. Included are descriptions of the ANL Components Technology Division Test Program and the EBR-II Nuclear Instrument Test Facilities (NITF) used for the in-reactor testing and a summary of program test results from EBR-II

  12. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    International Nuclear Information System (INIS)

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

    1998-01-01

    This paper summarizes the results from the 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 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -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

  13. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

    This paper summarizes the results from the 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 -6} yr{sup -1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10{sup -6} yr{sup -1} and the contribution of seismic events is 1.7 10{sup -5} yr{sup -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.

  14. The EBR-II probabilistic risk assessment lessons learned regarding passive safety

    International Nuclear Information System (INIS)

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

    1994-01-01

    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 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -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

  15. Operating and test experience of EBR-II

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1991-01-01

    EBR-2 has operated for 27 years, the longest for any Liquid Metal Reactor (LMR) power plant. During that time, much has been learned about successful LMR operation and design. The basic lesson is that conversatism in design can pay significant dividends in operating reliability. Furthermore, such conservatism need not mean high cost. The EBR-2 system emphasizes simplicity, minimizing the number of valves in the heat transport system, for example, and simplifying the primary heat-transport-system layout. Another lesson is that emphasizing reliability of the steam generating system at the sodium-water interface (by using duplex tubes in the case of EBR-2) has been well worth the higher initial costs; no problems with leakage have been encountered in EBR-2's operating history. Locating spent fuel storage in the primary tank and providing for decay heat removal by natural connective flow have also been contributors to EBR-2's success. The ability to accommodate loss of forced cooling or loss of heat sink passively has resulted in benefits for simplification, primarily through less reliance on emergency power and in not requiring the secondary sodium or steam systems to be safety grade. Also, the ''piped-pool '' arrangement minimizes thermal stress to the primary tank and enhances natural convective flow. These benefits have been realized through a history of operation that has seen EBR-2 evolve through four major phases in its test programs, culminating in its present mission as the Integral Fast Rector (IFR) prototype. 20 refs., 8 figs., 1 tab

  16. A transient overpower experiment in EBR-II

    International Nuclear Information System (INIS)

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

    1994-01-01

    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

  17. Off-normal performance of EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    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

  18. Proof tests of irradiated and unirradiated EBR-II subassembly ducts

    International Nuclear Information System (INIS)

    Ruther, W.E.; Chopra, P.S.; Lambert, J.D.B.

    1977-01-01

    A series of dynamic pressure tests have been conducted within EBR-II subassembly ducts. The tests were designed to simulate bursting of a driver-fuel element in a cluster of such elements at their burnup limit during off-normal conditions in EBR-II. The major objective of the tests was to assure that such failure, which might cause rapid release of stored fission gas, would not deform or otherwise damage subassembly ducts in a way that would hinder movement of a control rod. The test results are described

  19. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR (Integral Fast Reactor) program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. (author)

  20. Potential safety enhancements to nuclear plant control: proof testing at EBR-II

    International Nuclear Information System (INIS)

    Lindsay, R.W.; Chisholm, G.H.

    1984-01-01

    Future changes in nuclear plant control and protective systems will reflect an evolutionary improvement through increased use of computers coupled with a better integration of man and machine. Before improvements can be accepted into the licensed commercial plant environment, significant testing must be accomplished to answer safety questions and to prove the worth of new ideas. The Experimental Breeder Reactor-II (EBR-II) is being used as a test-bed for both in-house development and testing for others in a DOE sponsored Man-Machine Integration program. The ultimate result of the development and testing would be a control system for which safety credit could be taken in the licensing process

  1. Experience with automatic reactor control at EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.; Larson, H.A.; Christensen, L.J.

    1985-01-01

    Satisfactory operation of the ACRDS has extended the capabilities of EBR-II to a transient test facility, achieving automatic transient control. Test assemblies can now be irradiated in transient conditions overlapping the slower transient capability of the TREAT reactor

  2. Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

    International Nuclear Information System (INIS)

    Eberle, C.S.; Dean, E.M.; Angelo, P.L.

    1995-01-01

    A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

  3. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

  4. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    International Nuclear Information System (INIS)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

    2016-01-01

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

  5. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  6. Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1989-01-01

    Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

    1992-01-01

    The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

  9. SASSYS-1 computer code verification with EBR-II test data

    International Nuclear Information System (INIS)

    Warinner, D.K.; Dunn, F.E.

    1985-01-01

    The EBR-II natural circulation experiment, XX08 Test 8A, is simulated with the SASSYS-1 computer code and the results for the latter are compared with published data taken during the transient at selected points in the core. The SASSYS-1 results provide transient temperature and flow responses for all points of interest simultaneously during one run, once such basic parameters as pipe sizes, initial core flows, and elevations are specified. The SASSYS-1 simulation results for the EBR-II experiment XX08 Test 8A, conducted in March 1979, are within the published plant data uncertainties and, thereby, serve as a partial verification/validation of the SASSYS-1 code

  10. Simulation and verification of the EBR-II automatic control rod drive system with continuous system modeling codes

    International Nuclear Information System (INIS)

    Larson, H.A.; Dean, E.M.

    1985-01-01

    The two computer programs are successful in modeling the EBR-II ACRDS. In fact, this is very convenient for a presampling of the consequences of a desired power movement. The ACRDS is to be modified so that the error signal is a comparison between demand position and measured position. Purpose of this change is to permit pseudo-random binary types of reactivity transfer function experiments at EBR-II. Questions asked about the computer software and hardware to accommodate this change can be quickly answered with either of the verified codes discussed here

  11. Operating limits for subassembly deformation in EBR-II

    International Nuclear Information System (INIS)

    Bottcher, J.H.

    1977-01-01

    The deformation of a subassembly in response to the core environment is frequently the life limiting factor for that component in an LMFBR. Deformation can occur as diametral and axial growth or bowing of the subassembly. Such deformation has caused several handling problems in both the core and the storage basket of EBR-II and may also have contributed to reactivity anomalies during reactor operation. These problems generally affect plant availability but the reactivity anomalies could lead to a potential safety hazard. Because of these effects the deformation mechanisms must be understood and modeled. Diametral and axial growth of subassembly ducts in EBR-II is due to swelling and creep and is a function of temperature, neutron fluence and stress. The source of stress in a duct is the hydraulic pressure difference across the wall. By coupling the calculated subassembly growth rate to the available clearance in the core or storage basket a limiting neutron fluence, or exposure, can be established

  12. Water treatment in the EBR-II steam system

    International Nuclear Information System (INIS)

    Klein, M.A.; Hurst, H.

    1975-01-01

    Boiler-water treatment in the EBR-II steam system consists of demineralizing makeup water and using hydrazine to remove traces of oxygen and morpholine to adjust pH to 8.8-9.2. This treatment is called a ''zero-solids'' method, because the chemical agents and reaction products are either volatile or form water and do not contribute solids to the boiler water. A continuous blowdown is cooled, filtered, and deionized to remove impurities and maintain high purity of the water. If a cooling-water leak occurs, phosphate is added to control scaling, and the ''zero-solids'' eatment is suspended until the leak is repaired. Water streams are sampled at six points to control water purity. Examination of the steam drum and an evaporator show the metal surfaces to be in excellent condition with minimal corrosion. The EBR-II steam-generating plant has accumulated over 85,000 hours of in-service operation and has operated successfully for over ten years with the ''zero-solids'' treatment. (auth)

  13. Microstructural comparison of HT-9 irradiated in HFIR and EBR-II

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1985-05-01

    A series of specimens of HT-9 heat 91354 have been examined following irradiation in HFIR to 39 dpa at 300, 400, 500 and 600 0 C and following irradiation in EBR-II to 29 dpa at 390 and 500 0 C. HFIR irradiation was found to have promoted helium bubble formation at all temperatures and voids at 400 0 C. Cavitation had not been observed at lower fluence, nor was it found in EBR-II irradiated specimens. The onset of void swelling in HFIR is attributed to helium generation. The observations provide an explanation for saturation of ductile-brittle transition temperature shifts with increasing fluence

  14. Advances in criticality predictions for EBR-II

    International Nuclear Information System (INIS)

    Schaefer, R.W.; Imel, G.R.

    1994-01-01

    Improvements to startup criticality predictions for the EBR-II reactor have been made. More exact calculational models, methods and data are now used, and better procedures for obtaining experimental data that enter into the prediction are in place. Accuracy improved by more than a factor of two and the largest ECP error observed since the changes is only 18 cents. An experimental method using subcritical counts is also being implemented

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

    International Nuclear Information System (INIS)

    Demmer, R.; Heintzelman, J.; Squires, L.; Meservey, R.

    2009-01-01

    The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. 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. 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. 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 about 100,000 gallons of wash water. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to evaluate 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 workshop was conducted by a trained facilitator using Value Engineering techniques to elicit the most technically sound solutions from the workshop participants. A brainstorming session was held to identify possible alternative treatment methods that would meet the primary functions and criteria of neutralizing the hazards, maximizing byproduct removal and minimizing waste generation. An initial list of some 20 probable alternatives was evaluated and refined down

  16. Sodium technology at EBR-II

    International Nuclear Information System (INIS)

    Holmes, J.T.; Smith, C.R.F.; Olson, W.H.

    1976-01-01

    Since the installation of purity monitoring systems in 1967, the control of the purity of the primary and secondary sodium and cover gas systems at the Experimental Breeder Reactor II (EBR-II) has been excellent. A rigorous monitoring program is being used to assure that operating limits for more than 25 chemical and radioactive impurities are not exceeded. The program involves the use of sophisticated sampling and analysis techniques and on-line monitors for both sodium and cover gas systems. Sodium purity control is accomplished by essentially continuous cold trapping of a small side stream of the total circulating sodium. The cold traps have been found to be very effective for the removal of the major chemical impurities (oxygen and hydrogen) and tritium but are almost ineffective for 131 I and 137 Cs that enter the sodium from fuel cladding breaks. Purging with pure argon maintains the cover gas purity

  17. Assessment calculation of MARS-LMR using EBR-II SHRT-45R

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C.; Ha, K.S.

    2016-10-15

    Highlights: • Neutronic and thermal-hydraulic behavior predicted by MARS-LMR is validated with EBR-II SHRT-45R test data. • Decay heat model of ANS-94 give better prediction of the fission power. • The core power is well predicted by reactivity feedback during initial transient, however, the predicted power after approximately 200 s is over-estimated. The study of the reactivity feedback model of the EBR-II is necessary for the better calculation of the power. • Heat transfer between inter-subassemblies is the most important parameter, especially, a low flow and power subassembly, like non-fueled subassembly. - Abstract: KAERI has designed a prototype Gen-IV SFR (PGSFR) with metallic fuel. And the safety analysis code for the PGSFR, MARS-LMR, is based on the MARS code, and supplemented with various liquid metal related features including sodium properties, heat transfer, pressure drop, and reactivity feedback models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT)”. Argonne National Laboratory (ANL) has technically supported and participated in this program. One of benchmark analysis tests is SHRT-45R, which is an unprotected loss of flow test in an EBR-II. So, sodium natural circulation and reactivity feedbacks are major phenomena of interest. A benchmark analysis was conducted using MARS-LMR with original input data provided by ANL. MARS-LMR well predicts the core flow and power change by reactivity feedbacks in the core. Except the results of the XX10, the temperature and flow in the XX09 agreed well with the experiments. Moreover, sensitivity tests were carried out for a decay heat model, reactivity feedback model, inter-subassembly heat transfer, internal heat structures and so on, to evaluate their sensitivity and get a better prediction. The decay heat model of ANS-94 shows

  18. Assessment calculation of MARS-LMR using EBR-II SHRT-45R

    International Nuclear Information System (INIS)

    Choi, C.; Ha, K.S.

    2016-01-01

    Highlights: • Neutronic and thermal-hydraulic behavior predicted by MARS-LMR is validated with EBR-II SHRT-45R test data. • Decay heat model of ANS-94 give better prediction of the fission power. • The core power is well predicted by reactivity feedback during initial transient, however, the predicted power after approximately 200 s is over-estimated. The study of the reactivity feedback model of the EBR-II is necessary for the better calculation of the power. • Heat transfer between inter-subassemblies is the most important parameter, especially, a low flow and power subassembly, like non-fueled subassembly. - Abstract: KAERI has designed a prototype Gen-IV SFR (PGSFR) with metallic fuel. And the safety analysis code for the PGSFR, MARS-LMR, is based on the MARS code, and supplemented with various liquid metal related features including sodium properties, heat transfer, pressure drop, and reactivity feedback models. In order to validate the newly developed MARS-LMR, KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT)”. Argonne National Laboratory (ANL) has technically supported and participated in this program. One of benchmark analysis tests is SHRT-45R, which is an unprotected loss of flow test in an EBR-II. So, sodium natural circulation and reactivity feedbacks are major phenomena of interest. A benchmark analysis was conducted using MARS-LMR with original input data provided by ANL. MARS-LMR well predicts the core flow and power change by reactivity feedbacks in the core. Except the results of the XX10, the temperature and flow in the XX09 agreed well with the experiments. Moreover, sensitivity tests were carried out for a decay heat model, reactivity feedback model, inter-subassembly heat transfer, internal heat structures and so on, to evaluate their sensitivity and get a better prediction. The decay heat model of ANS-94 shows

  19. An evaluation of multigroup flux predictions in the EBR-II core

    International Nuclear Information System (INIS)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required

  20. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-12-31

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

  1. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Performance of advanced oxide fuel pins in EBR-II

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

    1986-05-01

    The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

  4. Safety aspects of advanced fuels irradiations in EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.

    1975-09-01

    Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

  5. EBR-II rotating plug seal maintenance

    International Nuclear Information System (INIS)

    Allen, K.J.

    1986-01-01

    The EBR-II rotating plug seals require frequent cleaning and maintenance to keep the plugs from sticking during fuel handling. Time consuming cleaning on the cover gas and air sides of the dip ring seal is required to remove oxidation and sodium reaction products that accumulate and stop plug rotation. Despite severely limited access, effective seal cleaning techniques have removed 11 800 lb (5 352 kg) of deposits from the seals since 1964. Temperature control modifications and repairs have also required major maintenance work. Suggested seal design recommendations could significantly reduce maintenance on future similar seals

  6. EBR-II and TREAT Digitization Project

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, George W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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).

  7. EBR-II and TREAT Digitization Project

    International Nuclear Information System (INIS)

    Griffith, George W.; Rabiti, Cristian

    2015-01-01

    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).

  8. EBR-II water-to-sodium leak detection system

    International Nuclear Information System (INIS)

    Wrightson, M.M.; McKinley, K.; Ruther, W.E.; Holmes, J.T.

    1976-01-01

    The water-to-sodium leak detection system installed at EBR-II in April, 1975, is described in detail. Topics covered include operational characteristics, maintenance problems, alarm functions, background hydrogen level data, and future plans for refinements to the system. Particular emphasis is given to the failures of eight of the ten leak detectors due to sodium-to-vacuum leakage, and the program anticipated for complete recovery of the system

  9. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

    1986-01-01

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  10. Simulation and operation of the EBR-II automatic control rod drive system

    International Nuclear Information System (INIS)

    Lehto, W.K.; Larson, H.A.; Dean, E.M.; Christensen, L.J.

    1985-01-01

    An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control-rod-drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE Operational Reliability Testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In addition, the ACRDS is used for steady-state operation and will be qualified to control power ascent from initial critical to full power

  11. Simulation and operation of the EBR-II automatic control rod drive system

    International Nuclear Information System (INIS)

    Lehto, W.K.; Larson, H.A.; Dean, E.M.; Christensen, L.J.

    1985-01-01

    An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control-rod-drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE Operational Reliability Testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In additions, the ACRDS is used for steady-state operation and will be qualified to control power ascent from initial critical to full power

  12. Comparisons of PRD [power-reactivity-decrements] components for various EBR-II configurations

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1986-01-01

    Comparison of detailed calculations of contributions by region and component of the power-reactivity-decrements (PRD) for four differing loading configurations of the Experimental Breeder Reactor-II (EBR-II) are given. The linear components and Doppler components are calculated. The non-linear (primarily subassembly bowing) components are deduced by differences relative to measured total PRD values. Variations in linear components range from about 10% to as much as about 100% depending upon the component. The deduced non-linear components differ both in magnitude and sign as functions of reactor power. Effects of differing assumptions of the nature of the fuel-to-clad interactions upon the PRD components are also calculated

  13. Breached fuel pin contamination from Run Beyond Cladding Breach (RBCB) tests in EBR-II

    International Nuclear Information System (INIS)

    Colburn, R.P.; Strain, R.V.; Lambert, J.D.B.; Ukai, S.; Shibahara, I.

    1988-09-01

    Studies indicate there may be a large economic incentive to permit some continued reactor operation with breached fuel pin cladding. A major concern for this type of operation is the potential spread of contamination in the primary coolant system and its impact on plant maintenance. A study of the release and transport of contamination from naturally breached mixed oxide Liquid Metal Reactor (LMR) fuel pins was performed as part of the US Department of Energy/Power Reactor and Nuclear Fuel Development Corporation (DOE/PNC) Run Beyond Cladding Breach (RBCB) Program at EBR-II. The measurements were made using the Breached Fuel Test Facility (BFTF) at EBR-II with replaceable deposition samplers located approximately 1.5 meters from the breached fuel test assemblies. The effluent from the test assemblies containing the breached fuel pins was routed up through the samplers and past dedicated instrumentation in the BFTF before mixing with the main coolant flow stream. This paper discusses the first three contamination tests in this program. 2 refs., 5 figs., 2 tabs

  14. Results and implications of the EBR-II inherent safety demonstration tests

    International Nuclear Information System (INIS)

    Planchon, H.P.; Golden, G.H.; Sackett, J.I.; Mohr, D.; Chang, L.K.; Feldman, E.E.; Betten, P.R.

    1987-01-01

    On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants

  15. Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques

    International Nuclear Information System (INIS)

    Jacqmin, R.P.; Finck, P.J.; Palmiotti, G.

    1994-01-01

    This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution

  16. Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

    1982-01-01

    LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

  17. Dynamic modeling and simulation of EBR-II steam generator system

    International Nuclear Information System (INIS)

    Berkan, R.C.; Upadhyaya, B.R.

    1989-01-01

    This paper presents a low order dynamic model of the Experimental breeder Reactor-II (EBR-II) steam generator system. The model development includes the application of energy, mass and momentum balance equations in state-space form. The model also includes a three-element controller for the drum water level control problem. The simulation results for low-level perturbations exhibit the inherently stable characteristics of the steam generator. The predictions of test transients also verify the consistency of this low order model

  18. EBR-II fuel handling console digital upgrade

    International Nuclear Information System (INIS)

    Peters, G.G.; Wiege, D.D.; Christensen, L.J.

    1995-01-01

    The main fuel handling console and control system at the Experimental Breeder Reactor II (EBR-II) are being upgraded to a computerized system using high-end workstations for the operator interface and a programmable logic controller (PLC) for the control system. Two-dimensional (2D) and three-dimensional (3D) computer graphics will be provided for the operator which will show the relative position of under-sodium fuel handling equipment. This equipment is operated remotely with no means of directly viewing the transfer. This paper describes various aspects of the modification including reasons for the upgrade, capabilities the new system provides over the old control system, philosophies and rationale behind the new design, testing and simulation work, diagnostic features, and the advanced graphics techniques used to display information to the operator

  19. Reactivity-induced time-dependencies of EBR-II linear and non-linear feedbacks

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1988-01-01

    Time-dependent linear feedback reactivities are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a kinetic code analysis of an experiment in which the change in power resulted from the dropping of a control rod. Shown with these linear reactivities are the reactivity associated with the control-rod shaft contraction and also time-dependent non-linear (mainly bowing) component deduced from the inverse kinetics of the experimentally measured fission power and the calculated linear reactivities. (author)

  20. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1993-01-01

    Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

  1. Operational reliability testing of FBR fuel in EBR-II

    International Nuclear Information System (INIS)

    Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

    1991-01-01

    The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

  2. Parametric investigation of fracture of EBR-II ducts

    International Nuclear Information System (INIS)

    Chopra, P.S.; Moustakakis, B.

    1977-01-01

    Results of preliminary static and dynamic finite element fracture mechanics analyses that were conducted to analytically simulate the dynamic fracture behavior of EBR-II ducts are presented. The loads considered are those that may arise because of rapid release of fission gases from a failed fuel element inside a duct, obtained from some previous tests and a recent analytical model. In spite of the motivation for the present work, the analytical procedures described may have a wider general application in the fail-safe design of structures

  3. Bowing-reactivity trends in EBR-II assuming zero-swelling ducts

    International Nuclear Information System (INIS)

    Meneghetti, D.

    1994-01-01

    Predicted trends of duct-bowing reactivities for the Experimental Breeder Reactor II (EBR-II) are correlated with predicted row-wise duct deflections assuming use of idealized zero-void-swelling subassembly ducts. These assume no irradiation induced swellings of ducts but include estimates of the effects of irradiation-creep relaxation of thermally induced bowing stresses. The results illustrate the manners in which at-power creeps may affect subsequent duct deflections at zero power and thereby the trends of the bowing component of a subsequent power reactivity decrement

  4. Considerations for advanced reactor design based on EBR-II experience

    International Nuclear Information System (INIS)

    King, R. W.

    1999-01-01

    The long-term success of the Experimental Breeder Reactor-II (EBR-II) provides several insights into fundamental characteristics and design features of a nuclear generating station that enhance safety, operability, and maintainability. Some of these same characteristics, together with other features, offer the potential for operational lifetimes well beyond the current licensing time frame, and improved reliability that could potentially reduce amortized capital costs as well as overall operation and maintenance costs if incorporated into advanced plant designs. These features and characteristics are described and the associated benefits are discussed

  5. Transient performance of EBR-II driver fuel

    International Nuclear Information System (INIS)

    Buzzell, J.A.; Hudman, G.D.; Porter, D.L.

    1981-01-01

    The first phases of qualification of the EBR-II driver fuel for repeated transient overpower operation have recently been completed. The accomplishments include prediction of the transient fuel and cladding performance through ex-core testing and fuel-element modeling studies, localized in-core power testing during steady-state operation, and whole-core multiple transient testing. The metallic driver fuel successfully survived 56 transients, spaced over a 45-day period, with power increases of approx. 160% at rates of approx. 1%/s with a 720-second hold at full power. The performance results obtained from both ex-core and n-core tests indicate that the fuel is capable of repeated transient operation

  6. Data systems in FFTF and EBR-II

    International Nuclear Information System (INIS)

    Warrick, R.P.; Ritter, W.M.

    1980-02-01

    This paper describes the Data System used to monitor operation and collect experimental data in FFTF. This data system has evolved since initial inception from a relatively simple, single computer system monitoring a relatively few (approx. 1000) instrument channels important for operation to one which has increased capability to support the long-range testing needs in FFTF. The system, while still relatively simple, now contains multiple computers which normally perform independent functions. The computers, however, provide backup processing for certain simple tasks. Operator interfacing is provided through CRT's. The output capabilities of the system are described. A description of the Data System in EBR-II is also included

  7. Feedback components of a U20Pu10Zr-fueled compared to a U10Zr-fueled EBR-II

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1988-01-01

    Calculated feedback components of the regional contributions of the power reactivity decrements (PRDs) and of the temperature coefficients of reactivity of a U20Pu10Zr-fueled and of a U10Zr-fueled Experimental Breeder Reactor II (EBR-II) are compared. The PRD components are also separated into power-to-flow dependent and solely power dependent parts. The effects of these values upon quantities useful for indicating the comparative potential inherent safety characteristics of these EBR-II loadings are presented

  8. In-reactor cladding breach of EBR-II driver-fuel elements

    International Nuclear Information System (INIS)

    Seidel, B.R.; Einziger, R.E.

    1977-01-01

    Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    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

  10. Modification of EBR-II plant to conduct loss-of-flow-without-scram tests

    Energy Technology Data Exchange (ETDEWEB)

    Messick, N C; Betten, P R; Booty, W F; Christensen, L J; Fryer, R M; Mohr, D; Planchon, H P; Radtke, W H

    1987-04-01

    This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests.

  11. Modification of EBR-II plant to conduct loss-of-flow-without-scram tests

    International Nuclear Information System (INIS)

    Messick, N.C.; Betten, P.R.; Booty, W.F.; Christensen, L.J.; Fryer, R.M.; Mohr, D.; Planchon, H.P.; Radtke, W.H.

    1987-01-01

    This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests. (orig.)

  12. Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Horak, J.A.

    1985-01-01

    Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

  13. Applications of the EBR-II Probabilistic Risk Assessment

    International Nuclear Information System (INIS)

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

    1993-01-01

    A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future

  14. Seismic response of the EBR-II to the Mt. Borah earthquake

    International Nuclear Information System (INIS)

    Gale, J.G.; Lehto, W.K.

    1985-01-01

    On October 28, 1983, an earthquake of magnitude 7.3 occurred in the mountains of central Idaho at a distance of 114-km from the ANL-West site. The earthquake tripped the seismic sensors in the EBR-II reactor shutdown system causing a reactor scram. Visual and operability checks of structures, components, and systems showed no indication of damage or system abnormalities and reactor restart was initiated. As a result of the earthquake, questions arose as to the magnitude of the actual stress levels in critical components and what value of ground acceleration could be experienced without damage to reactor structures. EBR-II was designed prior to implementation of present day requirements for seismic qualification and appropriate analyses had not been conducted. A lumped-mass, finite element model of the primary tank, support structure, and the reactor was generated and analyzed using the response spectrum technique. The analysis showed that the stress levels in the primary tank system were very low during the Mount Borah earthquake and that the system could experience seismic loadings three to four times those of the Mount Borah earthquake without exceeding yield stresses in any of the components

  15. EBR-II argon cooling system restricted fuel handling I and C upgrade

    International Nuclear Information System (INIS)

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

    1995-01-01

    The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  18. Computer imaging of EBR-II handling equipment

    International Nuclear Information System (INIS)

    Hansen, L.H.; Peters, G.G.

    1994-10-01

    This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. The system described in this paper uses actual signals to move a three-dimensional graphics model in real-time in response to movements of equipment in the plant. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. This paper will present details on how the 3D model was created and how real-time dynamic behavior was added to each of the moving components

  19. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  20. Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

  1. System modeling of spent fuel transfers at EBR-II

    International Nuclear Information System (INIS)

    Imel, G.R.; Houshyar, A.

    1994-01-01

    The unloading of spent fuel from the Experimental Breeder Reactor-II (EBR-II) for interim storage and subsequent processing in the Fuel Cycle Facility (FCF) is a multi-stage process, involving complex operations at a minimum of four different facilities at the Argonne National Laboratory-West (ANL-W) site. Each stage typically has complicated handling and/or cooling equipment that must be periodically maintained, leading to both planned and unplanned downtime. A program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. Routine operation of the reactor for fuels performance and materials testing occurred simultaneously in FY 1994 with the blanket unloading. In the summer of 1994, Congress dictated the October 1, 1994 shutdown of EBR-2. Consequently, all blanket S/As and fueled drivers will be removed from the reactor tank and replaced with stainless steel assemblies (which are needed to maintain a precise configuration within the grid so that the under sodium fuel handling equipment can function). A system modeling effort was conducted to determine the means to achieve the objective for the blanket and fuel unloading program, which under the current plan requires complete unloading of the primary tank of all fueled assemblies in 2 1/2 years. A simulation model of the fuel handling system at ANL-W was developed and used to analyze different unloading scenarios; the model has provided valuable information about required resources and modifications to equipment and procedures. This paper reports the results of this modeling effort

  2. Time constants and feedback transfer functions of EBR-II subassembly types

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1986-01-01

    Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

  3. Time constants and feedback transfer functions of EBR-II subassembly types

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1987-01-01

    Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel. (author)

  4. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. 5 refs., 4 figs

  5. EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

    International Nuclear Information System (INIS)

    Mott, J.E.; Radtke, W.H.; King, R.W.

    1986-02-01

    The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

  6. On-line sodium and cover as purity monitors gas operating tools at EBR-II

    International Nuclear Information System (INIS)

    Smith, C.R.F.; Richardson, W.J.; Holmes, J.T.

    1976-01-01

    Plugging temperature indicators, electrochemical oxygen meters and hydrogen diffusion meters are the on-line sodium purity monitors now in use at EBR-II. On-line gas chromatographs are used to monitor helium, hydrogen, oxygen and nitrogen impurities in the argon cover gases. Monitors for tritium-in-sodium and for hydrocarbons-in-cover gas have been developed and are scheduled for installation in the near future. An important advantage of on-line monitors over the conventional grab-sampling techniques is the speed of response to changing reactor conditions. This helps us to identify the source of the impurity, whether the cause may be transient or constant, and take corrective action as necessary. The oxygen meter is calibrated monthly against oxygen in sodium determined by the vanadium wire equilibration method. The other instruments either do not require calibration or are self-calibrating. The ranges, sensitivity and response times of all of the on-line purity monitors has proven satisfactory under EBR-II operating conditions

  7. EBR-II Cover Gas Cleanup System upgrade process control system structure

    International Nuclear Information System (INIS)

    Carlson, R.B.; Staffon, J.D.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) Cover Gas Cleanup System (CGCS) control system was upgraded in 1991 to improve control and provide a graphical operator interface. The upgrade consisted of a main control computer, a distributed control computer, a front end input/output computer, a main graphics interface terminal, and a remote graphics interface terminal. This paper briefly describes the Cover Gas Cleanup System and the overall control system; describes the main control computer hardware and system software features in more detail; and, then, describes the real-time control tasks, and how they interact with each other, and how they interact with the operator interface task

  8. Fail-safety of the EBR-II steam generator system

    International Nuclear Information System (INIS)

    Chopra, P.S.; Stone, C.C.; Hutter, E.; Barney, W.K.; Staker, R.G.

    1976-01-01

    Fail-safe analyses of the EBR-II steam-generator system show that a postulated non-instantaneous leak of water or steam into sodium, through a duplex tube or a tubesheet, at credible leak rates will not structurally damage the evaporators and superheaters. However, contamination of the system and possible shell wastage by sodium-water reaction products may render the system inoperable for a period exceeding six months. This period would be shortened to three months if the system were modified by adding a remotely operated water dump system, a steam vent system, a secondary sodium superheater relief line, and a tubesheet leak-detection system

  9. Automated start-up of EBR-II

    International Nuclear Information System (INIS)

    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

  10. Sensitivity Analysis of Uncertainty Parameter based on MARS-LMR Code on SHRT-45R of EBR II

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Seok-Ju; Kang, Doo-Hyuk; Seo, Jae-Seung [System Engineering and Technology Co., Daejeon (Korea, Republic of); Bae, Sung-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Hae-Yong [Sejong University, Seoul (Korea, Republic of)

    2016-10-15

    In order to assess the uncertainty quantification of the MARS-LMR code, the code has been improved by modifying the source code to accommodate calculation process required for uncertainty quantification. In the present study, a transient of Unprotected Loss of Flow(ULOF) is selected as typical cases of as Anticipated Transient without Scram(ATWS) which belongs to DEC category. The MARS-LMR input generation for EBR II SHRT-45R and execution works are performed by using the PAPIRUS program. The sensitivity analysis is carried out with Uncertainty Parameter of the MARS-LMR code for EBR-II SHRT-45R. Based on the results of sensitivity analysis, dominant parameters with large sensitivity to FoM are picked out. Dominant parameters selected are closely related to the development process of ULOF event.

  11. Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan); Muranaka, Kohmei; Asai, Takayuki [Graduate School of Engineering, University of Fukui (Japan); Rooijen, W.F.G. van, E-mail: rooijen@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan)

    2014-08-15

    Highlights: • The IAEA EBR-II benchmarks SHRT-17 and SHRT-45R are analyzed with a 1D system code. • The calculated result of SHRT-17 corresponds well to the measured results. • For SHRT-45R ERANOS is used for various core parameters and reactivity coefficients. • SHRT-45R peak temperature is overestimated with the ERANOS feedback coefficients. • The peak temperature is well predicted when the feedback coefficient is reduced. - Abstract: Benchmark problems of several experiments in EBR-II, proposed by ANL and coordinated by the IAEA, are analyzed using the plant system code NETFLOW++ and the neutronics code ERANOS. The SHRT-17 test conducted as a loss-of-flow test is calculated using only the NETFLOW++ code because it is a purely thermal–hydraulic problem. The measured data were made available to the benchmark participants after the results of the blind benchmark calculations were submitted. Our work shows that major parameters of the plant are predicted with good accuracy. The SHRT-45R test, an unprotected loss of flow test is calculated using the NETFLOW++ code with the aid of delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to perform a semi-coupled analysis of the neutronics – thermal–hydraulic problem. The measured data are compared with our calculated results. In our work, the peak temperature is underestimated, indicating that the reactivity feedback coefficients are too strong. When the reactivity feedback coefficient for thermal expansion is adjusted, good agreement is obtained in general for the calculated plant parameters, with a few exceptions.

  12. Computer imaging of EBR-II fuel handling equipment

    International Nuclear Information System (INIS)

    Peters, G.G.; Hansen, L.H.

    1995-01-01

    This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. The system described in this paper uses actual signals to drive a three-dimensional computer-generated model in real-time in response to movements of equipment in the plant This paper will present details on how the 3D model of the intank equipment was created and how real-time dynamic behavior was added to each of the moving components

  13. The EBR-II X501 Minor Actinide Burning Experiment

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  15. Development and testing of a diagnostic system for intelligen distributed control at EBR-2

    International Nuclear Information System (INIS)

    Edwards, R.M.; Ruhl, D.W.; Klevans, E.H.; Robinson, G.E.

    1990-01-01

    A diagnostic system is under development for demonstration of Intelligent Distributed Control at the Experimental Breeder Reactor (EBR--II). In the first phase of the project a diagnostic system is being developed for the EBR-II steam plant based on the DISYS expert systems approach. Current testing uses recorded plant data and data from simulated plant faults. The dynamical simulation of the EBR-II steam plant uses the Babcock and Wilcox (B ampersand W) Modular Modeling System (MMS). At EBR-II the diagnostic system operates in the UNIX workstation and receives live plant data from the plant Data Acquisition System (DAS). Future work will seek implementation of the steam plant diagnostic in a distributed manner using UNIX based computers and Bailey microprocessor-based control system. 10 refs., 6 figs

  16. EBR-II secondary sodium loop Plugging Temperature Indicator control system upgrade

    International Nuclear Information System (INIS)

    Carlson, R.B.; Gehrman, R.L.

    1995-01-01

    The Experimental Breeder Reactor II (EBR-II) secondary sodium coolant loop Plugging Temperature Indicator (PTI) control system was upgraded in 1993 to a real-time computer based system. This was done to improve control, to remove obsolete and high maintenance equipment, and to provide a graphical CRT based operator interface. A goal was to accomplish this inexpensively using small, reliable computer and display hardware with a minimum of purchased software. This paper describes the PTI system, the upgraded control system and its operator interface, and development methods and tools. The paper then assesses how well the system met its goals, discusses lessons learned and operational improvements noted, and provides some recommendations and suggestions on applying small real-time control systems of this type

  17. EBR-II facility for cleaning and maintenance of LMR components

    International Nuclear Information System (INIS)

    Washburn, R.A.

    1986-01-01

    The cleaning and maintenance of EBR-II sodium wetted components is accomplished in a separate hands-on maintenance facility known as the Sodium Components Maintenance Shop (SCMS). Sodium removal is mostly done using alcohol but steam or water is used. The SCMS has three alcohol cleaning systems: one for small nonradioactive components, one for small radioactive components, and one for large radioactive components. The SCMS also has a water-wash station for the removal of sodium with steam or water. An Alcohol Recovery Facility removes radioactive contaminants from the alcohol and reclaims the alcohol for reuse. Associated with the large components cleaning system is a major component handling system

  18. Performance of commercially produced mixed-oxide fuels in EBR-II

    International Nuclear Information System (INIS)

    Hales, J.W.; Lawrence, L.A.

    1980-11-01

    Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

  19. Safety philosophy in upgrading the EBR-II plant protection system

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1976-01-01

    The EBR-II plant protection system (PPS) has been substantially modified, upgrading its performance to more fully comply with modern safety philosophy and criteria. The upgrading effort required that the total reactor system be evaluated for possible faults and that a PPS be designed to accommodate them. The result was deletion of a number of existing trip functions and upgrading of others. Particular attention was given to loss of primary pumping power and reactivity insertion events. The design and performance criteria for the PPS has been more firmly established, understanding of the PPS function has been improved and the reactor has been subjected to fewer spurious trips, improving operational reliability

  20. Experimental and theoretical investigations on the dynamic response of EBR-II ducts under pressure pulse loading

    International Nuclear Information System (INIS)

    Chopra, P.S.; Srinivas, S.

    1975-01-01

    In order to assess the potential damage to hexagonal subassembly ducts (cans) that may result from rapid gas release from a failed element the EBR-II project has conducted experiments and analyses. Additional experimental and analytical investigations are now being conducted to assure fail-safety of the ducts. Fail-safety is defined as the ability of a duct to withstand pressure pulses from failed elements during all reactor conditions without damage to adjacent ducts or any other problems in fuel handling. The results of 93 EBR-II duct tests conducted primarily by Koenig have been reported previously. The results of empirical correlations of some of these tests to determine the influence of several variables on the pressure pulse experienced by a duct and on the duct deformation are presented. The variables include the type of gas contained in the simulated element (tube), the element and duct materials, the presence or absence of flow restrictors in the element, and the way gas was released. 8 references. (auth)

  1. Systematic variation of threshold reaction rates in EBR-II

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Combs, B.L.; Davis, A.I.

    1976-01-01

    Characterization of neutron flux, fluence, and spectra in fast reactor irradiation environments is presently being carried out at HEDL utilizing the multiple foil technique. These fluences and spectra are then used to correlate damage effects data to produce damage functions or equations to predict materials effects under future irradiation conditions. The neutron flux and spectrum, then, act as a transfer function to relate present observations to future effects in the same or different environments and thus consistent fluence evaluations are of utmost importance. As part of a continuing program to establish the data base to meet consistency requirements, a systematic correlation of data from a recent dosimetry test in EBR-II is being made. The paper presents preliminary results of some of these correlations involving threshold reactions

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

    International Nuclear Information System (INIS)

    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. Software engineering for the EBR-II data acquisition system conversion

    International Nuclear Information System (INIS)

    Schorzman, W.

    1988-01-01

    The original data acquisition system (DAS) for the Experimental Breeder Reactor II (EBR-II) was placed into service with state-of-the-art computer and peripherals in 1970. Software engineering principles for real-time data acquisition were in their infancy, and the original software design was dictated by limited hardware resources. The functional requirements evolved from creative ways to gather and display data. This abstract concept developed into an invaluable tool for system analysis, data reporting, and as a plant monitor for operations. In this paper the approach is outlined to the software conversion project with the restraints of operational transparency and 6 weeks for final conversion and testing. The outline is then compared with the formal principles of software engineering to show the way that bridge the gap can be bridged between the theoretical and real world by analyzing the work and listing the lessons learned

  4. Dynamic response of the EBR-II secondary sodium system to postulated leaks of steam and water into sodium

    International Nuclear Information System (INIS)

    Srinivas, S.; Chopra, P.S.; Stone, C.C.

    1976-01-01

    The paper presents evaluations of the dynamic response of a steam generator system to postulated leaks of steam and water into sodium. This work is part of a comprehensive fail-safe analysis of the EBR-II steam generator system

  5. EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Paolo Balestra; Carlo Parisi; Andrea Alfonsi

    2016-02-01

    The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

  6. EBR-II spent fuel treatment demonstration project

    International Nuclear Information System (INIS)

    Benedict, R.W.; Henslee, S.P.

    1997-01-01

    For approximately 10 years, Argonne National Laboratory was developed a fast reactor fuel cycle based on dry processing. When the US fast reactor program was canceled in 1994, the fuel processing technology, called the electrometallurgical technique, was adapted for treating unstable spent nuclear fuel for disposal. While this technique, which involves electrorefining fuel in a molten salt bath, is being developed for several different fuel categories, its initial application is for sodium-bonded metallic spent fuel. In June 1996, the Department of Energy (DOE) approved a radiation demonstration program in which 100 spent driver assemblies and 25 spent blanket assemblies from the Experimental Breeder Reactor-II (EBR-II) will be treated over a three-year period. This demonstrated will provide data that address issues in the National Research Council's evaluation of the technology. The planned operations will neutralize the reactive component (elemental sodium) in the fuel and produce a low enriched uranium product, a ceramic waste and a metal waste. The fission products and transuranium elements, which accumulate in the electrorefining salt, will be stabilized in the glass-bonded ceramic waste form. The stainless steel cladding hulls, noble metal fission products, and insoluble residues from the process will be stabilized in a stainless steel/zirconium alloy. Upon completion of a successful demonstration and additional environmental evaluation, the current plans are to process the remainder of the DOE sodium bonded fuel

  7. Eutectic penetration times in irradiated EBR-II driver fuel elements

    International Nuclear Information System (INIS)

    Betten, P.R.; Bottcher, J.H.; Seidel, B.R.

    1983-01-01

    The experimental test procedure employed the use of a high-temperature furnace which heated pre-irradiated elements to temperature and maintained the environment until element-cladding breach occurred. Pre-irradiated elements of the Mark-II design were first encapsulated in a close-fitting sealed tube that was instrumented with a pressure transducer at the top of the tube and a thermocouple at the element's top-of-fuel axial location. The volume of the capsule was evacuated in order to better identify the pressure pulse which would occur on breach and to minimize contaminants. Next, a three-zone fast-recovery furnace was heated and an axial temperature profile, similar to that experienced in the EBR-II core, was established. The encapsulated element was then quickly inserted into the furnace and remained there until clad breach occurred. The element was then removed from the furnace immediately. Visual and metallurgical examination of the rupture site was done later. A total of seven elements were tested in the above manner

  8. The EBR-II materials-surveillance program. 4: Results of SURV-4 and SURV-6

    International Nuclear Information System (INIS)

    Ruther, W.E.; Hayner, G.O.; Carlson, B.G.; Ebersole, E.R.; 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. For both the irradiated and thermally aged samples, 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 addition to the fifteen types of metal samples, graphite blocks were irradiated in the SURV subassemblies to determine the effect of irradiation on the graphite neutron shield. In this report, the properties of these materials irradiated at 370 C to a total fluence of 2.2 x 10 22 n/cm 2 (over 2,994 days) are compared with those of similar specimens thermally aged at 370 C for 2,994 days in the storage basket of the reactor. The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, impact strength, and creep

  9. Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1986-09-01

    Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

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

    International Nuclear Information System (INIS)

    Pahl, R. G.

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

    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

  12. EBR-II Cover Gas Cleanup System upgrade distributed control and front end computer systems

    International Nuclear Information System (INIS)

    Carlson, R.B.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) Cover Gas Cleanup System (CGCS) control system was upgraded in 1991 to improve control and provide a graphical operator interface. The upgrade consisted of a main control computer, a distributed control computer, a front end input/output computer, a main graphics interface terminal, and a remote graphics interface terminal. This paper briefly describes the Cover Gas Cleanup System and the overall control system; gives reasons behind the computer system structure; and then gives a detailed description of the distributed control computer, the front end computer, and how these computers interact with the main control computer. The descriptions cover both hardware and software

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

    International Nuclear Information System (INIS)

    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 x 10 22 n/cm 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 x 10 22 n/cm 2 (the maximum fluence of the SURV-4 samples) to 3.2 x 10 22 n/cm 2

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    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

  15. EBR-II Cover Gas Cleanup System (CGCS) upgrade graphical interface design

    International Nuclear Information System (INIS)

    Staffon, J.D.; Peters, G.G.

    1992-01-01

    Technology advances in the past few years have prompted an effort at Argonne National Laboratory to replace existing equipment with high performance digital computers and color graphic displays. Improved operation of process systems can be achieved by utilizing state-of-the-art computer technology in the areas of process control and process monitoring. The Cover Gas Cleanup System (CGCS) at EBR-II is the first system to be upgraded with high performance digital equipment. The upgrade consisted of a main control computer, a distributed control computer, a front end input/output computer, a main graphics interface terminal, and a remote graphics interface terminal. This paper describes the main control computer and the operator interface control software

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

    International Nuclear Information System (INIS)

    Lehto, W.K.; Gross, K.C.

    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 ampersand 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

  17. Run-beyond-clad-breach oxide testing in EBR-2

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Bottcher, J.H.; Strain, R.V.; Gross, K.C.; Lee, M.J.; Webb, J.P.; Colburn, R.P.; Ukai, S.; Nomura, S.; Odo, T.; Shikakura, S.

    1990-01-01

    Fourteen tests sponsored by the US and Japan were used to study reliability of breached LMR oxide fuel pins during continued operation in EBR-II for a range of conditions and parameters. The fuel-sodium reaction product governed pin behavior. It extended primary breaches by swelling and promoted secondary failures, yet it inhibited loss of fuel and fission products and enhanced release of delayed neutrons used in monitoring breach condition. Fission gas and cesium, the main contaminants from failures, could be adequately controlled. This positive EBR-II experience suggested that limited operation with failed fuel may be feasible in commercial LMR's. 16 refs., 14 figs., 4 tabs

  18. Design of a reactor inlet temperature controller for EBR-2 using state feedback

    International Nuclear Information System (INIS)

    Vilim, R.B.; Planchon, H.P.

    1990-01-01

    A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs

  19. Metal waste forms from treatment of EBR-II spent fuel

    International Nuclear Information System (INIS)

    Abraham, D. P.

    1998-01-01

    Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

  20. BMFT-CEA-US-DOE Exchange on KNK II-Rapsodie-EBR II operating experience, German contributions for the second expert meeting at Idaho Falls, USA, October 27 and 28, 1982

    International Nuclear Information System (INIS)

    1982-10-01

    The meeting at Idaho Falls was the follow-up meeting of the first expert meeting on EBR II- Rapsodie- KNK II operating experience, which took place at the Karlsruhe Research Center in March 1980. The present report compiles the ten German papers presented at the Idaho Falls meeting, discussing various aspects of experience gained by the operation of KNK II

  1. Safety related considerations for operation with defected elements in EBR-II

    International Nuclear Information System (INIS)

    Fryer, R.M.; Sackett, J.I.; Lambert, J.D.B.

    1976-01-01

    Traditionally, EBR-II has employed the 'shutdown and remove' philosophy when breached fuel elements are encountered. This mode of operation maintained in-plant inventories of fission products at low levels and allowed certain fission product detection systems to be employed as automatic plant shutdown devices. Information from fuel failure propagation studies and fast reactor operation indicates that shutdown under these conditions is unwarranted. Analytical studies, as well as fast reactor experience, further indicate that failure propagation, if it occurs at all, will not cross adjacent subassembly boundaries. Therefore, the 'shutdown and remove' philosophy can be liberalized to allow the demonstration of safety during a run-beyond-clad-breach mode of operation. This mode of operation is essential to the demonstration of the economics of commercial LMFBR systems

  2. Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup

    International Nuclear Information System (INIS)

    Stone, I.Z.; Jost, J.W.; Baker, R.B.

    1979-01-01

    A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations

  3. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  4. Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

    International Nuclear Information System (INIS)

    Chang, L.K.; Koenig, J.F.; Porter, D.L.

    1987-01-01

    In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

  5. EBR-II blanket fuel leaching test using simulated J-13 well water.

    Energy Technology Data Exchange (ETDEWEB)

    Fonnesbeck, J. E.

    1998-05-15

    A pulsed-flow leaching test is being conducted using three EBR-II blanket fuel segments. These samples are immersed in simulated J-13 well water. The samples are kept at a constant temperature of 90 C. Leachate is exchanged weekly and analyzed for various nuclides which are of interest from a mobility and longevity point of view. Our primary interest is in the longer-lived species such as {sup 99}Tc, {sup 237}Np, and {sup 241}Am. In addition, the behavior of U, Pu, {sup 90}Sr, and {sup 137}Cs are being analyzed. During the course of this experiment, an interesting observation has been made involving one of the samples which could indicate the possible rapid ''anoxic'' oxidation of uranium metal to UO{sub 2}.

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

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    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

  7. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Surveillance application using patten recognition software at the EBR-II Reactor Facility

    International Nuclear Information System (INIS)

    Olson, D.L.

    1992-01-01

    The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodium Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller

  10. Development of a graphical user interface allowing use of the SASSYS LMR systems analysis code as an EBR-II interactive simulator

    International Nuclear Information System (INIS)

    Garner, P.L.; Briggs, L.L.; Gross, K.C.; Ku, J.Y.; Staffon, J.D.

    1994-01-01

    The SASSYS computer program for safety analyses of liquid-metal- cooled fast reactors has been adapted for use as the simulation engine under the graphical user interface provided by the GRAFUN and HIST programs and the Data Views software package under the X Window System on UNIX-based computer workstations to provide a high fidelity, real-time, interactive simulator of the Experimental Breeder Reactor Number II (EBR-II) plant. In addition to providing analysts with an interactive way of performing safety case studies, the simulator can be used to investigate new control room technologies and to supplement current operator training

  11. Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

    International Nuclear Information System (INIS)

    Klueh, R.L.; Vitek, J.M.

    1984-01-01

    The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

  12. Microsegregation observed in Fe-35.5Ni-7.5Cr irradiated in EBR-II

    International Nuclear Information System (INIS)

    Brager, H.R.; Garner, F.A.

    1984-01-01

    At 593 0 C one alloy, Fe-35.5Ni-7.5Cr, which was particularly resistant to swelling in EBR-II, increased in density 0.9% at 7.6 x 10 22 n/cm 2 (E > 0.1 MeV). Examination by energy dispersive x-ray analysis revealed that substantial oscillations occur in the nickel content of the alloy, varying from 25 to 53% about the nominal level of 35.5%. These oscillations exhibit a period of approx.200 nm. Regions enriched in nickel are depleted in chromium and iron, and the reverse is true in regions of low nickel content. This spinodal-like process produces a net densification and also appears to eventually destroy the swelling resistance of the alloy. Once voids form in the nickel-poor chromium-rich regions, further segregation of nickel to void surfaces is expected to accelerate the loss of swelling resistance

  13. EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1981-01-01

    The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

  14. Software engineering for the EBR-II data acquisition system conversion

    International Nuclear Information System (INIS)

    Schorzman, W.

    1988-01-01

    The purpose of this paper is to outline how EBR-II engineering approached the data acquisition system (DAS) software conversion project with the restraints of operational transparency and six weeks for final implementation and testing. Software engineering is a relatively new discipline that provides a structured philosopy for software conversion. The software life cycle is structured into six basic steps: 1) initiation, 2) requirements definition, 3) design, 4) programming, 5) testing, and 6) operations. These steps are loosely defined and can be altered to fit specific software applications. DAS software is encompassed from three sources: 1) custom software, 2) system software, and 3) in-house application software. A data flow structure is used to describe the DAS software. The categories are: 1) software used to bring signals into the central processer, 2) software that transforms the analog data to engineering units and then logs the data in the data store, and 3) software used to transport and display the data. The focus of this paper is to describe how the conversion team used a structured engineering approach and utilized the resources available to produce a quality system on time. Although successful, the conversion process provided some pit falls and stumbling blocks. Working through these obstacles enhanced our understanding and surfaced in the form of LESSONS LEARNED, which are gracefully shared in this paper

  15. Experimental study of the transition from forced to natural circulation in EBR-II at low power and flow

    International Nuclear Information System (INIS)

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

    1979-01-01

    A series of tests have been conducted in EBR-II which studied the dynamics of the transition from forced to natural circulation flow in a liquid-metal-cooled fast breeder reactor (LMFBR). Each test was initiated by abruptly tripping an electromagnetic pump which supplies 5 to 6% of the normal full operational primary flow rate. The ensuing flow coast-down reached a minimum value after which the flow increased as natural circulation was established. The effects of secondary system flow through the intermediate heat exchanger and reactor decay power level on the minimum in-core flow rates and maximum in-core temperatures were examined

  16. Tensile and fracture properties of EBR-II-irradiated V-15Cr-5Ti containing helium

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Horak, J.A.

    1986-01-01

    The alloy V-15Cr-5Ti was cyclotron-implanted with 80 appM He and subsequently irradiated in the Experimental Breeder Reactor (EBR-II) to 30 dpa. The same alloy was also irradiated in the 10, 20, and 30% cold-worked conditions. Irradiation temperatures ranged from 400 to 700 0 C. No significant effects of helium on mechanical properties were found in this temperature range although the neutron irradiation shifted the temperature of transition from cleavage to ductile fracture to about 625 0 C. Ten percent cold work was found to have a beneficial effect in reducing the tendency for cleavage fracture following irradiation, but high levels (20%) were observed to reduce ductility. Still higher levels (30%) improved ductility by inducing recovery during the elevated-temperature irradiation. Swelling was found to be negligible, but precipitates - titanium oxides or carbonitrides - contained substantial cavities

  17. Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Wiffen, F.W.

    1985-01-01

    The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs

  18. Integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

    1984-01-01

    Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept

  19. Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Mariani, R.D.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

  20. Alternate form and placement of short lived reactor waste and associated fuel hardware for decommissioning of EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Planchon, H.P.; Singleterry, R.C. Jr.

    1995-12-01

    Upon the termination of EBR-II operation in 1994, the mission has progressed to decommissioning and waste cleanup of the facility. The simplest method to achieve this goal is to bury the raw fuel and activated steel in an approved burial ground or deep geologic repository. While this might be simple, it could be very expensive, consume much needed burial space for other materials, and leave large amounts of fissile easily available to future generations. Also, as with any operation, an associated risk to personnel and the public from the buried waste exists. To try and reduce these costs and risks, alternatives to burial are sought. One alternative explored here for EBR-II is to condition the fuel and store the fission products and steel either permanently or temporarily in the sealed primary boundary of the decommissioned reactor. The first problem is to identify which subassemblies are going to be conditioned and their current composition and decay time. The next problem is to identify the conditioning process and determine the composition and form of the waste streams. The volume, mass, heat, and curie load of the waste streams needs to be determined so a waste-assembly can be designed. The reactor vessel and internals need to be analyzed to determine if they can handle these loads. If permanent storage is the goal, then mechanisms for placing the waste-assembly in the reactor vessel and sealing the vessel are needed. If temporary storage is the goal, then mechanisms for waste-assembly placement and retrieval are needed. This paper answers the technical questions of volume, mass, heat, and curie loads while just addressing the other questions found in a safety analysis. The final conclusion will compare estimated risks from the burial option and this option.

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

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Y. [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Sencer, B.H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Garner, F.A. [Radiation Effects Consulting, Richland, WA 99354 (United States); Marquis, E.A., E-mail: emarq@umich.edu [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2015-12-15

    AISI 304 stainless steel was irradiated at 416 °C and 450 °C at a 4.4 × 10{sup −9} and 3.05 × 10{sup −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.

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

    Science.gov (United States)

    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.

  3. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  4. Tensile and fracture properties of EBR-II-irradiated V-15Cr-5Ti containing helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Horak, J.A.

    1986-01-01

    The alloy V-15Cr-5Ti was cyclotron-implanted with 80 appM He and subsequently irradiated in the Experimental Breeder Reactor (EBR-II) to 30 dpa. The same alloy was also irradiated in the 10, 20, and 30% cold-worked conditions. Irradiation temperatures ranged from 400 to 700/sup 0/C. No significant effects of helium on mechanical properties were found in this temperature range although the neutron irradiation shifted the temperature of transition from cleavage to ductile fracture to about 625/sup 0/C. Ten percent cold work was found to have a beneficial effect in reducing the tendency for cleavage fracture following irradiation, but high levels (20%) were observed to reduce ductility. Still higher levels (30%) improved ductility by inducing recovery during the elevated-temperature irradiation. Swelling was found to be negligible, but precipitates - titanium oxides or carbonitrides - contained substantial cavities.

  5. Validation of the REBUS-3/RCT methodologies for EBR-II core-follow analysis

    International Nuclear Information System (INIS)

    McKnight, R.D.

    1992-01-01

    One of the many tasks to be completed at EBR-2/FCF (Fuel Cycle Facility) regarding fuel cycle closure for the Integral Fast Reactor (IFR) is to develop and install the systems to be used for fissile material accountancy and control. The IFR fuel cycle and pyrometallurgical process scheme determine the degree of actinide of actinide buildup in the reload fuel assemblies. Inventories of curium, americium and neptunium in the fuel will affect the radiation and thermal environmental conditions at the fuel fabrication stations, the chemistry of reprocessing, and the neutronic performance of the core. Thus, it is important that validated calculational tools be put in place for accurately determining isotopic mass and neutronic inputs to FCF for both operational and material control and accountancy purposes. The primary goal of this work is to validate the REBUS-2/RCT codes as tools which can adequately compute the burnup and isotopic distribution in binary- and ternary-fueled Mark-3, Mark-4, and Mark-5 subassemblies. 6 refs

  6. Integral-capture measurements and cross-section adjustments for Nd, Sm, and Eu

    International Nuclear Information System (INIS)

    Anderl, R.A.; Schmittroth, F.; Harker, Y.D.

    1981-07-01

    Integral-capture reaction rates are reported for 143 Nd, 144 Nd, 145 Nd, 147 Sm, 151 Eu, 152 Eu, 153 Eu, and 154 Eu irradiated in different neutron spectra in EBR-II. These reaction rates are based primarily on mass-spectrometric measurements of the isotopic atom ratios of the capture product to the target nuclide. The neutron spectra are characterized using passive neutron dosimetry and spectrum-unfolding with the FERRET least-squares data analysis code. Reaction rates for the neutron spectrum monitors were determined by the radiometric technique using Ge(Li) spectrometers. These rates are also reported here. The integral data for the rare-earth samples and for the spectrum monitors were used in multigroup flux/cross-section adtustment analyses with FERRET to generate adjustments to 47 group representations of the ENDF/B-IV capture cross sections for the rare-earth isotopes. These adjusted cross sections are in good agreement with recent differential data and with adjusted cross sections based on STEK integral data. Examples are given of the use of the adjusted cross sections and covariance matrices for cross-section evaluation

  7. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    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

  8. Training experience at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report

  9. EBR-II blanket fuel leaching test using simulated J-13 well water

    International Nuclear Information System (INIS)

    Fonnesbeck, J. E.

    1999-01-01

    This paper discusses the results of a pulsed-flow leaching test using simulated J-13 well water leachant. This test was performed on three blanket fuel segments from the ANL-W EBR-II nuclear reactor which were originally made up of depleted uranium (DU). This experiment was designed to mimic conditions which would exist if, upon disposal of this material in a geological repository, it came in direct contact with groundwater. These segments were contained in pressure vessels and maintained at a constant temperature of 90 C. Weekly aliquots of leachate were taken from the three vessels and replaced with an equal volume of fresh leachant. These weekly aliquots were analyzed for both 90 Sr and 137 Cs. The results of the pulsed-flow leach test showed the formation of uranium oxide (UO 2 ) and uranium hydride (UH 3 ) particulate with rapid release of the 137 Cs and 90 Sr to the leachant. On the fifth week of sampling, one of the vessels became over pressurized and vented gas when opened. The most reasonable explanation for the presence of gas in this vessel is that the unoxidized uranium metal in the blanket segment could have reacted with the surrounding water leachant to form hydrogen. However, an investigation is currently being undertaken to both qualify and quantify H 2 formation during uranium spent nuclear fuel corrosion in water

  10. Improvements in EBR-2 core depletion calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Hill, R.N.; Sakamoto, S.

    1991-01-01

    The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Tsai, H.C.; Nowicki, L.J. [Argonne National Lab., IL (United States)] [and others

    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.

  12. Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

    International Nuclear Information System (INIS)

    Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

    1975-01-01

    The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, S.D.; Gese, N.J. [Separations Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Wurth, L.A. [Zinc Air Inc., 5314-A US Hwy 2 West, Columbia Falls, MT 59912 (United States)

    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. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  15. A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  16. Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1994-10-01

    From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

  17. Actinide behavior in the integral fast reactor. Progress report, May 1, 1992--April 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory`s site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core.

  18. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-01-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II

  19. High flux-fluence measurements in fast reactors

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Ulseth, J.A.

    1977-01-01

    Characterization of irradiation environments for fuels and materials tests in fast reactors requires determination of the neutron flux integrated over times as long as several years. An accurate integration requires, therefore, passive dosimetry monitors with long half-life or stable products which can be conveniently measured. In addition, burn-up, burn-in, and burn-out effects must be considered in high flux situations and use of minimum quantities of dosimeter materials is often desirable. These conditions force the use of dosimeter and dosimeter container designs, measured products, and techniques that are different from those that are used in critical facilities and other well-characterized benchmark fields. Recent measurements in EBR-II indicate that high-accuracy results can be attained and that tie-backs to benchmark field technique calibrations can be accomplished

  20. A status report on the integral fast reactor fuels and safety program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

  1. Safety analysis report. Decontamination and decommissioning of the EBR-I Complex

    International Nuclear Information System (INIS)

    Commander, J.C.; Macbeth, P.J.; Michels, D.E.

    1975-06-01

    The safety aspects of the planned EBR-I Complex decontamination and decommissioning operations are assessed. The major operations are: (1) removal of NaK from the EBR-I primary and secondary coolant systems, (2) processing of the NaK to produce solid caustic for disposal, (3) decontamination of contaminated areas of EBR-I and ZPR-III, (4) removal of items that cannot be decontaminated economically to acceptably safe levels, (5) isolation of contaminated areas, (6) demolition of the AFSR Shielding, and (7) removal of contaminated vessels from the NaK storage pit. It may be concluded that although potential hazards do exist from explosion, chemical burns and low-level radioactive exposure from the D and D operation, these hazards represent acceptable risks provided that the established procedures and precautions are followed. (U.S.)

  2. An integrated real-time diagnostic concept using expert systems, qualitative reasoning and quantitative analysis

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Kumara, S.; Levine, S.H.

    1989-01-01

    An approach for an integrated real-time diagnostic system is being developed for inclusion as an integral part of a power plant automatic control system. In order to participate in control decisions and automatic closed loop operation, the diagnostic system must operate in real-time. Thus far, an expert system with real-time capabilities has been developed and installed on a subsystem at the Experimental Breeder Reactor (EBR-II) in Idaho, USA. Real-time simulation testing of advanced power plant concepts at the Pennsylvania State University has been developed and was used to support the expert system development and installation at EBR-II. Recently, the US National Science Foundation (NSF) and the US Department of Energy (DOE) have funded a Penn State research program to further enhance application of real-time diagnostic systems by pursuing implementation in a distributed power plant computer system including microprocessor based controllers. This paper summarizes past, current, planned, and possible future approaches to power plant diagnostic systems research at Penn State. 34 refs., 9 figs

  3. Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

    1993-01-01

    This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

  4. Experimental confirmation of the design to minimize vibration and wear in 61-pin wire-spaced EBR-II subassemblies

    International Nuclear Information System (INIS)

    Fukuda, S.K.

    1978-05-01

    Examinations of HEDL 61-pin subassemblies comprised of 5.84 mm (0.230) inch diameter mixed-oxide fuel pins with 1.02 mm (0.040'') diameter spacer wire (PNL-9, -10, -11, HEDL-N-E, -N-F), showed severe cladding and spacer wire wear after irradiation in EBR-II. A comparison of a large number of design, fabrication, and irradiation parameters for all of the HEDL subassemblies indicated that the porosity per ring of fuel pins correlated significantly with the occurrence of wear on the fuel pins. The porosity per ring is the clearance between the flat-to-flat pin bundle dimension and the inner hex can dimension divided by the number of hexagonal fuel pin rings in the subassembly. The porosity per ring for PNL-9, -10, -11 and HEDL-N-E was 0.15 mm/ring (6 mils/ring) and 0.18 mm/ring (7 mils/ring) for the HEDL-N-F subassembly. Since the original FTR subassembly design had a porosity/ring spread of 0.04 mm/ring to 0.16 mm/ring (1.67 to 6.11 mils/ring) an additional series of irradiation tests was conducted to confirm that a tighter fuel pin bundle would eliminate the wear

  5. Intercomparison study of atmospheric methane and carbon dioxide concentrations measured at the Ebre River Delta Station

    Science.gov (United States)

    Occhipinti, Paola; Morguí, Josep Anton; Àgueda, Alba; Batet, Oscar; Borràs, Sílvia; Cañas, Lídia; Curcoll, Roger; Grossi, Claudia; Nofuentes, Manel; Vazquez, Eusebi; Rodó, Xavier

    2015-04-01

    In the framework of the ClimaDat project, IC3 has established a network of eight monitoring stations across the Iberian Peninsula and the Canarian Archipelago with the aim of studying climate processes. The monitoring station at the Ebre River Delta (DEC3) is located in the Ebre River Delta Natural Park (40° 44' N; 0° 47' E) and it is characterized by the typical North-Western Mediterranean climate. Since 2013, atmospheric greenhouse gases (GHG) and 222Rn tracer gas together with the meteorological parameters are continuously measured from a 10 m a.g.l. height tower. Atmospheric GHG (CO2, CH4, CO and N2O) concentrations are determined using a Picarro analyzer G2301 (CO2 and CH4) and a modified gas chromatograph (GC) Agilent 6890N (CO2, CH4, CO and N2O). Open data access is available from the www.climadat.es website. Data collected at the DEC3 station are also submitted to the InGOS platform since this station is part of the InGOS European infrastructure project. Researchers from the Laboratory of the Atmosphere and the Oceans (LAO) at IC3 have performed an intercomparison study at the DEC3 site between three different Picarro analyzers (two Picarro G2301 and one Picarro G2301M), a Los Gatos Research (LGR) analyzer and the GC system already installed at the station. The aim of this study is to compare and assess the measuring agreement between the four optical gas analyzers and the GC. In the first part of the experiment, all instruments have been calibrated using NOAA gases as primary standards analyzing five Praxair provided targets to evaluate the precision of the measuring instruments. Max Plank Institute (MPI) gases have been used as secondary standards for the GC whereas Praxair provided tanks are used as secondary standards for the Picarro and the LGR analyzers. In the second part of the experiment, atmospheric GHG were measured from natural atmospheric air taken from a 10 m a.g.l. inlet. Daily cycles of GHG measurements were carried out using different

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

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Y., E-mail: yina.huang@materials.ox.ac.uk [University of Wisconsin, Madison, WI 53706 (United States); Wiezorek, J.M.K. [University of Pittsburgh, Pittsburgh, PA 15260 (United States); Garner, F.A. [Radiation Effects Consulting, 2003 Howell Ave., Richland, WA 99354 (United States); Freyer, P.D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Okita, T. [University of Tokyo, Tokyo (Japan); Sagisaka, M.; Isobe, Y. [Nuclear Fuel Industries, Ltd., Osaka (Japan); Allen, T.R. [University of Wisconsin, Madison, WI 53706 (United States)

    2015-10-15

    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

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

    Science.gov (United States)

    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

  8. Functional and operational design requirements for decontamination and decommissioning of the EBR-I Mark-II NaK: Final report

    International Nuclear Information System (INIS)

    Brown, B.W.; Crandall, D.L.; Dafoe, R.E.; Dolenc, M.R.; LaRue, D.M.

    1987-09-01

    Approximately 180 gal of sodium/potassium (NaK) eutectic liquid metal were severely radioactively contaminated during a meltdown of the Mark-II core of the Experimental Breeder Reactor-I (EBR-I) in November 1955. This contaminated NaK, which is contained in four vessels, is currently stored in an underground bunker located at the Army Reentry Vehicle Facility Site (ARVFS) located approximately at the center of the Idaho National Engineering Laboratory (INEL). This document presents the Functional and Operational Requirements (F and ORs) for the D and D of the contaminated NaK and the ARVFS bunker site. This project will chemically deactivate the NaK; dispose of the radioactively contaminated product at a designated burial site; chemically deactivate any residual NaK in the containers, and dispose of the containers at a designated burial site; decontaminate and decommission any contaminated process equipment used in these operations, and decontaminate and decommission the ARVFS bunker site. Completion of the above technical objectives will allow for the effective disposition of the NaK, and will return the ARFVS bunker and immediate area to a reusable condition. Upon completion, the ARVFS NaK, which is now considered a significant potential hazard, will be removed from the Surplus Facilities Management Program priority listing of projects. 33 refs., 8 figs

  9. Multifrequency tests in the EBR-II reactor plant

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.; Gross, K.C.

    1989-01-01

    A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

  10. CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

    International Nuclear Information System (INIS)

    Hess, A.L.

    1994-01-01

    A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

  11. ANL calculational methodologies for determining spent nuclear fuel source term

    International Nuclear Information System (INIS)

    McKnight, R. D.

    2000-01-01

    Over the last decade Argonne National Laboratory has developed reactor depletion methods and models to determine radionuclide inventories of irradiated EBR-II fuels. Predicted masses based on these calculational methodologies have been validated using available data from destructive measurements--first from measurements of lead EBR-II experimental test assemblies and later using data obtained from processing irradiated EBR-II fuel assemblies in the Fuel Conditioning Facility. Details of these generic methodologies are described herein. Validation results demonstrate these methods meet the FCF operations and material control and accountancy requirements

  12. The influence of mechanical deformation on the irradiation creep of AISI 316 stainless steel irradiated in the EBR-II and FFTF fast reactors

    International Nuclear Information System (INIS)

    Garner, F.A.; Gilbert, E.R.

    2007-01-01

    Irradiation creep of stainless steels is thought not to be very responsive to material and environmental variables. To test this perception earlier unpublished experiments conducted in the EBR-II reactor on AISI 316 have been analyzed. While swelling is dependent on the cold-work level at 400-480 o C, the post-transient irradiation creep rate, often called the creep compliance B0, is not dependent on cold-work level. If the tube reaches pressures on reactor start-up that generate above-yield stresses in unirradiated steel, then plastic strains occur prior to significant irradiation, but the post-transient strain rate is identical to that of material that did not exceed the yield stress on start-up. It is shown that both stress-free and stress-affected swelling are isotropic and that the Soderberg relationship is maintained. At temperatures above ∼540 o C thermal creep and stored energy begin to assert themselves, with creep rates accelerating with cold-work and becoming non-linear with stress. These results are in agreement with a similar study on titanium-modified 316 steel in FFTF. (author)

  13. Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

    International Nuclear Information System (INIS)

    Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

    1976-02-01

    An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

  14. Decontamination and decommissioning of the EBR-I Complex. Final report

    International Nuclear Information System (INIS)

    Kendall, E.W.; Wang, D.K.

    1975-07-01

    This final report covers the Decontamination and Decommissioning (D and D) of the Experimental Breeder Reactor No. 1 (EBR-I) Complex funded under Contract No. AT(10-1)-1375. The major effort consisted of removal and processing of 5500 gallons of sodium/potassium (NaK) coolant from the EBR-I reactor system. Tests were performed to assess the explosive hazards of NaK and KO 2 in various environments and in contact with various contaminants likely to be encountered in the removal and processing operations. A NaK process plant was designed and constructed and the operation was successfully completed. Lesser effort was required for D and D of the Zero Power Reactor (ZPR-III) Facility, the Argonne Fast Source Reactor (AFSR) Shielding, and removal of contaminated NaK from the storage pit. The D and D effort was completed by 13 June 1975, ahead of schedule. (auth)

  15. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

  16. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

  17. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  18. ERB-II operating experience

    International Nuclear Information System (INIS)

    Smith, R.N.; Cissel, D.W.; Smith, R.R.

    1977-01-01

    As originally designed and operated, EBR-II successfully demonstrated the concept of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle (mini-nuclear park). Subsequent operation has been as an irradiation facility, a role which will continue into the foreseeable future. Since the beginning of operation in 1961, operating experience of EBR-II has been very satisfactory. Most of the components and systems have performed well. In particular, the mechanical performance of heat-removal systems has been excellent. A review of the operating experience reveals that all the original design objectives have been successfully demonstrated. To date, no failures or incidents resulting in serious in-core or out-of-core consequences have occurred. No water-to-sodium leaks have been detected over the life of the plant. At the present time, the facility is operating very well and continuously except for short shutdowns required by maintenance, refueling, modification, and minor repair. A plant factor of 76.9% was achieved for the calendar year 1976

  19. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Wade, D.C.

    1990-01-01

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  20. Rekindled interest in pyrometallurgical processing

    International Nuclear Information System (INIS)

    Burris, L.

    1986-01-01

    The IFR with its integral, on-site fuel recycle revived a concept pioneered at EBR-II. The reactor concept has become very attractive due to the advances in metal fuel performance over the past 15 years and in the understanding of the safety of metal-fueled reactors. The proposed fuel cycle carries out Lawroski's call for development of a low-cost fuel cycle for fast reactors to help them become economically competitive. The IFR represents a new direction in breeder developments. The next two years will be devoted to establishing experimentally the chemical feasibility of the pyrometallurgical process. Once it becomes feasible, the EBR-II fuel cycle facility can be refurbished and the process using IFR-type fuel irradiated in EBR-II

  1. The Integral Fast Reactor: A practical approach to waste management

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology

  2. Brassinosteroids improve photosystem II efficiency, gas exchange, antioxidant enzymes and growth of cowpea plants exposed to water deficit.

    Science.gov (United States)

    Lima, J V; Lobato, A K S

    2017-01-01

    Water deficit is considered the main abiotic stress that limits agricultural production worldwide. Brassinosteroids (BRs) are natural substances that play roles in plant tolerance against abiotic stresses, including water deficit. This research aims to determine whether BRs can mitigate the negative effects caused by water deficiency, revealing how BRs act and their possible contribution to increased tolerance of cowpea plants to water deficit. The experiment was a factorial design with the factors completely randomised, with two water conditions (control and water deficit) and three levels of brassinosteroids (0, 50 and 100 nM 24-epibrassinolide; EBR is an active BRs). Plants sprayed with 100 nM EBR under the water deficit presented significant increases in Φ PSII , q P and ETR compared with plants subjected to the water deficit without EBR. With respect to gas exchange, P N , E and g s exhibited significant reductions after water deficit, but application of 100 nM EBR caused increases in these variables of 96, 24 and 33%, respectively, compared to the water deficit + 0 nM EBR treatment. To antioxidant enzymes, EBR resulted in increases in SOD, CAT, APX and POX, indicating that EBR acts on the antioxidant system, reducing cell damage. The water deficit caused significant reductions in Chl a , Chl b and total Chl, while plants sprayed with 100 nM EBR showed significant increases of 26, 58 and 33% in Chl a , Chl b and total Chl, respectively. This study revealed that EBR improves photosystem II efficiency, inducing increases in Φ PSII , q P and ETR. This substance also mitigated the negative effects on gas exchange and growth induced by the water deficit. Increases in SOD, CAT, APX and POX of plants treated with EBR indicate that this steroid clearly increased the tolerance to the water deficit, reducing reactive oxygen species, cell damage, and maintaining the photosynthetic pigments. Additionally, 100 nM EBR resulted in a better dose-response of cowpea

  3. Integrating Seasonal Oscillations into Basel II Behavioural Scoring Models

    Directory of Open Access Journals (Sweden)

    Goran Klepac

    2007-09-01

    Full Text Available The article introduces a new methodology of temporal influence measurement (seasonal oscillations, temporal patterns for behavioural scoring development purposes. The paper shows how significant temporal variables can be recognised and then integrated into the behavioural scoring models in order to improve model performance. Behavioural scoring models are integral parts of the Basel II standard on Internal Ratings-Based Approaches (IRB. The IRB approach much more precisely reflects individual risk bank profile.A solution of the problem of how to analyze and integrate macroeconomic and microeconomic factors represented in time series into behavioural scorecard models will be shown in the paper by using the REF II model.

  4. Plant automation

    International Nuclear Information System (INIS)

    Christensen, L.J.; Sackett, J.I.; Dayal, Y.; Wagner, W.K.

    1989-01-01

    This paper describes work at EBR-II in the development and demonstration of new control equipment and methods and associated schemes for plant prognosis, diagnosis, and automation. The development work has attracted the interest of other national laboratories, universities, and commercial companies. New initiatives include use of new control strategies, expert systems, advanced diagnostics, and operator displays. The unique opportunity offered by EBR-II is as a test bed where a total integrated approach to automatic reactor control can be directly tested under real power plant conditions

  5. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  6. Development of an integrated signal validation system and application to operating power plants

    International Nuclear Information System (INIS)

    Upadhyaya, B.R.; Holbert, K.E.; Kerlin, T.W.

    1989-01-01

    The objective of the university-industry joint research program at the University of Tennessee and Combustion Engineering, Inc. is to develop and implement a comprehensive signal validation system for current power plants and future advanced reactors. The integrated system consists of several parallel signal processing modules. The multi-modular decision information is combined to detect, isolate and characterize faulty signals. The signal validation system has been implemented in a VAX workstation and applied to operational data from a pressurized water reactor (PWR) and the Experimental Breeder Reactor-II (EBR-II). The use of the various signal validation techniques may be extended to predictive maintenance advising, instrument calibration verification, and to the development of intelligent instrumentation systems. 18 refs., 6 figs

  7. Integrated Procurement Management System, Version II

    Science.gov (United States)

    Collier, L. J.

    1985-01-01

    Integrated Procurement Management System, Version II (IPMS II) is online/ batch system for collecting developing, managing and disseminating procurementrelated data at NASA Johnson Space Center. Portions of IPMS II adaptable to other procurement situations.

  8. Advanced condition monitoring techniques and plant life extension studies at EBR-2

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Perry, W.H.; King, R.W.

    1991-01-01

    Numerous advanced techniques have been evaluated and tested at EBR-2 as part of a plant-life extension program for detection of degradation and other abnormalities in plant systems. Two techniques have been determined to be of considerable assistance in planning for the extended-life operation of EBR-2. The first, a computer-based pattern-recognition system (System State Analyzer or SSA) is used for surveillance of the primary system instrumentation, primary sodium pumps and plant heat balances. This surveillance has indicated that the SSA can detect instrumentation degradation and system performance degradation over varying time intervals and can be used to provide derived signal values to replace signals from failed sensors. The second technique, also a computer-based pattern-recognition system (Sequential Probability Ratio Test or SPRT) is used to validate signals and to detect incipient failures in sensors and components or systems. It is being used on the failed fuel detection system and is experimentally used on the primary coolant pumps. Both techniques are described and experience with their operation presented

  9. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-01-01

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF

  10. PARALLEL MEASUREMENT AND MODELING OF TRANSPORT IN THE DARHT II BEAMLINE ON ETA II

    International Nuclear Information System (INIS)

    Chambers, F W; Raymond, B A; Falabella, S; Lee, B S; Richardson, R A; Weir, J T; Davis, H A; Schultze, M E

    2005-01-01

    To successfully tune the DARHT II transport beamline requires the close coupling of a model of the beam transport and the measurement of the beam observables as the beam conditions and magnet settings are varied. For the ETA II experiment using the DARHT II beamline components this was achieved using the SUICIDE (Simple User Interface Connecting to an Integrated Data Environment) data analysis environment and the FITS (Fully Integrated Transport Simulation) model. The SUICIDE environment has direct access to the experimental beam transport data at acquisition and the FITS predictions of the transport for immediate comparison. The FITS model is coupled into the control system where it can read magnet current settings for real time modeling. We find this integrated coupling is essential for model verification and the successful development of a tuning aid for the efficient convergence on a useable tune. We show the real time comparisons of simulation and experiment and explore the successes and limitations of this close coupled approach

  11. Integrating the fuel cycle at IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1992-01-01

    During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

  12. A state of the art on metallic fuel technology development

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

  13. Safety characteristics of the integral fast reactor concept

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

    1985-01-01

    The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

  14. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  15. Modeling of constituent redistribution in U-Pu-Zr metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-12-01

    A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

  16. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    International Nuclear Information System (INIS)

    Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L.

    1991-08-01

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

  17. Adaptive robust control of the EBR-II reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

    Simulation results are presented for an adaptive H ∞ controller, a fixed H ∞ 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 ∞ controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H ∞ and classical controllers. This makes for a superior and more robust controller

  18. A measurement-based method for predicting margins and uncertainties for unprotected accidents in the Integral Fast Reactor concept

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1990-01-01

    A measurement-based method for predicting the response of an LMR core to unprotected accidents has been developed. The method processes plant measurements taken at normal operation to generate a stochastic model for the core dynamics. This model can be used to predict three sigma confidence intervals for the core temperature and power response. Preliminary numerical simulations performed for EBR-2 appear promising. 6 refs., 2 figs

  19. Numerical integration subprogrammes in Fortran II-D

    Energy Technology Data Exchange (ETDEWEB)

    Fry, C. R.

    1966-12-15

    This note briefly describes some integration subprogrammes written in FORTRAN II-D for the IBM 1620-II at CARDE. These presented are two Newton-Cotes, Chebyshev polynomial summation, Filon's, Nordsieck's and optimum Runge-Kutta and predictor-corrector methods. A few miscellaneous numerical integration procedures are also mentioned covering statistical functions, oscillating integrands and functions occurring in electrical engineering.

  20. A state of the art on metallic fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

  1. Application of advanced technology to LMR control

    International Nuclear Information System (INIS)

    Lindsay, R.W.

    1989-01-01

    Key issues must be resolved to preserve the nuclear option; including new considerations for safety, economics, waste, transportation, diversion, etc. The programs at the Experimental Breeder Reactor II (EBR-II) are now carefully focused to provide answers to the above concerns in connection with the Integral Fast Reactor program at Argonne. Safety features that are inherent in plant design, coupled with automating plant control to help achieve the above objectives are more than just an issue of installing controllers and exotic algorithms, they include the complete integration of plant design, control strategy, and information presentation. Current technology development, both at Argonne and elsewhere includes efforts relating to the use of Artificial Intelligence, sensor/signal validation in many forms, pattern recognition, optimal develop and/or adopt promising technologies, and integrate them into an operating power plant for proof of value. After they have proven useful at EBR-II, it is expected that they can be incorporated into advanced designs such as PRISM and/or included in backfit activities as well. 6 refs

  2. Application of advanced technology to LMR control

    International Nuclear Information System (INIS)

    Lindsay, R.W.

    1989-01-01

    This paper reports that key issues must be resolved to preserve the nuclear option; including new considerations for safety, economics, waste, transportation, diversion, etc. The programs at the Experimental Breeder Reactor II (EBR-II) are now carefully focused to provide answers to the above concerns in connection with the Integral Fast Reactor program at Argonne. Safety features that are inherent in plant design, coupled with automating plant control to help achieve the above objectives are more than just an issue of installing controllers and exotic algorithms, they include the complete integration of plant design, control strategy, and information presentation. Current technology development, both at Argonne and elsewhere includes efforts relating to the use of Artificial Intelligence, sensor/signal validation in many forms, pattern recognition, optimal control technologies, etc. The eBR-II effort is to identify needs, develop and/or adopt promising technologies, and integrate them into an operating power plant for proof of value. After they have proven useful at EBR-II, it is expected that they can be incorporated into advanced designs such as PRISM and/or included in backfit activities as well

  3. Frequency response testing at Experimental Breeder Reactor II using discrete-level periodic signals

    International Nuclear Information System (INIS)

    Rhodes, W.D.; Larson, H.A.

    1990-01-01

    The Experimental Breeder Reactor 2 (EBR-2) reactivity-to-power frequency-response function was measured with pseudo-random, discrete-level, periodic signals. The reactor power deviation was small with insignificant perturbation of normal operation and in-place irradiation experiments. Comparison of results with measured rod oscillator data and with theoretical predictions show good agreement. Moreover, measures of input signal quality (autocorrelation function and energy spectra) confirm the ability to enable this type of frequency response determination at EBR-2. Measurements were made with the pseudo-random binary sequence, quadratic residue binary sequence, pseudo-random ternary sequence, and the multifrequency binary sequence. 10 refs., 7 figs., 3 tabs

  4. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  5. Interim waste storage for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

    1991-01-01

    The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

  6. EBR-II in-vessel natural circulation experiments on hot and cold pool stratification

    International Nuclear Information System (INIS)

    Ragland, W.A.; Feldman, E.E.

    1990-01-01

    The Experimental Breeder Reactor II is located in a cylindrical pool of liquid sodium which is part of the cold-leg of the primary flow circuit. A vertical string of 32 thermocouples spans the 8 m tank height, at each of two diametrically opposed locations in the primary tank. Local temperatures were measured with these 64 thermocouples during dynamic tests. The instantaneous spacial temperature distribution obtained from a string of thermocouples can be viewed on a personal computer. The animation which results from displaying successive spacial distributions provide a very effective way to quickly obtain physical insights. The design of the two strings of thermocouples, the software used to create the animation, measured data from three different types of tests--two unprotected reactor transients, and one with the reactor at decay power levels and the reactor cover lifted, are discussed. 5 refs., 3 figs

  7. Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

    International Nuclear Information System (INIS)

    William S. Charlton

    1999-01-01

    An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels

  8. Modifications to HFEF/S for IFR fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Forrester, R.J.; Carnes, M.D.; Rigg, R.H.

    1988-01-01

    Modifications have begun to the Hot Fuel Examination Facility-South (HFEF/S) in order to demonstrate the technology of the integral fast reactor (IFR) fuel cycle. This paper describes the status of the modifications to the facility and briefly reviews the status of the development of the process equipment. The HFEF/S was the demonstration facility for the early Experimental Breeder Reactor II (EBR-II) melt refining/injection-casting fuel cycle. Then called the Fuel Cycle Facility, ∼400 EBR-II fuel assemblies were recycled in the two hot cells of the facility during the 1964-69 period. Since then it has been utilized as a fuels examination facility. The objective of the IFR fuel cycle program is to upgrade HFEF/S to current standards, install new process equipment, and demonstrate the commercial feasibility of the IFR pyroprocess fuel cycle

  9. Assessment of Integrated Information System (IIS) in organization ...

    African Journals Online (AJOL)

    Assessment of Integrated Information System (IIS) in organization. ... to enable the Information System (IS) managers, as well as top management to understand the ... since organisational and strategic aspects in IIS should also be considered.

  10. Non-destructive assay of EBR-II blanket elements using resonance transmission analysis

    International Nuclear Information System (INIS)

    Klann, R.T.; Poenitz, W.P.

    1998-01-01

    Resonance transmission analysis utilizing a faltered reactor beam was examined as a means of determining the 239 Pu content in Experimental Breeder Reactor-II depleted uranium blanket elements. The technique uses cadmium and gadolinium falters along with a 239 Pu fission chamber to isolate the 0.3 eV resonance in 239 Pu. In the energy range of this resonance (0.1 eV to 0.5 ev), the total microscopic cross-section of 239 Pu is significantly greater than the cross-sections of 238 U and 235 U. This large difference allows small changes in the 239 Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and 239 Pu foils indicate a significant change in response based on the 239 Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of 239 Pu up to approximately two weight percent

  11. Validation and application of a physics database for fast reactor fuel cycle analysis

    International Nuclear Information System (INIS)

    McKnight, R.D.; Stillman, J.A.; Toppel, B.J.; Khalil, H.S.

    1994-01-01

    An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations

  12. Time series analysis of nuclear instrumentation in EBR-II

    International Nuclear Information System (INIS)

    Imel, G.R.

    1996-01-01

    Results of a time series analysis of the scaler count data from the 3 wide range nuclear detectors in the Experimental Breeder Reactor-II are presented. One of the channels was replaced, and it was desired to determine if there was any statistically significant change (ie, improvement) in the channel's response after the replacement. Data were collected from all 3 channels for 16-day periods before and after detector replacement. Time series analysis and statistical tests showed that there was no significant change after the detector replacement. Also, there were no statistically significant differences among the 3 channels, either before or after the replacement. Finally, it was determined that errors in the reactivity change inferred from subcritical count monitoring during fuel handling would be on the other of 20-30 cents for single count intervals

  13. Fast Reactor Knowledge Preservation Efforts. An Overview

    International Nuclear Information System (INIS)

    Grandy, Christopher

    2013-01-01

    • ARC-AFR Program is involved in a number of knowledge preservation activities; • Recovery of Information from EBR-II, FFTF, and TREAT is very important; • Recovery of Information from Office of Scientific and Technical Information (OSTI) and conversion to electronic format; • Organizing some data into electronic databases – EBR-II Plant Testing Data, FFTF Plant Testing Data, TREAT Test Data, SFR Fuels and Materials Irradiation Data, etc.; • Information is being used to support existing U.S. SFR programs along with international programs such as the IAEA CRP EBR-II Safety Benchmark

  14. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

    Smith, R.R.; Cissel, D.W.

    1978-01-01

    Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

  15. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-09-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

  16. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-01-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

  17. Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1983-01-01

    Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation

  18. Scram reliability under seismic conditions at the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Roglans, J.; Wang, C.Y.; Hill, D.J.

    1993-01-01

    A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

  19. Steam generator tube integrity program: Phase II, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.

  20. Steam generator tube integrity program: Phase II, Final report

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted

  1. Use of an isotope separator at the INEL

    International Nuclear Information System (INIS)

    Anderl, R.A.

    1977-01-01

    An electromagnetic isotope separator with a retardation lens as a collector was used to prepare highly enriched samples of Nd-143, -144, -145, -146, -148, -150, Sm-147, -149; Eu-151, -152, -153, -154. The 50 μg to 75 μg samples, deposited on 1 mil nickel foil or 0.5 mil vanadium foil, are part of a sample set to be irradiated in EBR-II as part of an integral-capture cross-section measurement program at the INEL. The isotope separator and the apparatus used for the sample preparation are described

  2. Method for the determination of technical specifications limiting temperature in EBR-II operation

    International Nuclear Information System (INIS)

    Chang, L.K.; Hill, D.J.; Ku, J.Y.

    2004-01-01

    The methodology and analysis procedure to qualify the Mark-V and Mark-VA fuels for the Experimental Breeder Reactor II are summarized in this paper. Fuel performance data and design safety criteria are essential for thermal-hydraulic analyses and safety evaluations. Normal and off-normal operation duty cycles and transient classifications are required for the safety assessment of the fuels. Design safety criteria for steady-state normal and transient off-normal operations were developed to ensure structural integrity of the fuel pin. The maximum allowable coolant outlet temperatures and powers of subassemblies for steady-state normal operation conditions were first determined in a row-by-row basis by a thermal-hydraulic and fuel damage analysis, in which a trial-and-error approach was used to predict the maximum subassembly coolant outlet temperatures and powers that satisfy the design safety criteria for steady-state normal operation conditions. The limiting steady-state temperature and power were then used as the initial subassembly thermal conditions for the off-normal transient analysis to assess the safety performance of the fuel pin for anticipated, unlikely and extremely unlikely events. If the design safety criteria for the off-normal events are not satisfied, then the initial steady-state subassembly temperatures and/or powers are reduced and an iterative procedure is employed until the design safety criteria for off-normal conditions are satisfied, and the initial subassembly outlet coolant temperature and power are the technical specification limits for reactor operation. (author)

  3. Measurements of actinide transmutation in the hard spectrum of a fast reactor

    International Nuclear Information System (INIS)

    Trybus, C.L.; Collins, P.J.; Maddison, D.W.; Bunde, K.A.; Pallmtag, S.; Palmiotti, G.

    1994-01-01

    Measurements of fission and capture in 235 U, 238 U, 239 Pu and 237 Np and in their product actinides have been made following irradiation in the metal-fuel core of EBR-II. The reactor has a peak flux around 500keV and the data complement measurements in the softer spectrum of an LMFBR. Irradiations were made at the same time for a set of standard dosimeter samples. These provide a test of calculated spectra and are also used for validation of steel activations and calculated atomic displacement rates. Calculation were made with modem transport codes using ENDF/B-5.2 data. Comparisons are made, using a simple homogeneous model, producing a similar spectrum, using ENDF/B-6.2 and JEFF-2 data

  4. Improvements in fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-12-01

    Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

  5. Integrated-light spectroscopy of globular clusters at the infrared Ca II lines

    Science.gov (United States)

    Armandroff, Taft E.; Zinn, Robert

    1988-01-01

    Integrated-light spectroscopy has been obtained for 27 globular clusters at the Ca II IR triplet. Line strengths and radial velocities have been measured from the spectra. For the well-studied clusters in the sample, the strength of the Ca II lines is very well correlated with previous metallicity estimates. Thus, the triplet is useful as a metallicity indicator in globular cluster integrated-light spectra. The greatly reduced effect of interstellar extinction at these wavelengths (compared to the blue region of the spectrum) has permitted observations of some of the most heavily reddened clusters in the Galaxy. For several such clusters, the Ca II triplet metallicities are in poor agreement with metallicity estimates from IR photometry by Malkan (1981). The strength of an interstellar band at 8621A has been used to estimate the amount of extinction towards these clusters. Using the new metallicity and radial-velocity data, the metallicity distribution, kinematics, and spatial distribution of the disk globular cluster system have been analyzed. Results very similar to those of Zinn (1985) have been found. The relation of the disk globulars to the stellar thick disk is discussed.

  6. Treatment of EBR-I NaK mixed waste at Argonne National Laboratory and subsequent land disposal at the Idaho National Engineering and Environmental Laboratory

    International Nuclear Information System (INIS)

    Herrmann, S. D.; Buzzell, J. A.; Holzemer, M. J.

    1998-01-01

    Sodium/potassium (NaK) liquid metal coolant, contaminated with fission products from the core meltdown of Experimental Breeder Reactor I (EBR-I) and classified as a mixed waste, has been deactivated and converted to a contact-handled, low-level waste at Argonne's Sodium Component Maintenance Shop and land disposed at the Radioactive Waste Management Complex. Treatment of the EBR-I NaK involved converting the sodium and potassium to its respective hydroxide via reaction with air and water, followed by conversion to its respective carbonate via reaction with carbon dioxide. The resultant aqueous carbonate solution was solidified in 55-gallon drums. Challenges in the NaK treatment involved processing a mixed waste which was incompletely characterized and difficult to handle. The NaK was highly radioactive, i.e. up to 4.5 R/hr on contact with the mixed waste drums. In addition, the potential existed for plutonium and toxic characteristic metals to be present in the NaK, resultant from the location of the partial core meltdown of EBR-I in 1955. Moreover, the NaK was susceptible to degradation after more than 40 years of storage in unmonitored conditions. Such degradation raised the possibility of energetic exothermic reactions between the liquid NaK and its crust, which could have consisted of potassium superoxide as well as hydrated sodium/potassium hydroxides

  7. Treatment of EBR-I NaK mixed waste at Argonne National Laboratory and subsequent land disposal at the Idaho National Engineering and Environmental Laboratory.

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, S. D.; Buzzell, J. A.; Holzemer, M. J.

    1998-02-03

    Sodium/potassium (NaK) liquid metal coolant, contaminated with fission products from the core meltdown of Experimental Breeder Reactor I (EBR-I) and classified as a mixed waste, has been deactivated and converted to a contact-handled, low-level waste at Argonne's Sodium Component Maintenance Shop and land disposed at the Radioactive Waste Management Complex. Treatment of the EBR-I NaK involved converting the sodium and potassium to its respective hydroxide via reaction with air and water, followed by conversion to its respective carbonate via reaction with carbon dioxide. The resultant aqueous carbonate solution was solidified in 55-gallon drums. Challenges in the NaK treatment involved processing a mixed waste which was incompletely characterized and difficult to handle. The NaK was highly radioactive, i.e. up to 4.5 R/hr on contact with the mixed waste drums. In addition, the potential existed for plutonium and toxic characteristic metals to be present in the NaK, resultant from the location of the partial core meltdown of EBR-I in 1955. Moreover, the NaK was susceptible to degradation after more than 40 years of storage in unmonitored conditions. Such degradation raised the possibility of energetic exothermic reactions between the liquid NaK and its crust, which could have consisted of potassium superoxide as well as hydrated sodium/potassium hydroxides.

  8. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  9. The Sodium Process Facility at Argonne National Laboratory-West

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1998-01-01

    Argonne National Laboratory-West (ANL-W) has approximately 680,000 liters of raw sodium stored in facilities on site. As mandated by the State of Idaho and the US Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The sodium will be processed in three separate and distinct campaigns: the 290,000 liters of Fermi-1 primary sodium, the 50,000 liters of the Experimental Breeder Reactor-II (EBR-II) secondary sodium, and the 330,000 liters of the EBR-II primary sodium. The Fermi-1 and the EBR-II secondary sodium contain only low-level of radiation, while the EBR-II primary sodium has radiation levels up to 0.5 mSv (50 mrem) per hour at 1 meter. The EBR-II primary sodium will be processed last, allowing the operating experience to be gained with the less radioactive sodium prior to reacting the most radioactive sodium. The sodium carbonate will be disposed of in 270 liter barrels, four to a pallet. These barrels are square in cross-section, allowing for maximum utilization of the space on a pallet, minimizing the required landfill space required for disposal

  10. The Sodium Process Facility at Argonne National Laboratory-West

    Energy Technology Data Exchange (ETDEWEB)

    Michelbacher, J.A.; Henslee, S.P. McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1998-07-01

    Argonne National Laboratory-West (ANL-W) has approximately 680,000 liters of raw sodium stored in facilities on site. As mandated by the State of Idaho and the US Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The sodium will be processed in three separate and distinct campaigns: the 290,000 liters of Fermi-1 primary sodium, the 50,000 liters of the Experimental Breeder Reactor-II (EBR-II) secondary sodium, and the 330,000 liters of the EBR-II primary sodium. The Fermi-1 and the EBR-II secondary sodium contain only low-level of radiation, while the EBR-II primary sodium has radiation levels up to 0.5 mSv (50 mrem) per hour at 1 meter. The EBR-II primary sodium will be processed last, allowing the operating experience to be gained with the less radioactive sodium prior to reacting the most radioactive sodium. The sodium carbonate will be disposed of in 270 liter barrels, four to a pallet. These barrels are square in cross-section, allowing for maximum utilization of the space on a pallet, minimizing the required landfill space required for disposal.

  11. Top-quark mass measurement from dilepton events at CDF II.

    Science.gov (United States)

    Abulencia, A; Acosta, D; Adelman, J; Affolder, T; Akimoto, T; Albrow, M G; Ambrose, D; Amerio, S; Amidei, D; Anastassov, A; Anikeev, K; Annovi, A; Antos, J; Aoki, M; Apollinari, G; Arguin, J-F; Arisawa, T; Artikov, A; Ashmanskas, W; Attal, A; Azfar, F; Azzi-Bacchetta, P; Azzurri, P; Bacchetta, N; Bachacou, H; Badgett, W; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Baroiant, S; Bartsch, V; Bauer, G; Bedeschi, F; Behari, S; Belforte, S; Bellettini, G; Bellinger, J; Belloni, A; Ben-Haim, E; Benjamin, D; Beretvas, A; Beringer, J; Berry, T; Bhatti, A; Binkley, M; Bisello, D; Bishai, M; Blair, R E; Blocker, C; Bloom, K; Blumenfeld, B; Bocci, A; Bodek, A; Boisvert, V; Bolla, G; Bolshov, A; Bortoletto, D; Boudreau, J; Bourov, S; Boveia, A; Brau, B; Bromberg, C; Brubaker, E; Budagov, J; Budd, H S; Budd, S; Burkett, K; Busetto, G; Bussey, P; Byrum, K L; Cabrera, S; Campanelli, M; Campbell, M; Canelli, F; Canepa, A; Carlsmith, D; Carosi, R; Carron, S; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chang, S H; Chapman, J; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Chlebana, F; Cho, I; Cho, K; Chokheli, D; Chou, J P; Chu, P H; Chuang, S H; Chung, K; Chung, W H; Chung, Y S; Ciljak, M; Ciobanu, C I; Ciocci, M A; Clark, A; Clark, D; Coca, M; Connolly, A; Convery, M E; Conway, J; Cooper, B; Copic, K; Cordelli, M; Cortiana, G; Cruz, A; Cuevas, J; Culbertson, R; Cyr, D; DaRonco, S; D'Auria, S; D'Onofrio, M; Dagenhart, D; de Barbaro, P; De Cecco, S; Deisher, A; De Lentdecker, G; Dell'Orso, M; Demers, S; Demortier, L; Deng, J; Deninno, M; De Pedis, D; Derwent, P F; Dionisi, C; Dittmann, J; Dituro, P; Dörr, C; Dominguez, A; Donati, S; Donega, M; Dong, P; Donini, J; Dorigo, T; Dube, S; Ebina, K; Efron, J; Ehlers, J; Erbacher, R; Errede, D; Errede, S; Eusebi, R; Fang, H C; Farrington, S; Fedorko, I; Fedorko, W T; Feild, R G; Feindt, M; Fernandez, J P; Field, R; Flanagan, G; Flores-Castillo, L R; Foland, A; Forrester, S; Foster, G W; Franklin, M; Freeman, J C; Fujii, Y; Furic, I; Gajjar, A; Gallinaro, M; Galyardt, J; Garcia, J E; Garcia Sciverez, M; Garfinkel, A F; Gay, C; Gerberich, H; Gerchtein, E; Gerdes, D; Giagu, S; Giannetti, P; Gibson, A; Gibson, K; Ginsburg, C; Giolo, K; Giordani, M; Giunta, M; Giurgiu, G; Glagolev, V; Glenzinski, D; Gold, M; Goldschmidt, N; Goldstein, J; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González, O; Gorelov, I; Goshaw, A T; Gotra, Y; Goulianos, K; Gresele, A; Griffiths, M; Grinstein, S; Grosso-Pilcher, C; Grundler, U; Guimaraes da Costa, J; Haber, C; Hahn, S R; Hahn, K; Halkiadakis, E; Hamilton, A; Han, B-Y; Handler, R; Happacher, F; Hara, K; Hare, M; Harper, S; Harr, R F; Harris, R M; Hatakeyama, K; Hauser, J; Hays, C; Hayward, H; Heijboer, A; Heinemann, B; Heinrich, J; Hennecke, M; Herndon, M; Heuser, J; Hidas, D; Hill, C S; Hirschbuehl, D; Hocker, A; Holloway, A; Hou, S; Houlden, M; Hsu, S-C; Huffman, B T; Hughes, R E; Huston, J; Ikado, K; Incandela, J; Introzzi, G; Iori, M; Ishizawa, Y; Ivanov, A; Iyutin, B; James, E; Jang, D; Jayatilaka, B; Jeans, D; Jensen, H; Jeon, E J; Jones, M; Joo, K K; Jun, S Y; Junk, T R; Kamon, T; Kang, J; Karagoz-Unel, M; Karchin, P E; Kato, Y; Kemp, Y; Kephart, R; Kerzel, U; Khotilovich, V; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, M S; Kim, S B; Kim, S H; Kim, Y K; Kirby, M; Kirsch, L; Klimenko, S; Klute, M; Knuteson, B; Ko, B R; Kobayashi, H; Kondo, K; Kong, D J; Konigsberg, J; Kordas, K; Korytov, A; Kotwal, A V; Kovalev, A; Kraus, J; Kravchenko, I; Kreps, M; Kreymer, A; Kroll, J; Krumnack, N; Kruse, M; Krutelyov, V; Kuhlmann, S E; Kusakabe, Y; Kwang, S; Laasanen, A T; Lai, S; Lami, S; Lami, S; Lammel, S; Lancaster, M; Lander, R L; Lannon, K; Lath, A; Latino, G; Lazzizzera, I; Lecci, C; LeCompte, T; Lee, J; Lee, J; Lee, S W; Lefèvre, R; Leonardo, N; Leone, S; Levy, S; Lewis, J D; Li, K; Lin, C; Lin, C S; Lindgren, M; Lipeles, E; Liss, T M; Lister, A; Litvintsev, D O; Liu, T; Liu, Y; Lockyer, N S; Loginov, A; Loreti, M; Loverre, P; Lu, R-S; Lucchesi, D; Lujan, P; Lukens, P; Lungu, G; Lyons, L; Lys, J; Lysak, R; Lytken, E; Mack, P; MacQueen, D; Madrak, R; Maeshima, K; Maki, T; Maksimovic, P; Manca, G; Margaroli, F; Marginean, R; Marino, C; Martin, A; Martin, M; Martin, V; Martínez, M; Maruyama, T; Matsunaga, H; Mattson, M E; Mazini, R; Mazzanti, P; McFarland, K S; McGivern, D; McIntyre, P; McNamara, P; McNulty, R; Mehta, A; Menzemer, S; Menzione, A; Merkel, P; Mesropian, C; Messina, A; von der Mey, M; Miao, T; Miladinovic, N; Miles, J; Miller, R; Miller, J S; Mills, C; Milnik, M; Miquel, R; Miscetti, S; Mitselmakher, G; Miyamoto, A; Moggi, N; Mohr, B; Moore, R; Morello, M; Movilla Fernandez, P; Mülmenstädt, J; Mukherjee, A; Mulhearn, M; Muller, Th; Mumford, R; Murat, P; Nachtman, J; Nahn, S; Nakano, I; Napier, A; Naumov, D; Necula, V; Neu, C; Neubauer, M S; Nielsen, J; Nigmanov, T; Nodulman, L; Norniella, O; Ogawa, T; Oh, S H; Oh, Y D; Okusawa, T; Oldeman, R; Orava, R; Osterberg, K; Pagliarone, C; Palencia, E; Paoletti, R; Papadimitriou, V; Papikonomou, A; Paramonov, A A; Parks, B; Pashapour, S; Patrick, J; Pauletta, G; Paulini, M; Paus, C; Pellett, D E; Penzo, A; Phillips, T J; Piacentino, G; Piedra, J; Pitts, K; Plager, C; Pondrom, L; Pope, G; Portell, X; Poukhov, O; Pounder, N; Prakoshyn, F; Pronko, A; Proudfoot, J; Ptohos, F; Punzi, G; Pursley, J; Rademacker, J; Rahaman, A; Rakitin, A; Rappoccio, S; Ratnikov, F; Reisert, B; Rekovic, V; van Remortel, N; Renton, P; Rescigno, M; Richter, S; Rimondi, F; Rinnert, K; Ristori, L; Robertson, W J; Robson, A; Rodrigo, T; Rogers, E; Rolli, S; Roser, R; Rossi, M; Rossin, R; Rott, C; Ruiz, A; Russ, J; Rusu, V; Ryan, D; Saarikko, H; Sabik, S; Safonov, A; Sakumoto, W K; Salamanna, G; Salto, O; Saltzberg, D; Sanchez, C; Santi, L; Sarkar, S; Sato, K; Savard, P; Savoy-Navarro, A; Scheidle, T; Schlabach, P; Schmidt, E E; Schmidt, M P; Schmitt, M; Schwarz, T; Scodellaro, L; Scott, A L; Scribano, A; Scuri, F; Sedov, A; Seidel, S; Seiya, Y; Semenov, A; Semeria, F; Sexton-Kennedy, L; Sfiligoi, I; Shapiro, M D; Shears, T; Shepard, P F; Sherman, D; Shimojima, M; Shochet, M; Shon, Y; Shreyber, I; Sidoti, A; Sill, A; Sinervo, P; Sisakyan, A; Sjolin, J; Skiba, A; Slaughter, A J; Sliwa, K; Smirnov, D; Smith, J R; Snider, F D; Snihur, R; Soderberg, M; Soha, A; Somalwar, S; Sorin, V; Spalding, J; Spinella, F; Squillacioti, P; Stanitzki, M; Staveris-Polykalas, A; St Denis, R; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Stuart, D; Suh, J S; Sukhanov, A; Sumorok, K; Sun, H; Suzuki, T; Taffard, A; Tafirout, R; Takashima, R; Takeuchi, Y; Takikawa, K; Tanaka, M; Tanaka, R; Tecchio, M; Teng, P K; Terashi, K; Tether, S; Thom, J; Thompson, A S; Thomson, E; Tipton, P; Tiwari, V; Tkaczyk, S; Toback, D; Tollefson, K; Tomura, T; Tonelli, D; Tönnesmann, M; Torre, S; Torretta, D; Tourneur, S; Trischuk, W; Tsuchiya, R; Tsuno, S; Turini, N; Ukegawa, F; Unverhau, T; Uozumi, S; Usynin, D; Vacavant, L; Vaiciulis, A; Vallecorsa, S; Varganov, A; Vataga, E; Velev, G; Veramendi, G; Veszpremi, V; Vickey, T; Vidal, R; Vila, I; Vilar, R; Vollrath, I; Volobouev, I; Würthwein, F; Wagner, P; Wagner, R G; Wagner, R L; Wagner, W; Wallny, R; Walter, T; Wan, Z; Wang, M J; Wang, S M; Warburton, A; Ward, B; Waschke, S; Waters, D; Watts, T; Weber, M; Wester, W C; Whitehouse, B; Whiteson, D; Wicklund, A B; Wicklund, E; Williams, H H; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, C; Worm, S; Wright, T; Wu, X; Wynne, S M; Yagil, A; Yamamoto, K; Yamaoka, J; Yamashita, Y; Yang, C; Yang, U K; Yao, W M; Yeh, G P; Yoh, J; Yorita, K; Yoshida, T; Yu, I; Yu, S S; Yun, J C; Zanello, L; Zanetti, A; Zaw, I; Zetti, F; Zhang, X; Zhou, J; Zucchelli, S

    2006-04-21

    We report a measurement of the top-quark mass using events collected by the CDF II detector from pp collisions at square root of s = 1.96 TeV at the Fermilab Tevatron. We calculate a likelihood function for the top-quark mass in events that are consistent with tt --> bl(-)nu(l)bl'+ nu'(l) decays. The likelihood is formed as the convolution of the leading-order matrix element and detector resolution functions. The joint likelihood is the product of likelihoods for each of 33 events collected in 340 pb(-1) of integrated luminosity, yielding a top-quark mass M(t) = 165.2 +/- 6.1(stat) +/- 3.4(syst) GeV/c2. This first application of a matrix-element technique to tt --> bl+ nu(l)bl'- nu(l') decays gives the most precise single measurement of M(t) in dilepton events. Combined with other CDF run II measurements using dilepton events, we measure M(t) = 167.9 +/- 5.2(stat) +/- 3.7(syst) GeV/c2.

  12. Status of IFR fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; McFarlane, H.F.

    1993-01-01

    The next major step in Argonne's Integral Fast Reactor (IFR) Program is demonstration of the pyroprocess fuel cycle, in conjunction with continued operation of EBR-II. The Fuel Cycle Facility (FCF) is being readied for this mission. This paper will address the status of facility systems and process equipment, the initial startup experience, and plans for the demonstration program

  13. Efficacy of intra-articular hyaluronic acid injections and exercise-based rehabilitation programme, administered as isolated or integrated therapeutic regimens for the treatment of knee osteoarthritis.

    Science.gov (United States)

    Saccomanno, Maristella F; Donati, Fabrizio; Careri, Silvia; Bartoli, Matteo; Severini, Gabriele; Milano, Giuseppe

    2016-05-01

    To assess the efficacy of intra-articular hyaluronic acid (HA) injections and exercise-based rehabilitation (EBR) programme, administered as isolated or integrated for the treatment of knee osteoarthritis. One hundred sixty-five patients affected by moderate degrees of knee OA were randomly divided into three groups. Group 1 (HA) underwent three HA injections (one every 2 weeks); group 2 (EBR) underwent 20 treatment sessions in a month of an individualized programme; and group 3 (HA + EBR) received both treatments simultaneously. Primary outcome was the Italian version of the Western Ontario and McMaster Universities (WOMAC) Osteoarthritis Index; secondary outcome was the evaluation of active range of movement (AROM). All patients were evaluated before and 1, 3 and 6 months after treatment. Significance was set at p injections and individualized rehabilitation programmes administered in isolation or in combination are effective in improving knee function and pain relief. The combined treatment showed the greatest pain relief at 1-month follow-up compared to either in isolation. Compared to the previous studies, this is the first study, which proposed an EBR programme tailored to the compartment of the knee joint most involved in the degenerative process. I.

  14. Determination of static and dynamic reactivity effects in KNK II

    International Nuclear Information System (INIS)

    Essig, C.

    1987-11-01

    In the frame of a pre-study of the KNK II test program two series of experiments related to inherent safety characteristics of sodium cooled breeder reactors have been elaborated, which are one basis for the performance of experiments of the Loss Of Flow (LOF) type and the Loss Of Heat Sink (LOHS) type. Tests of this type at KNK II would -different from the earlier tests at RAPSODIE and EBR-II- provide a demonstration of the inherently safe performance in case of a significantly non-zero Doppler effect. With a suitable execution, the foreseen series of experiments allow, as explained in this report, a substantial separation of the reactivity contributions and the determination of reactivity effects, i.e. the time constants of the recouplings. The performance and evaluation of these experiments with respect to the inherent safety potential will once more underline the distinguished role of KNK II for the development of fast breeders [de

  15. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  16. Toward an Integrated Psychological Approach - Phase II

    International Development Research Centre (IDRC) Digital Library (Canada)

    Trauma, Development and Peacebuilding : Toward an Integrated Psychological Approach - Phase II. Over the past decade, the peace, conflict and development community has begun to question the value of medicalized approaches such as post-traumatic stress disorder (PTSD) in dealing with aftermath of political violence ...

  17. Management of super-grade plutonium in spent nuclear fuel

    International Nuclear Information System (INIS)

    McFarlane, H. F.; Benedict, R. W.

    2000-01-01

    This paper examines the security and safeguards implications of potential management options for DOE's sodium-bonded blanket fuel from the EBR-II and the Fermi-1 fast reactors. The EBR-II fuel appears to be unsuitable for the packaging alternative because of DOE's current safeguards requirements for plutonium. Emerging DOE requirements, National Academy of Sciences recommendations, draft waste acceptance requirements for Yucca Mountain and IAEA requirements for similar fuel also emphasize the importance of safeguards in spent fuel management. Electrometallurgical treatment would be acceptable for both fuel types. Meeting the known requirements for safeguards and security could potentially add more than $200M in cost to the packaging option for the EBR-II fuel

  18. Experimental Breeder Reactor-II automatic control-rod-drive system

    International Nuclear Information System (INIS)

    Christensen, L.J.

    1983-01-01

    A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

  19. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  20. dena grid study II. Integration of renewable energy sources into the German power supply system until 2020; dena - Netzstudie II. Integration erneuerbarer Energien in die deutsche Stromversorgung bis 2020

    Energy Technology Data Exchange (ETDEWEB)

    Agricola, Annegret C.; Seidl, Hannes; Zoch, Immo (comps.)

    2010-12-15

    With its broad system approach, dena Grid Study II describes new ways of developing Germany's integrated grid with regard to the expansion of renewable energy sources and taking not only an economically optimised conventional power generation system but also European power trading into account. Basis for detailed grid planning. Having investigated many different transmission technologies, grid management measures and options for increasing flexibility on both the supply and demand side, the study does much more than simply identify grid expansion requirements, it shows the path towards overall optimisation of the energy supply system. In addition to robustly identifying grid expansion requirements, dena Grid Study II provides a very sound basis for further detailed grid planning and the identification of specific power routes in the German power transmission grid. Increasing grid expansion requirements. dena Grid Study II investigates the expansion of electricity generation from renewable energy sources to 39 percent by 2020. However, this is just an interim goal. By as soon as 2030, the German Federal Government is planning to raise the proportion of renewable electricity generation to 50 percent. This shows that even after 2020, further adaptation of grid infrastructure will still be necessary. Practical action required. If the grid expansion required to integrate renewable energy sources is to be achieved rapidly, there is an urgent need to back up the grid expansion scenarios presented in dena Grid Study II with practical action. (orig.)

  1. Decontamination and decommissioning of the EBR-I complex. Topical report No. 3. NAK disposal pilot plant test

    International Nuclear Information System (INIS)

    Commander, J.C.; Lewis, L.; Hammer, R.

    1975-06-01

    Decontamination and decommissioning of the Experimental Breeder Reactor No. 1 (EBR-I) requires processing of the primary coolant, an eutectic solution of sodium and potassium (NaK), remaining in the EBR-I primary and secondary coolant systems. While developing design criteria for the NaK processing system, reasonable justification was provided for the development of a pilot test plant for field testing some of the process concepts and proposed hardware. The objective of this activity was to prove the process concept on a low-cost, small-scale test bed. The pilot test plant criteria provided a general description of the test including: the purpose, location, description of test equipment available, waste disposal requirements, and a flow diagram and conceptual equipment layout. The pilot plant test operations procedure provided a detailed step-by-step procedure for operation of the pilot plant to obtain the desired test data and operational experience. It also spelled out the safety precautions to be used by operating personnel, including the requirement for alkali metals training certification, use of protective clothing, availability of fire protection equipment, and caustic handling procedures. The pilot plant test was performed on May 16, 1974. During the test, 32.5 gallons or 240 lb of NaK was successfully converted to caustic by reaction with water in a caustic solution. (auth)

  2. LUCID Upgrade for ATLAS Luminosity Measurement in Run II

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00444244; The ATLAS collaboration

    2016-01-01

    The main ATLAS luminosity monitor, LUCID, and its read-out electronics have been completely rebuilt for the LHC Run II in order to cope with a higher center of mass energy ($\\sqrt{s}$=13 TeV) and the 25 ns bunch-spacing. The LUCID detector is measuring Cherenkov light produced in photomultiplier quartz windows and in quartz optical fibers. It has a novel calibration system that uses radioactive $^{207}$Bi sources that produce internal-conversion electrons with energy above the Cherenkov threshold in quartz. The new electronics can count signals with amplitude above a predefined threshold (hits) as well as the integrated pulseheight of the signals, which makes it possible to measure luminosity with complementary methods. The new detector, calibration system and electronics will be described, together with the results of the 2015 luminosity measurement.

  3. Spontaneous eye blink rate as predictor of dopamine-related cognitive function-A review.

    Science.gov (United States)

    Jongkees, Bryant J; Colzato, Lorenza S

    2016-12-01

    An extensive body of research suggests the spontaneous eye blink rate (EBR) is a non-invasive indirect marker of central dopamine (DA) function, with higher EBR predicting higher DA function. In the present review we provide a comprehensive overview of this literature. We broadly divide the available research in studies that aim to disentangle the dopaminergic underpinnings of EBR, investigate its utility in diagnosis of DA-related disorders and responsivity to drug treatment, and, lastly, investigate EBR as predictor of individual differences in DA-related cognitive performance. We conclude (i) EBR can reflect both DA receptor subtype D1 and D2 activity, although baseline EBR might be most strongly related to the latter, (ii) EBR can predict hypo- and hyperdopaminergic activity as well as normalization of this activity following treatment, and (iii) EBR can reliably predict individual differences in performance on many cognitive tasks, in particular those related to reward-driven behavior and cognitive flexibility. In sum, this review establishes EBR as a useful predictor of DA in a wide variety of contexts. Copyright © 2016 Elsevier Ltd. All rights reserved.

  4. Procedures and instrumentation for sodium boiling experiments in EBR-II

    International Nuclear Information System (INIS)

    Crowe, R.D.

    1976-01-01

    The development of instrumentation capable of detecting localized coolant boiling in a liquid metal cooled breeder reactor (LMFBR) has a high priority in fast reactor safety. The detection must be rapid enough to allow corrective action to be taken before significant damage occurs to the core. To develop and test a method of boiling detection, it is desirable to produce boiling in a reactor and thereby introduce a condition in the reactor the original design concepts were chosen to preclude. The proposed boiling experiments are designed to safely produce boiling in the subassembly of a fast reactor and provide the information to develop boiling detection instrumentation without core damage or safety compromise. The experiment consists of the operation of two separate subassemblies, first, a gamma heated boiling subassembly which produces non-typical but highly conservative boiling and then a fission heated subassembly which simulates a prototypical boiling event. The two boiling subassemblies are designed to operate in the instrumentation subassembly test facility (INSAT) of Experiment Breeder Reactor II

  5. Urban Integrated Industrial Cogeneration Systems Analysis. Phase II final report

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    Through the Urban Integrated Industrial Cogeneration Systems Analysis (UIICSA), the City of Chicago embarked upon an ambitious effort to identify the measure the overall industrial cogeneration market in the city and to evaluate in detail the most promising market opportunities. This report discusses the background of the work completed during Phase II of the UIICSA and presents the results of economic feasibility studies conducted for three potential cogeneration sites in Chicago. Phase II focused on the feasibility of cogeneration at the three most promising sites: the Stockyards and Calumet industrial areas, and the Ford City commercial/industrial complex. Each feasibility case study considered the energy load requirements of the existing facilities at the site and the potential for attracting and serving new growth in the area. Alternative fuels and technologies, and ownership and financing options were also incorporated into the case studies. Finally, site specific considerations such as development incentives, zoning and building code restrictions and environmental requirements were investigated.

  6. Thermal Expansion Property of U-Zr Alloys and U-Zr-Ce Alloys as a Surrogate Metallic Fuel for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Jong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Youn Myung; Lee, Chan Bock

    2010-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. An extensive database on the performance of advanced metal fuels was generated as a result of the operation of these reactors and the IFR program. In this study, the U-Zr binary alloys and U-Zr-Ce ternary alloys as surrogate metallic fuel were fabricated in lower pressure Ar environment by gravity casting. The melt temperature was approximately 1,500 .deg. C. Thermal expansion of the fuel during normal operation is related with fuel performance in a reactor. Therefore, it is necessary to investigate the thermal expansion of the fuel in order to warrant a good prediction the fuel performance

  7. Feasibility of calibration of liquid sodium flowmeters by neutron activation techniques

    International Nuclear Information System (INIS)

    Kehler, P.

    1976-07-01

    Velocities of fluids in pipes can be measured by injecting radioactive tracers into the fluid and recording the activity downstream of the injection point. One convenient method of injecting radioactive tracers is by neutron activation of the fluid itself. The present report describes a FORTRAN program that can be used for the prediction of the counting rates of fluid flow tests performed with a pulsed neutron source and a scintillation detector. The program models the flow profile and the mixing of the fluid, the attenuation of neutrons and gamma rays in the fluid, and the geometric arrangement of the source and the detector. Using this program, an experiment for the measurement of the secondary sodium flow of the EBR-II was optimized. A pulsed D,T neutron source and a 5 in. x 5 in. NaI detector will be used in the EBR-II test. Under optimized conditions, the expected accuracy of the flow measurement is about 2 percent

  8. Design and field measurement of the BEPC-II interaction region dual-aperture quadrupoles

    International Nuclear Information System (INIS)

    Yin, Z.S.; Wu, Y.Z.; Zhang, J.F.; Chen, W.; Li, Y.J.; Li, L.; Hou, R.; Yin, B.G.; Sun, X.J.; Ren, F.L.; Wang, F.A.; Chen, F.S.; Yu, C.H.; Chen, C.

    2007-01-01

    With the Beijing Electron Positron Collider upgrade project (BEPC-II), two dual-aperture septum-style quadrupole magnets are used in the interaction region for the final focusing of the electron and positron beams. The BEPC-II lattice design calls for the same high quality integral quadrupole field and large good field region in both apertures for each magnet. Two-dimensional contour optimization and pole-end chamfer iteration are used to minimize the systematic harmonic errors. Unexpected non-systematic errors induced by the unsymmetrical structure and the manufacturing errors are compensated with the pole-end shimming. Magnet measurements with rotating coils are performed to guide and confirm the magnet design. This paper discusses the design consideration, optimizing procedure and measurement results of these dual-aperture magnets

  9. Leach behavior and mechanical-integrity studies of irradiated Epicor-II waste products

    International Nuclear Information System (INIS)

    Barletta, R.E.; Swyler, K.J.; Chan, S.F.; Davis, R.E.

    1982-01-01

    The leachability of Cs and Sr from cement solidified ion exchange media claimed to be representative of the Epicor-II prefilters (D-mix) is presented. The Cs and Sr release is significantly lower than that typically observed for organic ion exchange resin/cement composites. The effect of radiation up to a total dose of 10 7 Gy upon the leachability and mechanical integrity (as measured by MCC-11) of D-mix/cement composites has been investigated. No deleterious effects were found. 6 figures

  10. Integration of Error Compensation of Coordinate Measuring Machines into Feature Measurement: Part I—Model Development

    Science.gov (United States)

    Calvo, Roque; D’Amato, Roberto; Gómez, Emilio; Domingo, Rosario

    2016-01-01

    The development of an error compensation model for coordinate measuring machines (CMMs) and its integration into feature measurement is presented. CMMs are widespread and dependable instruments in industry and laboratories for dimensional measurement. From the tip probe sensor to the machine display, there is a complex transformation of probed point coordinates through the geometrical feature model that makes the assessment of accuracy and uncertainty measurement results difficult. Therefore, error compensation is not standardized, conversely to other simpler instruments. Detailed coordinate error compensation models are generally based on CMM as a rigid-body and it requires a detailed mapping of the CMM’s behavior. In this paper a new model type of error compensation is proposed. It evaluates the error from the vectorial composition of length error by axis and its integration into the geometrical measurement model. The non-explained variability by the model is incorporated into the uncertainty budget. Model parameters are analyzed and linked to the geometrical errors and uncertainty of CMM response. Next, the outstanding measurement models of flatness, angle, and roundness are developed. The proposed models are useful for measurement improvement with easy integration into CMM signal processing, in particular in industrial environments where built-in solutions are sought. A battery of implementation tests are presented in Part II, where the experimental endorsement of the model is included. PMID:27690052

  11. Integration of Error Compensation of Coordinate Measuring Machines into Feature Measurement: Part I—Model Development

    Directory of Open Access Journals (Sweden)

    Roque Calvo

    2016-09-01

    Full Text Available The development of an error compensation model for coordinate measuring machines (CMMs and its integration into feature measurement is presented. CMMs are widespread and dependable instruments in industry and laboratories for dimensional measurement. From the tip probe sensor to the machine display, there is a complex transformation of probed point coordinates through the geometrical feature model that makes the assessment of accuracy and uncertainty measurement results difficult. Therefore, error compensation is not standardized, conversely to other simpler instruments. Detailed coordinate error compensation models are generally based on CMM as a rigid-body and it requires a detailed mapping of the CMM’s behavior. In this paper a new model type of error compensation is proposed. It evaluates the error from the vectorial composition of length error by axis and its integration into the geometrical measurement model. The non-explained variability by the model is incorporated into the uncertainty budget. Model parameters are analyzed and linked to the geometrical errors and uncertainty of CMM response. Next, the outstanding measurement models of flatness, angle, and roundness are developed. The proposed models are useful for measurement improvement with easy integration into CMM signal processing, in particular in industrial environments where built-in solutions are sought. A battery of implementation tests are presented in Part II, where the experimental endorsement of the model is included.

  12. Precise measurement of the $W$-boson mass with the CDF II detector

    Energy Technology Data Exchange (ETDEWEB)

    Aaltonen, T.; /Helsinki Inst. of Phys.; Alvarez Gonzalez, B.; /Oviedo U. /Cantabria Inst. of Phys.; Amerio, S.; /INFN, Padua; Amidei, D.; /Michigan U.; Anastassov, A.; /Northwestern U. /Fermilab; Annovi, A.; /Frascati; Antos, J.; /Comenius U.; Apollinari, G.; /Fermilab; Appel, J.A.; /Fermilab; Arisawa, T.; /Waseda U.; Artikov, A.; /Dubna, JINR /Texas A-M

    2012-03-01

    We have measured the W-boson mass M{sub W} using data corresponding to 2.2 fb{sup -1} of integrated luminosity collected in p{bar p} collisions at {radical}s = 1.96 TeV with the CDF II detector at the Fermilab Tevatron collider. Samples consisting of 470 126 W {yields} e{nu} candidates and 624 708 W {yields} {mu}{nu} candidates yield the measurement M{sub W} = 80 387 {+-} 12{sub stat} {+-} 15{sub syst} = 80 387 {+-} 19 MeV/c{sup 2}. This is the most precise measurement of the W-boson mass to date and significantly exceeds the precision of all previous measurements combined.

  13. Cross-cultural examination of measurement invariance of the Beck Depression Inventory-II.

    Science.gov (United States)

    Dere, Jessica; Watters, Carolyn A; Yu, Stephanie Chee-Min; Bagby, R Michael; Ryder, Andrew G; Harkness, Kate L

    2015-03-01

    Given substantial rates of major depressive disorder among college and university students, as well as the growing cultural diversity on many campuses, establishing the cross-cultural validity of relevant assessment tools is important. In the current investigation, we examined the Beck Depression Inventory-Second Edition (BDI-II; Beck, Steer, & Brown, 1996) among Chinese-heritage (n = 933) and European-heritage (n = 933) undergraduates in North America. The investigation integrated 3 distinct lines of inquiry: (a) the literature on cultural variation in depressive symptom reporting between people of Chinese and Western heritage; (b) recent developments regarding the factor structure of the BDI-II; and (c) the application of advanced statistical techniques to the issue of cross-cultural measurement invariance. A bifactor model was found to represent the optimal factor structure of the BDI-II. Multigroup confirmatory factor analysis showed that the BDI-II had strong measurement invariance across both culture and gender. In group comparisons with latent and observed variables, Chinese-heritage students scored higher than European-heritage students on cognitive symptoms of depression. This finding deviates from the commonly held view that those of Chinese heritage somatize depression. These findings hold implications for the study and use of the BDI-II, highlight the value of advanced statistical techniques such as multigroup confirmatory factor analysis, and offer methodological lessons for cross-cultural psychopathology research more broadly. 2015 APA, all rights reserved

  14. Argonne Fuel Cycle Facility ventilation system -- modeling and results

    International Nuclear Information System (INIS)

    Mohr, D.; Feldman, E.E.; Danielson, W.F.

    1995-01-01

    This paper describes an integrated study of the Argonne-West Fuel Cycle Facility (FCF) interconnected ventilation systems during various operations. Analyses and test results include first a nominal condition reflecting balanced pressures and flows followed by several infrequent and off-normal scenarios. This effort is the first study of the FCF ventilation systems as an integrated network wherein the hydraulic effects of all major air systems have been analyzed and tested. The FCF building consists of many interconnected regions in which nuclear fuel is handled, transported and reprocessed. The ventilation systems comprise a large number of ducts, fans, dampers, and filters which together must provide clean, properly conditioned air to the worker occupied spaces of the facility while preventing the spread of airborne radioactive materials to clean am-as or the atmosphere. This objective is achieved by keeping the FCF building at a partial vacuum in which the contaminated areas are kept at lower pressures than the other worker occupied spaces. The ventilation systems of FCF and the EBR-II reactor are analyzed as an integrated totality, as demonstrated. We then developed the network model shown in Fig. 2 for the TORAC code. The scope of this study was to assess the measured results from the acceptance/flow balancing testing and to predict the effects of power failures, hatch and door openings, single-failure faulted conditions, EBR-II isolation, and other infrequent operations. The studies show that the FCF ventilation systems am very controllable and remain stable following off-normal events. In addition, the FCF ventilation system complex is essentially immune to reverse flows and spread of contamination to clean areas during normal and off-normal operation

  15. Measure and integration theory

    CERN Document Server

    Burckel, Robert B

    2001-01-01

    This book gives a straightforward introduction to the field as it is nowadays required in many branches of analysis and especially in probability theory. The first three chapters (Measure Theory, Integration Theory, Product Measures) basically follow the clear and approved exposition given in the author's earlier book on ""Probability Theory and Measure Theory"". Special emphasis is laid on a complete discussion of the transformation of measures and integration with respect to the product measure, convergence theorems, parameter depending integrals, as well as the Radon-Nikodym theorem. The fi

  16. Conceptual design report: Decontamination and decommissioning of the EBR-1 Mark-2 NaK

    International Nuclear Information System (INIS)

    Brown, B.W.; La Rue, D.M.; Stoll, F.E.; Dolenc, M.R.; Crandall, D.L.

    1987-09-01

    A conceptual design of a processing system for approximately 180 gallons of contaminated NaK (sodium/potassium eutectic alloy) is presented. This NaK resulted from an incident at EBR-1 at the Idaho National Engineering Laboratory in 1955. The proposed method of decommissioning the NaK is to chemically deactivate it by combining it with gaseous chlorine. This process will produce a solid mass of potassium chloride and sodium chloride salts which will mitigate the consequences of further chemical reaction of the NaK should the storage containers be breached in any way. Following the processing of the NaK, the storage vessels will be transferred to an appropriate storage facility for radioactive wastes. 10 refs., 12 figs., 1 tab

  17. Correlation of creep and swelling with fuel pin performance

    International Nuclear Information System (INIS)

    Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

    1975-09-01

    The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

  18. Data package addendum for COBRA-1A2 life extension to 400 EFPD

    International Nuclear Information System (INIS)

    Hecht, S.L.; Ermi, A.M.

    1994-01-01

    The COBRA-1A experiment was originally designed for irradiations up to 350 effective full power days (EFPD) in EBR-II. Three of the seven B7A test capsules were discharged after 88.6 EFPD (COBRA-1A1; EBR-II designation X516), while the remaining four capsules continued to be irradiated to a goal exposure of 300 EFPD (COBRA-1A2; EBR-II designation X516A). However, it was recently decided that COBRA-1A2 was to remain in the reactor during Run 170, giving and nominal end-of-life (EOL) exposure of 375 EFPD. Since the revised test exposure exceeds the design basis given in supporting analyses, amended analyses are provided herein, giving the technical bases for the extended irradiation. This report describes the safety analysis for the extension of the COBRA-1A2 test (X516A) to 400 effective full power days in FBR-II

  19. Heysham II/Torness AGR core integrity

    International Nuclear Information System (INIS)

    Birch, A.L.; Hampson, J.D.

    1985-01-01

    The design and construction process for the Heysham II/Torness AGR core structures is presented. The design intent utilizing all past experience in designing and building AGR core structures is described. The major aspects of the design criteria and the design conditions are outlined to demonstrate how the integrity of the Heysham II/Torness core is assured. Since no recognized codes of practice for graphite core design exist, the National Nuclear Corporation (NNC) have conceived design criteria utilizing reserve factors based on their design experience. Target reserve factors are defined for particular loading conditions including the ultimate 'safe-shutdown earthquake'. The substantial programme of computer analysis and RandD work to substantiate the design, including seismic qualification, is described. In keeping with their responsibility for the detailed core structure design and the fuel path geometry (guide tube system), NNC attach great importance to design/manufacture/construction liaison, which is demonstrated in the quality assurance section. (author)

  20. Validation of SAGE II ozone measurements

    Science.gov (United States)

    Cunnold, D. M.; Chu, W. P.; Mccormick, M. P.; Veiga, R. E.; Barnes, R. A.

    1989-01-01

    Five ozone profiles from the Stratospheric Aerosol and Gas Experiment (SAGE) II are compared with coincident ozonesonde measurements obtained at Natal, Brazil, and Wallops Island, Virginia. It is shown that the mean difference between all of the measurements is about 1 percent and that the agreement is within 7 percent at altitudes between 20 and 53 km. Good agreement is also found for ozone mixing ratios on pressure surfaces. It is concluded that the SAGE II profiles provide useful ozone information up to about 60 km altitude.

  1. Fast neutron irradiation results on Li2O, Li4SiO4, Li2ZrO3 and LiA102

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1983-04-01

    Several ceramic solid breeder materials were irradiated in the fast neutron flux of EBR-II to a burnup of 3 x 10 20 captures/cm 3 in the 500 0 C to 900 0 C temperature range. Performance data were obtained on structural integrity, lithium transport, pellet swelling, and grain growth. The data provide a basis for more accurately forecasting the feasibility and performance limitations for these materials

  2. Displacement cross sections and PKA spectra: tables and applications

    International Nuclear Information System (INIS)

    Doran, D.G.; Graves, N.J.

    1976-12-01

    Damage energy cross sections to 20 MeV are given for aluminum, vanadium, chromium, iron, nickel, copper, zirconium, niobium, molybdenum, tantalum, tungsten, lead, and 18Cr10Ni stainless steel. They are based on ENDF/B-IV nuclear data and the Lindhard energy partition model. Primary knockon atom (PKA) spectra are given for aluminum, iron, niobium, tantalum, and lead for neutron energies up to 15 MeV at approximately one-quarter lethargy intervals. The contributions of various reactions to both the displacement cross sections (taken to be proportional to the damage energy cross sections) and the PKA spectra are presented graphically. Spectral-averaged values of the displacement cross sections are given for several spectra, including approximate maps for the Experimental Breeder Reactor-II (EBR-II) and several positions in the Fast Test Reactor (FTR). Flux values are included to permit estimation of displacement rates. Graphs show integral PKA spectra for the five metals listed above for neutron spectra corresponding to locations in the EBR-II, the High Flux Isotope Reactor (HFIR), and a conceptual fusion reactor (UWMAK-I). Detailed calculations are given only for cases not previously documented. Uncertainty estimates are included

  3. Filtration experiments of the KNK II primary sodium

    International Nuclear Information System (INIS)

    Stamm, H.H.

    1987-01-01

    The separated particles of the KNK primary sodium are a result of a normal corrosion rate by using ferritic and austenitic steels in sodium. Similar particle distributions and concentrations were found in primary circuits of other breeder plants (EBR-II et al). The experimental results of the particle concentration in the KNK primary sodium were lower than the theoretical calculation. Based on a corrosion rate of 0.5 micron/y for austenitic steels and a corrosion rate of 1.5 mg/square-cm y for a ferritic steel used in KNK as structure material an equilibrium particle concentration of 1.066 mg/kg Na was calculated. The experimental results of particle size are not in agreement with theoretical calculation of the BACCHUS code. (orig.)

  4. EBR Strengthening Technique for Concrete, Long-Term Behaviour and Historical Survey

    Directory of Open Access Journals (Sweden)

    Christoph Czaderski

    2018-01-01

    Full Text Available Epoxy bonded steel plates (externally bonded reinforcemen: EBR for the strengthening of concrete structures were introduced to the construction industry in the late 1960s, and the use of fibre reinforced polymers (FRPs was introduced in the 1990s, which means that these techniques have already been used in construction for 50 and 25 years, respectively. In the first part of the paper, a historical survey of the development and introduction of these strengthening techniques into the construction industry are presented. The monitoring of such applications in construction is very important and gives more confidence to this strengthening technique. Therefore, in the second part of the paper, two long-term monitoring campaigns over an extraordinarily long duration will be presented. Firstly, a 47-year monitoring campaign on a concrete beam with an epoxy bonded steel plate and, secondly, a 20-year monitoring campaign on a road bridge with epoxy bonded CFRP (carbon fibre reinforced polymers strips are described. The paper is an expanded version of the paper presented at the SMAR2017 Conference.

  5. Development of 4S and related technologies (2). Long life metallic fuel

    International Nuclear Information System (INIS)

    Yacout, A.M.; Tsuboi, Y.; Ueda, N.

    2009-01-01

    This paper provides an overview of the long life metallic fuel to be used in the 4S reactor. The 4S fuel design is presented and implications of its characteristics on fuel performance are discussed. Main design characteristics include the long fuel life time of 30 years and the wider and longer fuel pins compared to EBR-II and FFTF fuel pins. The LIFE-METAL fuel performance code was used to evaluate the performance of the 4S fuel design. The code has been validated using post irradiation examination data of metallic fuel irradiated in EBR-II. The performance evaluation shows the benign nature of the design. The design enables the fuel to perform adequately during reactor operations without violating any of a conservative set of steady state design criteria. A survey evaluation of the fuel performance is also presented. This performance bounding evaluation took into account possible fuel swelling behavior and cladding temperature range that represents worst case scenarios. The evaluation showed that the fuel maintains its integrity even under those worst case conditions. (author)

  6. Calculation of displacement and helium production at the LAMPF irradiation facility

    International Nuclear Information System (INIS)

    Wechsler, M.S.; Davidson, D.R.; Sommer, W.F.; Greenwood, L.R.

    1985-01-01

    Differential and total displacement and helium-production rates are calculated for copper irradiated by spallation neutrons and 760-MeV protons at LAMPF. The calculations are performed using the SPECTOR and VNMTC computer codes, the latter being specially designed for spallation radiation-damage calculations. For comparison, similar SPECTER calculations are also described for irradiation of copper in the experimental breeder reactor (EBR-II) at the Argonne National Laboratory-West in Idaho, and in the rotating target neutron source (RTNS-II) at Lawrence Livermore Laboratory. The neutron energy spectra for LAMPF, EBR-II, and RTNS-II and the displacement and helium-production cross sections are shown

  7. Instruments Measuring Integrated Care: A Systematic Review of Measurement Properties.

    Science.gov (United States)

    Bautista, Mary Ann C; Nurjono, Milawaty; Lim, Yee Wei; Dessers, Ezra; Vrijhoef, Hubertus Jm

    2016-12-01

    Policy Points: Investigations on systematic methodologies for measuring integrated care should coincide with the growing interest in this field of research. A systematic review of instruments provides insights into integrated care measurement, including setting the research agenda for validating available instruments and informing the decision to develop new ones. This study is the first systematic review of instruments measuring integrated care with an evidence synthesis of the measurement properties. We found 209 index instruments measuring different constructs related to integrated care; the strength of evidence on the adequacy of the majority of their measurement properties remained largely unassessed. Integrated care is an important strategy for increasing health system performance. Despite its growing significance, detailed evidence on the measurement properties of integrated care instruments remains vague and limited. Our systematic review aims to provide evidence on the state of the art in measuring integrated care. Our comprehensive systematic review framework builds on the Rainbow Model for Integrated Care (RMIC). We searched MEDLINE/PubMed for published articles on the measurement properties of instruments measuring integrated care and identified eligible articles using a standard set of selection criteria. We assessed the methodological quality of every validation study reported using the COSMIN checklist and extracted data on study and instrument characteristics. We also evaluated the measurement properties of each examined instrument per validation study and provided a best evidence synthesis on the adequacy of measurement properties of the index instruments. From the 300 eligible articles, we assessed the methodological quality of 379 validation studies from which we identified 209 index instruments measuring integrated care constructs. The majority of studies reported on instruments measuring constructs related to care integration (33%) and patient

  8. Instruments Measuring Integrated Care: A Systematic Review of Measurement Properties

    Science.gov (United States)

    BAUTISTA, MARY ANN C.; NURJONO, MILAWATY; DESSERS, EZRA; VRIJHOEF, HUBERTUS JM

    2016-01-01

    Policy Points: Investigations on systematic methodologies for measuring integrated care should coincide with the growing interest in this field of research.A systematic review of instruments provides insights into integrated care measurement, including setting the research agenda for validating available instruments and informing the decision to develop new ones.This study is the first systematic review of instruments measuring integrated care with an evidence synthesis of the measurement properties.We found 209 index instruments measuring different constructs related to integrated care; the strength of evidence on the adequacy of the majority of their measurement properties remained largely unassessed. Context Integrated care is an important strategy for increasing health system performance. Despite its growing significance, detailed evidence on the measurement properties of integrated care instruments remains vague and limited. Our systematic review aims to provide evidence on the state of the art in measuring integrated care. Methods Our comprehensive systematic review framework builds on the Rainbow Model for Integrated Care (RMIC). We searched MEDLINE/PubMed for published articles on the measurement properties of instruments measuring integrated care and identified eligible articles using a standard set of selection criteria. We assessed the methodological quality of every validation study reported using the COSMIN checklist and extracted data on study and instrument characteristics. We also evaluated the measurement properties of each examined instrument per validation study and provided a best evidence synthesis on the adequacy of measurement properties of the index instruments. Findings From the 300 eligible articles, we assessed the methodological quality of 379 validation studies from which we identified 209 index instruments measuring integrated care constructs. The majority of studies reported on instruments measuring constructs related to care

  9. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

  10. Neural networks for sensor validation and plant monitoring

    International Nuclear Information System (INIS)

    Upadhyaya, B.R.; Eryurek, E.; Mathai, G.

    1990-01-01

    Sensor and process monitoring in power plants require the estimation of one or more process variables. Neural network paradigms are suitable for establishing general nonlinear relationships among a set of plant variables. Multiple-input multiple-output autoassociative networks can follow changes in plant-wide behavior. The backpropagation algorithm has been applied for training feedforward networks. A new and enhanced algorithm for training neural networks (BPN) has been developed and implemented in a VAX workstation. Operational data from the Experimental Breeder Reactor-II (EBR-II) have been used to study the performance of BPN. Several results of application to the EBR-II are presented

  11. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  12. Measuring integrated care.

    Science.gov (United States)

    Strandberg-Larsen, Martin

    2011-02-01

    The positive outcomes of coordination of healthcare services are to an increasing extent becoming clear. However the complexity of the field is an inhibiting factor for vigorously designed trial studies. Conceptual clarity and a consistent theoretical frame-work are thus needed. While researchers respond to these needs, patients and providers face the multiple challenges of today's healthcare environment. Decision makers, planners and managers need evidence based policy options and information on the scope of the integrated care challenges they are facing. The US managed care organization Kaiser Permanente has been put forward as an example for European healthcare systems to follow, although the evidence base is far from conclusive. The thesis has five objectives: 1) To contribute to the understanding of the concept of integration in healthcare systems and to identify measurement methods to capture the multi-dimensional aspects of integrated healthcare delivery. 2) To assess the level of integration of the Danish healthcare system. 3) To assess the use of joint health plans as a tool for coordination between the regional and local level in the Danish healthcare system. 4) To compare the inputs and performance of the Danish healthcare system and the managed care organization Kaiser Permanente, California, US. 5) To compare primary care clinicians' perception of clinical integration in two healthcare systems: Kaiser Permanente, Northern California and the Danish healthcare system. Further to examine the associations between specific organizational factors and clinical integration within each system. The literature was systematically searched to identify methods for measurement of integrated healthcare delivery. A national cross-sectional survey was conducted among major professional stake-holders at five different levels of the Danish healthcare system. The survey data were used to allow for analysis of the level of integration achieved. Data from the survey were

  13. Quantum measure and integration theory

    International Nuclear Information System (INIS)

    Gudder, Stan

    2009-01-01

    This article begins with a review of quantum measure spaces. Quantum forms and indefinite inner-product spaces are then discussed. The main part of the paper introduces a quantum integral and derives some of its properties. The quantum integral's form for simple functions is characterized and it is shown that the quantum integral generalizes the Lebesgue integral. A bounded, monotone convergence theorem for quantum integrals is obtained and it is shown that a Radon-Nikodym-type theorem does not hold for quantum measures. As an example, a quantum-Lebesgue integral on the real line is considered.

  14. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

    Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

    1996-01-01

    The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

  15. Measuring the degree of integration for an integrated service network

    Directory of Open Access Journals (Sweden)

    Chenglin Ye

    2012-09-01

    Full Text Available Background: Integration involves the coordination of services provided by autonomous agencies and improves the organization and delivery of multiple services for target patients. Current measures generally do not distinguish between agencies' perception and expectation. We propose a method for quantifying the agencies' service integration. Using the data from the Children's Treatment Network (CTN, we aimed to measure the degree of integration for the CTN agencies in York and Simcoe.  Theory and Methods: We quantified the integration by the agreement between perceived and expected levels of involvement and calculated four scores from different perspectives for each agency. We used the average score to measure the global network integration and examined the sensitivity of the global score.  Results: Most agencies' integration scores were less than 65%. As measured by the agreement between every other agency's perception and expectation, the overall integration of CTN in Simcoe and York was 44% (95% CI: 39% - 49% and 52% (95% CI: 48% - 56%, respectively. The sensitivity analysis showed that the global scores were robust.  Conclusion: Our method extends existing measures of integration and possesses a good extent of validity. We can also apply the method in monitoring improvement and linking integration with other outcomes.

  16. Measuring the degree of integration for an integrated service network

    Directory of Open Access Journals (Sweden)

    Chenglin Ye

    2012-09-01

    Full Text Available Background: Integration involves the coordination of services provided by autonomous agencies and improves the organization and delivery of multiple services for target patients. Current measures generally do not distinguish between agencies' perception and expectation. We propose a method for quantifying the agencies' service integration. Using the data from the Children's Treatment Network (CTN, we aimed to measure the degree of integration for the CTN agencies in York and Simcoe. Theory and Methods: We quantified the integration by the agreement between perceived and expected levels of involvement and calculated four scores from different perspectives for each agency. We used the average score to measure the global network integration and examined the sensitivity of the global score. Results: Most agencies' integration scores were less than 65%. As measured by the agreement between every other agency's perception and expectation, the overall integration of CTN in Simcoe and York was 44% (95% CI: 39% - 49% and 52% (95% CI: 48% - 56%, respectively. The sensitivity analysis showed that the global scores were robust. Conclusion: Our method extends existing measures of integration and possesses a good extent of validity. We can also apply the method in monitoring improvement and linking integration with other outcomes. 

  17. Non-additive measure and integral

    CERN Document Server

    Denneberg, Dieter

    1994-01-01

    Non-Additive Measure and Integral is the first systematic approach to the subject. Much of the additive theory (convergence theorems, Lebesgue spaces, representation theorems) is generalized, at least for submodular measures which are characterized by having a subadditive integral. The theory is of interest for applications to economic decision theory (decisions under risk and uncertainty), to statistics (including belief functions, fuzzy measures) to cooperative game theory, artificial intelligence, insurance, etc. Non-Additive Measure and Integral collects the results of scattered and often isolated approaches to non-additive measures and their integrals which originate in pure mathematics, potential theory, statistics, game theory, economic decision theory and other fields of application. It unifies, simplifies and generalizes known results and supplements the theory with new results, thus providing a sound basis for applications and further research in this growing field of increasing interest. It also co...

  18. Hot Fuel Examination Facility/South

    Energy Technology Data Exchange (ETDEWEB)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs.

  19. Hot Fuel Examination Facility/South

    International Nuclear Information System (INIS)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs

  20. Measurement of the inclusive jet cross section using the midpoint algorithm in Run II at CDF

    Energy Technology Data Exchange (ETDEWEB)

    Group, Robert Craig [Univ. of Florida, Gainesville, FL (United States)

    2006-01-01

    A measurement is presented of the inclusive jet cross section using the Midpoint jet clustering algorithm in five different rapidity regions. This is the first analysis which measures the inclusive jet cross section using the Midpoint algorithm in the forward region of the detector. The measurement is based on more than 1 fb-1 of integrated luminosity of Run II data taken by the CDF experiment at the Fermi National Accelerator Laboratory. The results are consistent with the predictions of perturbative quantum chromodynamics.

  1. Fuel cycle and waste management demonstration in the IFR Program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Benedict, R.W.; Laidler, J.J.; Battles, J.E.; Miller, W.E.

    1992-01-01

    Argonne's National Laboratory's Integral Fast Reactor (IFR) is the main element in the US advanced reactor development program. A unique fuel cycle and waste process technology is being developed for the IFR. Demonstration of this technology at engineering scale will begin within the next year at the EBR-II test facility complex in Idaho. This paper describes the facility being readied for this demonstration, the process to be employed, the equipment being built, and the waste management approach

  2. High precision measurements of hyperfine structure in Tm II, N2+ and Sc II

    International Nuclear Information System (INIS)

    Mansour, N.B.; Dinneen, T.P.; Young, L.

    1988-01-01

    We have applied the technique of collinear fast-ion-beam laser spectroscopy to measure the hyperfine structure (hfs) in Sc II, Tm II and N 2 + . Laser induced fluorescence was observed from a 50 keV ion beam which was superimposed with the output of an actively stabilized ring dye laser (rms bandwidth 2 and the excited 3d4p configuration of Sc II and in the 4f 13 6s and 4f 13 5d configurations of the Tm II. The fine and hyperfine structure of N 2 + has been observed in the (0,1) and (1,2) band of B 2 Σ/sub u/ + /minus/X 2 Σ/sub g/ + system. Higher resolution measurements of the metastable 3d 2 configuration in Sc II were also made by laser-rf double resonance. The experimental results will be compared with those obtained by multiconfiguration Hartree-Fock ab-initio calculations. 15 refs., 4 figs., 5 tabs

  3. Automated reasoning applications to design validation and sneak function analysis

    International Nuclear Information System (INIS)

    Stratton, R.C.

    1984-01-01

    Argonne National Laboratory (ANL) is actively involved in the LMFBR Man-Machine Integration (MMI) Safety Program. The objective of this program is to enhance the operational safety and reliability of fast-breeder reactors by optimum integration of men and machines through the application of human factors principles and control engineering to the design, operation, and the control environment. ANL is developing methods to apply automated reasoning and computerization in the validation and sneak function analysis process. This project provides the element definitions and relations necessary for an automated reasoner (AR) to reason about design validation and sneak function analysis. This project also provides a demonstration of this AR application on an Experimental Breeder Reactor-II (EBR-II) system, the Argonne Cooling System

  4. Silicon Photonics II Components and Integration

    CERN Document Server

    Lockwood, David J

    2011-01-01

    This book is volume II of a series of books on silicon photonics. It gives a fascinating picture of the state-of-the-art in silicon photonics from a component perspective. It presents a perspective on what can be expected in the near future. It is formed from a selected number of reviews authored by world leaders in the field, and is written from both academic and industrial viewpoints. An in-depth discussion of the route towards fully integrated silicon photonics is presented. This book will be useful not only to physicists, chemists, materials scientists, and engineers but also to graduate students who are interested in the fields of micro- and nanophotonics and optoelectronics.

  5. Comparison of methods for measuring flux gradients in type II superconductors

    International Nuclear Information System (INIS)

    Kroeger, D.M.; Koch, C.C.; Charlesworth, J.P.

    1975-01-01

    A comparison has been made of four methods of measuring the critical current density J/sub c/ in hysteretic type II superconductors, having a wide range of K and J/sub c/ values, in magnetic fields up to 70 kOe. Two of the methods, (a) resistive measurements and (b) magnetization measurements, were carried out in static magnetic fields. The other two methods involved analysis of the response of the sample to a small alternating field superimposed on the static field. The response was analyzed either (c) by measuring the third-harmonic content or (d) by integration of the waveform to obtain measure of flux penetration. The results are discussed with reference to the agreement between the different techniques and the consistency of the critical state hypothesis on which all these techniques are based. It is concluded that flux-penetration measurements by method (d) provide the most detailed information about J/sub c/ but that one must be wary of minor failures of the critical state hypothesis. Best results are likely to be obtained by using more than one method. (U.S.)

  6. Results from an in-plant demonstration of intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Messick, N.

    1993-01-01

    A learning systems-based reconfigurable controller was demonstrated on the deaerating feedwater heater at the Experimental Breeder Reactor II (EBR-II) on April 1, 1993. Failures of the normal pressure regulating process were introduced by reducing the steam flow to the heater by as much as 10%. The controller maintained pressure in the heater at acceptable levels for several minutes, whereas operator intervention would have otherwise been required within a few seconds. This experiment demonstrates the potential of advanced control techniques for improving safety, reliability, and performance of power plant operations as well as the utility of EBR-II as an experimental power plant controls facility

  7. Verification and implications of the multiple pin treatment in the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1994-01-01

    As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple pin treatment for the analysis of intra-subassembly thermal hydraulics has been included in the SASSYS-1 liquid metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemlies in a reactor, rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the Shutdown Heat Removal Tests in the EBR-II Reactor. Detailed multiple pin calculations for blanket subassemblies in the EBR-II reactor demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models

  8. Measuring integrated care

    DEFF Research Database (Denmark)

    Strandberg-Larsen, Martin

    2011-01-01

    respond to these needs, patients and providers face the multiple challenges of today's healthcare environment. Decision makers, planners and managers need evidence based policy options and information on the scope of the integrated care challenges they are facing. The US managed care organization Kaiser...... differences were found in the perception of clinical integration in the two settings. More primary care clinicians in the Northern California region of Kaiser Permanente reported being part of a clinical integrated environment than did Danish general practitioners. By measuring the level of clinical...... and performance of the Danish healthcare system and the managed care organization Kaiser Permanente, California, US. 5) To compare primary care clinicians' perception of clinical integration in two healthcare systems: Kaiser Permanente, Northern California and the Danish healthcare system. Further to examine...

  9. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  10. Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state.

  11. Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

    International Nuclear Information System (INIS)

    Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok

    2015-01-01

    Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state

  12. The graphics-based human interface to the DISYS diagnostic/control guidance system at EBR-2

    International Nuclear Information System (INIS)

    Edwards, R.M.; Chavez, C.; Kamarthi, S.; Dharap, S.; Lindsay, R.W.; Staffon, J.

    1990-01-01

    An initial graphics based interface to the real-time DISYS diagnostic system has been developed using the multi-tasking capabilities of the UNIX operating system and X-Windows 11 Xlib graphics library. This system is interfaced to live plant data at the Experimental Breeder Reactor (EBR-2) for the Argon Cooling System of fuel handling operations and the steam plant. The interface includes an intelligent process schematic which highlights problematic components and sensors based on the results of the diagnostic computations. If further explanation of a faulted component is required, the user can call up a display of the diagnostic computations presented in a tree-like diagram. Numerical data on the process schematic and optional diagnostic tree are updated as new real-time data becomes available. The initial X-Windows 11 based interface will be further enhanced using VI Corporation DATAVIEWS graphical data base software. 5 refs., 6 figs

  13. Progress and status of the integral fast reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

  14. Fuzzy measures and integrals

    Czech Academy of Sciences Publication Activity Database

    Mesiar, Radko

    2005-01-01

    Roč. 28, č. 156 (2005), s. 365-370 ISSN 0165-0114 R&D Projects: GA ČR(CZ) GA402/04/1026 Institutional research plan: CEZ:AV0Z10750506 Keywords : fuzzy measures * fuzzy integral * regular fuzzy integral Subject RIV: BA - General Mathematics Impact factor: 1.039, year: 2005

  15. Pulsed neutron measurement of single and two-phase liquid flow

    International Nuclear Information System (INIS)

    Kehler, P.

    1978-01-01

    Use of radioactive tracers for flow velocity measurements is well developed and documented. Measurement techniques involving pulsed sources of fast (14 MeV) neutrons for in-situ production of tracers can be considered as extensions of the old methods. Improvements offered by these Pulsed Neutron Activation (PNA) techniques over conventional radioisotope techniques are (1) non-intrusion into the system, (2) easier introduction and better mixing of the tracer, and (3) no requirement to handle large amounts of relatively long lived radioactive materials. Just as in conventional tracer techniques, flow velocity measurements by PNA methods can be based on the transit-time or the total-count method. A very significant difference of the PNA technique from conventional methods is that the induced activity is proportional to the density of the fluid, and that PNA techniques can be used for density measurements (of two-phase flows) in addition to flow velocity measurement. Original equations were derived that relate experimental data to the mass flow velocity and the average density. The accuracy of these equations is not effected by the flow regime. Experimental results are presented for tests performed on liquid sodium loops, on air--water loops, on the EBR-II reactor and on the LOFT reactor. Current instrumentation development programs (detectors, pulsed neutron sources) are discussed

  16. Use of a pattern recognition scheme to compensate for critical sensor failures

    International Nuclear Information System (INIS)

    Singer, R.M.; King, R.W.; Mott, J.

    1989-01-01

    A general mathematical approach (the SSA code) for embodying learned data from a complex system and combining it with a current observation to estimate the true current state of the system has been applied to a nuclear power plant (EBR-II). Sensor validation and generation of estimates signals based upon the plant operating state are used for replacement of signals from multiple faulted sensors on a near real-time basis. A direct experimental demonstration of the capability of the code to perform these tasks is presented in which multiple sensor faults in EBR-II are simulated. 2 refs., 4 figs

  17. 2 μm wavelength range InP-based type-II quantum well photodiodes heterogeneously integrated on silicon photonic integrated circuits.

    Science.gov (United States)

    Wang, Ruijun; Sprengel, Stephan; Muneeb, Muhammad; Boehm, Gerhard; Baets, Roel; Amann, Markus-Christian; Roelkens, Gunther

    2015-10-05

    The heterogeneous integration of InP-based type-II quantum well photodiodes on silicon photonic integrated circuits for the 2 µm wavelength range is presented. A responsivity of 1.2 A/W at a wavelength of 2.32 µm and 0.6 A/W at 2.4 µm wavelength is demonstrated. The photodiodes have a dark current of 12 nA at -0.5 V at room temperature. The absorbing active region of the integrated photodiodes consists of six periods of a "W"-shaped quantum well, also allowing for laser integration on the same platform.

  18. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  19. Method for the determination of technical specifications limiting temperature in EBR-II operation

    International Nuclear Information System (INIS)

    Chang, L.K.; Hill, D.J.; Ku, J.Y.

    1994-01-01

    The methodology and analysis procedure to qualify the Mark-V and Mark-VA fuels for the Experimental Breeder Reactor II are summarized in this paper. Fuel performance data and design safety criteria are essential for thermal-hydraulic analysis and safety evaluations. Normal and off-normal operation duty cycles and transient classifications are required for the safety assessment of the fuels. The temperature limits of subassemblies were first determined by a steady-state thermal-structural and fuel damage analysis, in which a trial-and-error approach was used to predict the maximum allowable fuel pin temperature that satisfies the design criteria for steady-state normal operation. The steady-state temperature limits were used as the basis of the off-normal transient analysis to assess the safety performance of the fuel for anticipated, unlikely and extremely unlikely events. If the design criteria for the off-normal events are not satisfied, then the subassembly temperature limit is reduced and an iterative procedure is employed until all design criteria are met

  20. Comparison of integral cross section values of several cross section libraries in the SAND-II format

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.

    1978-01-01

    A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

  1. The Community Integration Questionnaire - Revised: Australian normative data and measurement of electronic social networking.

    Science.gov (United States)

    Callaway, Libby; Winkler, Dianne; Tippett, Alice; Herd, Natalie; Migliorini, Christine; Willer, Barry

    2016-06-01

    Consideration of the relationship between meaningful participation, health and wellbeing underpins occupational therapy intervention, and drives measurement of community integration following acquired brain injury (ABI). However, utility of community integration measures has been limited to date by lack of normative data against which to compare outcomes, and none examine the growing use of electronic social networking (ESN) for social participation. This research had four aims: (i) develop and pilot items assessing ESN to add to the Community Integration Questionnaire, producing the Community Integration Questionnaire-Revised (CIQ-R); (ii) examine factor structure of the CIQ-R; (iii) collect Australian CIQ-R normative data; and (iv) assess test-retest reliability of the revised measure. Australia. A convenience sample of adults without ABI (N = 124) was used to develop and pilot ESN items. A representative general population sample of adults without ABI aged 18-64 years (N = 1973) was recruited to gather normative CIQ-R data. Cross-sectional survey. Demographic items and the CIQ-R. The CIQ-R demonstrated acceptable psychometric properties, with minor modification to the original scoring based on the factor analyses provided. Large representative general population CIQ-R normative data have been established, detailing contribution of a range of independent demographic variables to community integration. The addition of electronic social networking items to the CIQ-R offers a contemporary method of assessing community integration following ABI. Normative CIQ-R data enhance the understanding of community integration in the general population, allowing occupational therapists and other clinicians to make more meaningful comparisons between groups. © 2016 Occupational Therapy Australia.

  2. Results of acoustic measurements with an electric boiling generator at KNK II

    International Nuclear Information System (INIS)

    Aberle, J.

    1987-08-01

    With regard to an integral core surveillance in sodium-cooled breeder reactors acoustic measurement techniques are under development. To determine experimentally the acoustic transfer function of a reactor core and to demonstrate the detectability of local sodium boiling, experiments with a so-called Boiling Generator were carried out in the KNK II reactor. The main part of this Boiling Generator was an electrically heated pin bundle which was equipped with a local blockage to obtain cooling disturbances. In this report the results of the acoustic measurements carried out with the Boiling Generator are presented. Main topic of the evaluation is the determination of the acoustic transfer function between the core and the upper sodium plenum. The signal conditioning necessary prior to this investigation is also explained. Great effort was required to suppress electrical disturbances which superimposed the acoustic signals and could not be eliminated by the hardware during the experiments. Finally, the detectability of local boiling using acoustic measurements is considered

  3. Factors controlling metal fuel lifetime

    International Nuclear Information System (INIS)

    Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

    1986-01-01

    The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly

  4. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    Energy Technology Data Exchange (ETDEWEB)

    Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna (Austria); Jaćimović, Radojko [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Gregoire, Gilles; Destouches, Christophe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St. Paul-Lez-Durance (France)

    2016-12-21

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

  5. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    International Nuclear Information System (INIS)

    Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

    2016-01-01

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

  6. RBE of Monoenergetic Fast Neutrons: Cytogenetic Effects in Maize; EBR des Neutrons Rapides Monoenergeniques: Effets Cytogenetiques sur le Mais; Obeh monoehnergeticheskikh bystrykh nejtronov: tsitogeneticheskie izmeneniya u kukuruzy; EBR de los Neutrones Rapidos Monoenergeticos: Efectos Citogeneticos en el Maiz

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H. H.; Bateman, J. L.; Quastler, H.; Rossi, H. H. [Biology and Medical Departments, Brookhaven National Laboratory, Upton, NY (United States)

    1964-05-15

    The maize used in these experiments has the advantage for RBE studies of yielding a basically first-order dose-response curve with low as well as with high LET radiations. An exposure apparatus was used which produced essentially equal dose rates in five rings of seeds placed so as to intercept neutrons of 0.43, 0.65, 1.00, 1.50 and 1.80 MeV. The mutant sectors produced in leaves are believed to be due mostly to simple chromosome breakage and deletion. Experiments were performed at dosages that gave responses which were linear, below saturation levels, and overlapping in range for the monoenergetic fast neutrons and 250 kVp X-rays. RBE values, calculated from relative slopes of linear-regression lines for neutrons and X-rays ranged from 42 to 135 with an overall average of about 70. Fast neutrons of 0.43 MeV energy were the most efficient, of those used, in producing g{sub 2} mutant sectors. (author) [French] Pour les etudes sur l'EBR, le maft utilise au cours des experiences offre l'avantage de donner une courbe dose-reponse qui est essentiellement de premier ordre, que le TLE du rayonnement soit faible ou eleve. Les auteurs ont utilise un appareil d'irradiation assurant des debits de dose pratiquement egaux dans cinq couronnes de semences disposees de maniere a intercepter les neutrons de 0,43, 0,65, 1,00, 1,50 et 1,80 MeV. On pense que les secteurs mutants produits dans les feuilles sont dus essentiellement a une rupture et une disparition de chromosomes simples. Les auteurs ont fait des experiences a des doses qui ont donne des reponses lineaires, inferieures au niveau de saturation et se chevauchant dans le cas des neutrons rapides et des rayons X de 250 kV-crete. Les valeurs de l'EBR calculees d'apres la comparaison des pentes des courbes de regression lineaire pour les neutrons et les rayons X varient de 42 a 135, avec une valeur moyenne d'environ 70. Parmi les neutrons utilises, les neutrons rapides de 0,43 MeV ont ete les plus efficaces pour produire des

  7. GLOBULAR CLUSTER ABUNDANCES FROM HIGH-RESOLUTION, INTEGRATED-LIGHT SPECTROSCOPY. II. EXPANDING THE METALLICITY RANGE FOR OLD CLUSTERS AND UPDATED ANALYSIS TECHNIQUES

    Energy Technology Data Exchange (ETDEWEB)

    Colucci, Janet E.; Bernstein, Rebecca A.; McWilliam, Andrew [The Observatories of the Carnegie Institution for Science, 813 Santa Barbara St., Pasadena, CA 91101 (United States)

    2017-01-10

    We present abundances of globular clusters (GCs) in the Milky Way and Fornax from integrated-light (IL) spectra. Our goal is to evaluate the consistency of the IL analysis relative to standard abundance analysis for individual stars in those same clusters. This sample includes an updated analysis of seven clusters from our previous publications and results for five new clusters that expand the metallicity range over which our technique has been tested. We find that the [Fe/H] measured from IL spectra agrees to ∼0.1 dex for GCs with metallicities as high as [Fe/H] = −0.3, but the abundances measured for more metal-rich clusters may be underestimated. In addition we systematically evaluate the accuracy of abundance ratios, [X/Fe], for Na i, Mg i, Al i, Si i, Ca i, Ti i, Ti ii, Sc ii, V i, Cr i, Mn i, Co i, Ni i, Cu i, Y ii, Zr i, Ba ii, La ii, Nd ii, and Eu ii. The elements for which the IL analysis gives results that are most similar to analysis of individual stellar spectra are Fe i, Ca i, Si i, Ni i, and Ba ii. The elements that show the greatest differences include Mg i and Zr i. Some elements show good agreement only over a limited range in metallicity. More stellar abundance data in these clusters would enable more complete evaluation of the IL results for other important elements.

  8. Stability of lithium niobate on irradiation at elevated temperature

    International Nuclear Information System (INIS)

    Primak, W.; Gavin, A.P.; Anderson, T.T.; Monahan, E.

    1977-01-01

    In contrast to results obtained for neutron irradiation in a thermal reactor near room temperature, lithium niobate plates irradiated in the Experimental Breeder Reactor II (EBR-II) did not become metamict. This is attributed to the elevated temperature of the EBR-II. Ion bombardment experiments indicate that to avoid disordering of lithium niobate on irradiation, its temperature should be maintained above 673 K. Evidence for ionic conductivity was found at 873 K, indicating that it would be inadvisable to permit the temperature to rise that high, particularly with voltage across the plate. In reactor application as a microphone transducer, it is tentatively recommended that the lithium niobate be maintained in the middle of this temperature range for a major portion of reactor operating time

  9. A PRA case study of extended long term decay heat removal for shutdown risk assessment

    International Nuclear Information System (INIS)

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

    1992-01-01

    A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL). The results of this PRA have shown that the decay heat removal system for EBR-II is extremely robust and reliable. In addition, the methodology used demonstrates how the actions of other systems not normally used for actions of other systems not normally used for decay heat removal can be used to expand the mission time of the decay heat removal system and further increase its reliability. The methodology may also be extended to account for the impact of non-safety systems in enhancing the reliability of other dedicated safety systems

  10. Development of a risk-based in-service inspection program for an LMR

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.

    1996-01-01

    The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. Experimental Breeder Reactor II (EBR-II) is a liquid-metal-cooled fast reactor (LMR) that operated for 30 yr before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the LMR design. For example, it demonstrated passive response to plant upset events, low-pressure primary coolant, and compatibility of the coolant and reactor materials. This ISI program was based on work currently being done by an American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection

  11. NSLS-II filling pattern measurement

    International Nuclear Information System (INIS)

    Yong Hu; Dalesio, L.B.; Kiman Ha; Pinayev, I.

    2012-01-01

    Multi-bunch injection will be deployed at NSLS-II. High bandwidth diagnostic beam monitors with high speed digitizers are used to measure bunch-by-bunch charge variation. In order to minimize intensity-correlated orbit oscillations due to uneven bunch patterns, we need to measure the filling pattern (also named bunch pattern or bunch structure). This paper focuses on filling pattern measurements: how to measure bunch structure and make this information available in EPICS-based control system. This measurement requires combination of 3 types of beam monitors (Wall Current Monitor, Fast Current Transformer and Beam Position Monitor), data acquisition and controls (fast digitizer, EPICS software, etc.) and Event Timing system. High-bandwidth filling pattern monitor requires high-speed digitizer to sample its analog output signal. The evaluation results of commercial fast digitizer Agilent Acqiris and high bandwidth detector Bergoz FCT are presented. We have also tested the algorithm software for filling pattern measurement as well as the interface to event timing system. It appears that filling pattern measurement system is well understood and the tests for control hardware and software have given good results

  12. Postirradiation examination of beryllium pebbles

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1998-01-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

  13. Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Katsunori; Kohyama, Akira; Tanaka, Satoru; Namba, C.; Terai, T.; Kunugi, T.; Muroga, Takeo; Hasegawa, Akira; Sagara, A.; Berk, S.; Zinkle, Steven J.; Sze, Dai Kai; Petti, D. A.; Abdou, Mohamed A.; Morley, Neil B.; Kurtz, Richard J.; Snead, Lance L.; Ghoniem, Nasr M.

    2008-12-01

    This report describes an outline of the activities of the JUPITER-II collaboration (japan-USA program of Irradiation/Integration test for Fusion Research-II), Which has bee carried out through six years (2001-2006) under Phase 4 of the collabroation implemented by Amendment 4 of Annex 1 to the DOE (United States Department of Energy)-MEXT (Ministry of Education ,Culture,Sports,Science and Technology) Cooperation. This program followed the RTNS-II Program (Phase1:1982-4986), the FFTF/MOTA Program (Phase2:1987-1994) and the JUPITER Program (Phase 3: 1995-2000) [1].

  14. Direct liquid content measurement applicable for He II space cryostats

    International Nuclear Information System (INIS)

    Wanner, M.

    1988-01-01

    A direct calorimetric method for content measurement in the He II cryostat ISO was assessed. A well defined heat pulse into the He II bath causes a small temperature increase which can be measured and directly correlated to the liquid mass through the He II specific heat. To study this method under the potential zero gravity constraints of disconnected liquid volumes a setup was established for investigating heat transfer between separated liquid volumes. The results for different fluid configurations confirm that even for completely disconnected volumes the heat is almost immediately distributed throughout the whole liquid by evaporation and recondensation

  15. Irradiation performance of metallic fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

    1989-01-01

    Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

  16. The IFR status and prospects

    International Nuclear Information System (INIS)

    Till, C.E.

    1992-01-01

    The integral fast reactor (IFR) concept consists of a metal-fueled, liquid-metal-cooled fast reactor with its associated pyrometallurgical recovery and recycle of plutonium and higher actinides. The feasibility and operational stability of a liquid-metal-cooled fast reactor has been amply demonstrated in many countries. A comprehensive understanding of the irradiation effects on metal fuels permits extremely high fuel burnup without failures. Interest in the IFR concept grew rapidly when (in 1986) the benign response of the IFR was demonstrated by deliberately putting Experimental Breeder Reactor II (EBR-II) through unprotected loss-of-flow and loss-of-heat-sink transients; these transients were comfortably controlled by the inherent feedback of the system

  17. Comparison of integral cross section values of several cross section libraries in the SAND-II format

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.

    1976-09-01

    A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed

  18. Hot fuel examination facility element spacer wire-wrap machine

    International Nuclear Information System (INIS)

    Tobias, D.A.; Sherman, E.K.

    1989-01-01

    Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

  19. The EBR-II spent fuel treatment program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.

    1995-01-01

    Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option

  20. BelleII@home: Integrate volunteer computing resources into DIRAC in a secure way

    Science.gov (United States)

    Wu, Wenjing; Hara, Takanori; Miyake, Hideki; Ueda, Ikuo; Kan, Wenxiao; Urquijo, Phillip

    2017-10-01

    The exploitation of volunteer computing resources has become a popular practice in the HEP computing community as the huge amount of potential computing power it provides. In the recent HEP experiments, the grid middleware has been used to organize the services and the resources, however it relies heavily on the X.509 authentication, which is contradictory to the untrusted feature of volunteer computing resources, therefore one big challenge to utilize the volunteer computing resources is how to integrate them into the grid middleware in a secure way. The DIRAC interware which is commonly used as the major component of the grid computing infrastructure for several HEP experiments proposes an even bigger challenge to this paradox as its pilot is more closely coupled with operations requiring the X.509 authentication compared to the implementations of pilot in its peer grid interware. The Belle II experiment is a B-factory experiment at KEK, and it uses DIRAC for its distributed computing. In the project of BelleII@home, in order to integrate the volunteer computing resources into the Belle II distributed computing platform in a secure way, we adopted a new approach which detaches the payload running from the Belle II DIRAC pilot which is a customized pilot pulling and processing jobs from the Belle II distributed computing platform, so that the payload can run on volunteer computers without requiring any X.509 authentication. In this approach we developed a gateway service running on a trusted server which handles all the operations requiring the X.509 authentication. So far, we have developed and deployed the prototype of BelleII@home, and tested its full workflow which proves the feasibility of this approach. This approach can also be applied on HPC systems whose work nodes do not have outbound connectivity to interact with the DIRAC system in general.

  1. Effect of material variables on the irradiation performance of boron carbide

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Hollenberg, G.W.

    1980-01-01

    Boron carbide pellets were fabricated with variations in material parameters. These pellets were irradiated in the Experimental Breeder Reactor-II (EBR-II) to determine the effect of these variations on the performance. Helium release from the material and swelling of the pellets are the primary measures of performance. It was determined that material with a smaller grain size released more helium and swelled less. The pellets with boron-to-carbon ratios greater than 4 to 1 did not perform well. Iron additions improved the performance of the material while density variations had little effect

  2. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  3. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  4. Electron density measurements during ion beam transport on Gamble II

    International Nuclear Information System (INIS)

    Weber, B.V.; Hinshelwood, D.D.; Neri, J.M.; Ottinger, P.F.; Rose, D.V.; Stephanakis, S.J.; Young, F.C.

    1999-01-01

    High-sensitivity laser interferometry was used to measure the electron density created when an intense proton beam (100 kA, 1 MeV, 50 ns) from the Gamble II generator was transported through low-pressure gas as part of a project investigating Self-Pinched Transport (SPT) of intense ion beams. This measurement is non-perturbing and sufficiently quantitative to allow benchmarking of codes (particularly IPROP) used to model beam-gas interaction and ion-beam transport. Very high phase sensitivity is required for this measurement. For example, a 100-kA, 1-MeV, 10-cm-radius proton beam with uniform current density has a line-integrated proton density equal to n b L = 3 x 10 13 cm -2 . An equal electron line-density, n e L = n b L, (expected for transport in vacuum) will be detected as a phase shift of the 1.064 microm laser beam of only 0.05degree, or an optical path change of 1.4 x 10 -4 waves (about the size of a hydrogen atom). The time-history of the line-integrated electron density, measured across a diameter of the transport chamber at 43 cm from the input aperture, starts with the proton arrival time and decays differently depending on the gas pressure. The gas conditions included vacuum (10 -4 Torr air), 30 to 220 mTorr He, and 1 Torr air. The measured densities vary by three orders of magnitude, from 10 13 to 10 16 cm -2 for the range of gas pressures investigated. In vacuum, the measured electron densities indicate only co-moving electrons (n e L approximately n b L). In He, when the gas pressure is sufficient for ionization by beam particles and SPT is observed, n e L increases to about 10 n b L. At even higher pressures where electrons contribute to ionization, even higher electron densities are observed with an ionization fraction of about 2%. The diagnostic technique as used on the SPT experiment will be described and a summary of the results will be given. The measurements are in reasonable agreement with theoretical predictions from the IPROP code

  5. Information Technology, Type II Classroom Integration, and the Limited Infrastructure in Schools

    Science.gov (United States)

    Maddux, Cleborne D.; Johnson D. Lamont

    2006-01-01

    In this second special issue on Type II applications of information technology in education, the focus is on classroom integration. This editorial explores some possible explanations for the fact that information technology in schools has not fulfilled its considerable potential. One reason may be that individualized instruction is not part of the…

  6. Luminosity Measurements at LHCb for Run II

    CERN Multimedia

    Coombs, George

    2018-01-01

    A precise measurement of the luminosity is a necessary component of many physics analyses, especially cross-section measurements. At LHCb two different direct measurement methods are used to determine the luminosity: the “van der Meer scan” (VDM) and the “Beam Gas Imaging” (BGI) methods. A combined result from these two methods gave a precision of less than 2% for Run I and efforts are ongoing to provide a similar result for Run II. Fixed target luminosity is determined with an indirect method based on the single electron scattering cross-section.

  7. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  8. Actinide integral measurements in the CFRMF and integral tests for ENDF/B-V

    International Nuclear Information System (INIS)

    Anderl, R.A.

    1982-01-01

    Integral capture and/or fission rates have been reported earlier for several actinides irradiated in the fast neutron field of the Coupled Fast Reactivity Measurements Facility (CFRMF). These nuclides include 232 Th, 233 U, 235 U, 238 U, 237 Np, 239 Pu, 240 Pu, 242 Pu, 241 Am and 243 Am. This paper forucses on the utilization of these integral data for testing the respective cross sections on ENDF/B-V. Integral cross sections derived from the measured reaction rates are tabulated. Results are presented for cross-section data testing which includes integral testing based on a comparison of calculated and measured integral cross sections and testing based on least-squares-adjustment analyses

  9. Experimental and design experience with passive safety features of liquid metal reactors

    International Nuclear Information System (INIS)

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  10. Information Retrieval from SAGE II and MFRSR Multi-Spectral Extinction Measurements

    Science.gov (United States)

    Lacis, Andrew A.; Hansen, James E. (Technical Monitor)

    2001-01-01

    Direct beam spectral extinction measurements of solar radiation contain important information on atmospheric composition in a form that is essentially free from multiple scattering contributions that otherwise tend to complicate the data analysis and information retrieval. Such direct beam extinction measurements are available from the solar occultation satellite-based measurements made by the Stratospheric and Aerosol Gas Experiment (SAGE II) instrument and by ground-based Multi-Filter Shadowband Radiometers (MFRSRs). The SAGE II data provide cross-sectional slices of the atmosphere twice per orbit at seven wavelengths between 385 and 1020 nm with approximately 1 km vertical resolution, while the MFRSR data provide atmospheric column measurements at six wavelengths between 415 and 940 nm but at one minute time intervals. We apply the same retrieval technique of simultaneous least-squares fit to the observed spectral extinctions to retrieve aerosol optical depth, effective radius and variance, and ozone, nitrogen dioxide, and water vapor amounts from the SAGE II and MFRSR measurements. The retrieval technique utilizes a physical model approach based on laboratory measurements of ozone and nitrogen dioxide extinction, line-by-line and numerical k-distribution calculations for water vapor absorption, and Mie scattering constraints on aerosol spectral extinction properties. The SAGE II measurements have the advantage of being self-calibrating in that deep space provides an effective zero point for the relative spectral extinctions. The MFRSR measurements require periodic clear-day Langley regression calibration events to maintain accurate knowledge of instrument calibration.

  11. Development of a risk-based inservice inspection program for a liquid metal reactor

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.

    1996-01-01

    The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

  12. Essentials of measure theory

    CERN Document Server

    Kubrusly, Carlos S

    2015-01-01

    Classical in its approach, this textbook is thoughtfully designed and composed in two parts. Part I is meant for a one-semester beginning graduate course in measure theory, proposing an “abstract” approach to measure and integration, where the classical concrete cases of Lebesgue measure and Lebesgue integral are presented as an important particular case of general theory. Part II of the text is more advanced and is addressed to a more experienced reader. The material is designed to cover another one-semester graduate course subsequent to a first course, dealing with measure and integration in topological spaces. The final section of each chapter in Part I presents problems that are integral to each chapter, the majority of which consist of auxiliary results, extensions of the theory, examples, and counterexamples. Problems which are highly theoretical have accompanying hints. The last section of each chapter of Part II consists of Additional Propositions containing auxiliary and complementary results. Th...

  13. Grid Based Integration Technologies of Virtual Measurement System

    International Nuclear Information System (INIS)

    Zhang, D P; He, L S; Yang, H

    2006-01-01

    This paper presents a novel integrated architecture of measurement system for the new requirements of measurement collaboration, measurement resource interconnection and transparent access etc in the wide-area and across organization in the context of a grid. The complexity of integration on a grid arises from the scale, dynamism, autonomy, and distribution of the measurement resources. The main argument of this paper is that these complexities should be made transparent to the collaborative measurement, via flexible reconfigurable mechanisms and dynamic virtualization services. The paper is started by discussing the integration-oriented measurement architecture which provides collaborative measurement services to distributed measurement resources and then the measurement mechanisms are discussed which implements the transparent access and collaboration of measurement resources by providing protocols, measurement schedule and global data driven model

  14. IIS--Integrated Interactome System: a web-based platform for the annotation, analysis and visualization of protein-metabolite-gene-drug interactions by integrating a variety of data sources and tools.

    Science.gov (United States)

    Carazzolle, Marcelo Falsarella; de Carvalho, Lucas Miguel; Slepicka, Hugo Henrique; Vidal, Ramon Oliveira; Pereira, Gonçalo Amarante Guimarães; Kobarg, Jörg; Meirelles, Gabriela Vaz

    2014-01-01

    High-throughput screening of physical, genetic and chemical-genetic interactions brings important perspectives in the Systems Biology field, as the analysis of these interactions provides new insights into protein/gene function, cellular metabolic variations and the validation of therapeutic targets and drug design. However, such analysis depends on a pipeline connecting different tools that can automatically integrate data from diverse sources and result in a more comprehensive dataset that can be properly interpreted. We describe here the Integrated Interactome System (IIS), an integrative platform with a web-based interface for the annotation, analysis and visualization of the interaction profiles of proteins/genes, metabolites and drugs of interest. IIS works in four connected modules: (i) Submission module, which receives raw data derived from Sanger sequencing (e.g. two-hybrid system); (ii) Search module, which enables the user to search for the processed reads to be assembled into contigs/singlets, or for lists of proteins/genes, metabolites and drugs of interest, and add them to the project; (iii) Annotation module, which assigns annotations from several databases for the contigs/singlets or lists of proteins/genes, generating tables with automatic annotation that can be manually curated; and (iv) Interactome module, which maps the contigs/singlets or the uploaded lists to entries in our integrated database, building networks that gather novel identified interactions, protein and metabolite expression/concentration levels, subcellular localization and computed topological metrics, GO biological processes and KEGG pathways enrichment. This module generates a XGMML file that can be imported into Cytoscape or be visualized directly on the web. We have developed IIS by the integration of diverse databases following the need of appropriate tools for a systematic analysis of physical, genetic and chemical-genetic interactions. IIS was validated with yeast two

  15. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    subassemblies in EBR-II and selected encapsulated tests irradiated in GETR. Other comparative tests in thermal reactors indicate that fast flux and thermal flux results are similar. An adequate understanding of FCCI requires integration of in reactor tests and out-of-reactor applied and fundamental studies. To this end a Fuel Cladding Chemical Interaction Program has been established involving several ERDA laboratories and contractors. Other papers to be presented at this international meeting will describe: FCCI work being carried on out-of-reactor but simulating reactor irradiation conditions; studies using all available data sources aimed at illuminating the mechanism and developing models for FCCI; in-reactor and out-of-reactor tests using various techniques and materials whose objectives are to prevent serious FCCI or to mitigate its effect on fuel pin behavior; and the application of FCCI data to lifetime estimates and design criteria

  16. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    subassemblies in EBR-II and selected encapsulated tests irradiated in GETR. Other comparative tests in thermal reactors indicate that fast flux and thermal flux results are similar. An adequate understanding of FCCI requires integration of in reactor tests and out-of-reactor applied and fundamental studies. To this end a Fuel Cladding Chemical Interaction Program has been established involving several ERDA laboratories and contractors. Other papers to be presented at this international meeting will describe: FCCI work being carried on out-of-reactor but simulating reactor irradiation conditions; studies using all available data sources aimed at illuminating the mechanism and developing models for FCCI; in-reactor and out-of-reactor tests using various techniques and materials whose objectives are to prevent serious FCCI or to mitigate its effect on fuel pin behavior; and the application of FCCI data to lifetime estimates and design criteria.

  17. Fault tolerant strategies for automated operation of nuclear reactors

    International Nuclear Information System (INIS)

    Berkan, R.C.; Tsoukalas, L.

    1991-01-01

    This paper introduces an automatic control system incorporating a number of verification, validation, and command generation tasks with-in a fault-tolerant architecture. The integrated system utilizes recent methods of artificial intelligence such as neural networks and fuzzy logic control. Furthermore, advanced signal processing and nonlinear control methods are also included in the design. The primary goal is to create an on-line capability to validate signals, analyze plant performance, and verify the consistency of commands before control decisions are finalized. The application of this approach to the automated startup of the Experimental Breeder Reactor-II (EBR-II) is performed using a validated nonlinear model. The simulation results show that the advanced concepts have the potential to improve plant availability andsafety

  18. Direct electrical heating of irradiated metal fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1985-01-01

    The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

  19. Aportacions a la flora de les comarques transibèriques, II

    Directory of Open Access Journals (Sweden)

    Rovira, Ana M.

    1988-01-01

    Full Text Available Not availabe

    [fr] On expose quelques nouveautés taxonomiques et chorologiques pour la Flore des comarques catalanes du Sud de l'Ebre. On décrit deux nouveaux taxa: Brassica repanda (Willd. DC. subsp. dertosensis Molero & Rovira et Helianthemum pilosum (L. Pers. var. pseudoasperum Molero & Rovira. On propose aussi une nouvelle combinaison nomenclaturelle: Reseda lutea L. subsp. vivantii (P. Montserrat A. Rovira. Dans l’aspecte chorologique, Euphorbia lagascae Sprengel est indiquée pour la première fois en Catalogne: la plupart des autres espèces sont des nouveautés pour les comarques catalanes du Sud de l'Ebre. [ca] Presentem un recull de novetats corològiques i taxonòmiques per a les comarques catalanes situades alsud de l'Ebre. Es descriuen dos tàxons nous: Brassica repanda (Willd. DC. subsp. dertosensis Molero & Rovira i Helianthemum pilosum (L. Pers. var. pseudoasperum Molero & Rovira. S'estableix una nova combinació nomenclatural: Reseda lutea L. subsp. vivantii (P. Montserrat A. Rovira. En l'aspecte corològic. Euphorbia lagascae Sprengel representa una novetat per a Catalunya. La majoria de les altres citacions representen novetats pel conjunt d'aquestes comarques catalanes meridio-occidentals.

  20. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    Science.gov (United States)

    Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

    2016-12-01

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

  1. Photodiode measurements in Nucte-II

    International Nuclear Information System (INIS)

    Machida, M.; Aramaki, E.A.; Takahashi, T.; Ohara, M.; Nogi, N.

    1989-01-01

    Direct measurements of light emission from plasma produced by a field reversed theta-pinch NUCTE-II have been performed by using a set of photodiode detectors. The analysis shows that the plasma light emission can be related to the bremsstrahlung radiation and it is used to identify η=2 rotational instability parameters as rotation direction, angular velocity, and radial and axial displacement of the plasma column. A rough estimate for the temporal behaviour of the electro temperature has also been obtained by using the photodiode signal together with He-Ne laser interferometer and flux excluded signals. (author) [pt

  2. Measurements of He II Thermal Counterflow Using PIV Technique

    International Nuclear Information System (INIS)

    Zhang, T.; Van Sciver, S.W.

    2004-01-01

    Our previous experiments on the measurements of He II thermal counterflow using Particle Image Velocimetry (PIV) have shown that there exists a substantial discrepancy between the measured and theoretical values of normal fluid velocity. It was assumed that this is due to the slip velocity between tracer particles and liquid helium. In the present work, tracer particles with a much smaller mean diameter and a more uniform size distribution were selected in order to reduce the effect of slip velocity, and an improved two phase fluidized bed technique was used to introduce the particles into liquid helium. The normal fluid velocity of thermal counterflow was then measured using the PIV technique at various heat fluxes and bath temperatures. The experimental results, however, still show the existence of discrepancy between PIV measured particle velocities and the theoretical normal fluid velocity. A preliminary explanation of these results is given based on an interaction of tracer particles with the superfluid component in the He II

  3. Advanced fast reactor fuels program. Second annual progress report, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Baker, R.D.

    1978-12-01

    Results of steady-state (EBR-II) irradiation testing, off-normal irradiation design and testing, fuel-cladding compatibility, and chemical stability of uranium--plutonium carbide and nitride fuels are presented

  4. Possibility and necessity measures and integral equivalence

    Czech Academy of Sciences Publication Activity Database

    Chen, T.; Mesiar, Radko; Li, J.; Stupňanová, A.

    2017-01-01

    Roč. 86, č. 1 (2017), s. 62-72 ISSN 0888-613X Institutional support: RVO:67985556 Keywords : Integral equivalence * Necessity measure * Possibility measure * Survival function * Universal integral Subject RIV: BA - General Mathematics OBOR OECD: Statistics and probability Impact factor: 2.845, year: 2016 http://library.utia.cas.cz/separaty/2017/E/mesiar-0477092.pdf

  5. Setting up measuring campaigns for integrated wastewater modelling

    DEFF Research Database (Denmark)

    Vanrolleghem, P.A.; Schilling, W.; Rauch, Wolfgang

    1999-01-01

    The steps of calibration/confirmation of models in a suggested Ii-step procedure far analysis, planning and implementation of integrated urban wastewater management systems is focused upon in this paper. Based on ample experience obtained in comprehensive investigations throughout Europe recommen...... problems related to suspended solids, specific contaminants, hygienic hazards and total pollutant loss illustrate the recommendations presented. (C) 1999 IAWQ Published by Elsevier Science Ltd. All rights reserved....

  6. Verification study of the FORE-2M nuclear/thermal-hydraulilc analysis computer code

    International Nuclear Information System (INIS)

    Coffield, R.D.; Tang, Y.S.; Markley, R.A.

    1982-01-01

    The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments. (orig.)

  7. Peak radiated power measurement of the DOE Mark II container tag with integrated ST-676 sensor radio frequency identification device.

    Energy Technology Data Exchange (ETDEWEB)

    Jursich, Mark

    2010-04-01

    The total peak radiated power of the Department of Energy Mark II container tag was measured in the electromagnetic reverberation chamber facility at Sandia National Laboratories. The tag's radio frequency content was also evaluated for possible emissions outside the intentional transmit frequency band. No spurious emissions of any significance were found, and the radiated power conformed to the manufacturer's specifications.

  8. Precise measurement of the top-quark mass in the lepton+jets topology at CDF II.

    Science.gov (United States)

    Aaltonen, T; Abulencia, A; Adelman, J; Affolder, T; Akimoto, T; Albrow, M G; Amerio, S; Amidei, D; Anastassov, A; Anikeev, K; Annovi, A; Antos, J; Aoki, M; Apollinari, G; Arisawa, T; Artikov, A; Ashmanskas, W; Attal, A; Aurisano, A; Azfar, F; Azzi-Bacchetta, P; Azzurri, P; Bacchetta, N; Badgett, W; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Baroiant, S; Bartsch, V; Bauer, G; Beauchemin, P-H; Bedeschi, F; Behari, S; Bellettini, G; Bellinger, J; Belloni, A; Benjamin, D; Beretvas, A; Beringer, J; Berry, T; Bhatti, A; Binkley, M; Bisello, D; Bizjak, I; Blair, R E; Blocker, C; Blumenfeld, B; Bocci, A; Bodek, A; Boisvert, V; Bolla, G; Bolshov, A; Bortoletto, D; Boudreau, J; Boveia, A; Brau, B; Brigliadori, L; Bromberg, C; Brubaker, E; Budagov, J; Budd, H S; Budd, S; Burkett, K; Busetto, G; Bussey, P; Buzatu, A; Byrum, K L; Cabrera, S; Campanelli, M; Campbell, M; Canelli, F; Canepa, A; Carrillo, S; Carlsmith, D; Carosi, R; Carron, S; Casal, B; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chang, S H; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Chlebana, F; Cho, I; Cho, K; Chokheli, D; Chou, J P; Choudalakis, G; Chuang, S H; Chung, K; Chung, W H; Chung, Y S; Cilijak, M; Ciobanu, C I; Ciocci, M A; Clark, A; Clark, D; Coca, M; Compostella, G; Convery, M E; Conway, J; Cooper, B; Copic, K; Cordelli, M; Cortiana, G; Crescioli, F; Cuenca Almenar, C; Cuevas, J; Culbertson, R; Cully, J C; DaRonco, S; Datta, M; D'Auria, S; Davies, T; Dagenhart, D; de Barbaro, P; De Cecco, S; Deisher, A; De Lentdecker, G; De Lorenzo, G; Dell'Orso, M; Delli Paoli, F; Demortier, L; Deng, J; Deninno, M; De Pedis, D; Derwent, P F; Di Giovanni, G P; Dionisi, C; Di Ruzza, B; Dittmann, J R; D'Onofrio, M; Dörr, C; Donati, S; Dong, P; Donini, J; Dorigo, T; Dube, S; Efron, J; Erbacher, R; Errede, D; Errede, S; Eusebi, R; Fang, H C; Farrington, S; Fedorko, I; Fedorko, W T; Feild, R G; Feindt, M; Fernandez, J P; Field, R; Flanagan, G; Forrest, R; Forrester, S; Franklin, M; Freeman, J C; Furic, I; Gallinaro, M; Galyardt, J; Garcia, J E; Garberson, F; Garfinkel, A F; Gay, C; Gerberich, H; Gerdes, D; Giagu, S; Giannetti, P; Gibson, K; Gimmell, J L; Ginsburg, C; Giokaris, N; Giordani, M; Giromini, P; Giunta, M; Giurgiu, G; Glagolev, V; Glenzinski, D; Gold, M; Goldschmidt, N; Goldstein, J; Golossanov, A; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González, O; Gorelov, I; Goshaw, A T; Goulianos, K; Gresele, A; Grinstein, S; Grosso-Pilcher, C; Group, R C; Grundler, U; Guimaraes da Costa, J; Gunay-Unalan, Z; Haber, C; Hahn, K; Hahn, S R; Halkiadakis, E; Hamilton, A; Han, B-Y; Han, J Y; Handler, R; Happacher, F; Hara, K; Hare, D; Hare, M; Harper, S; Harr, R F; Harris, R M; Hartz, M; Hatakeyama, K; Hauser, J; Hays, C; Heck, M; Heijboer, A; Heinemann, B; Heinrich, J; Henderson, C; Herndon, M; Heuser, J; Hidas, D; Hill, C S; Hirschbuehl, D; Hocker, A; Holloway, A; Hou, S; Houlden, M; Hsu, S-C; Huffman, B T; Hughes, R E; Husemann, U; Huston, J; Incandela, J; Introzzi, G; Iori, M; Ivanov, A; Iyutin, B; James, E; Jang, D; Jayatilaka, B; Jeans, D; Jeon, E J; Jindariani, S; Johnson, W; Jones, M; Joo, K K; Jun, S Y; Jung, J E; Junk, T R; Kamon, T; Karchin, P E; Kato, Y; Kemp, Y; Kephart, R; Kerzel, U; Khotilovich, V; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, S B; Kim, S H; Kim, Y K; Kimura, N; Kirsch, L; Klimenko, S; Klute, M; Knuteson, B; Ko, B R; Kondo, K; Kong, D J; Konigsberg, J; Korytov, A; Kotwal, A V; Kraan, A C; Kraus, J; Kreps, M; Kroll, J; Krumnack, N; Kruse, M; Krutelyov, V; Kubo, T; Kuhlmann, S E; Kuhr, T; Kulkarni, N P; Kusakabe, Y; Kwang, S; Laasanen, A T; Lai, S; Lami, S; Lammel, S; Lancaster, M; Lander, R L; Lannon, K; Lath, A; Latino, G; Lazzizzera, I; LeCompte, T; Lee, J; Lee, J; Lee, Y J; Lee, S W; Lefèvre, R; Leonardo, N; Leone, S; Levy, S; Lewis, J D; Lin, C; Lin, C S; Lindgren, M; Lipeles, E; Lister, A; Litvintsev, D O; Liu, T; Lockyer, N S; Loginov, A; Loreti, M; Lu, R-S; Lucchesi, D; Lujan, P; Lukens, P; Lungu, G; Lyons, L; Lys, J; Lysak, R; Lytken, E; Mack, P; MacQueen, D; Madrak, R; Maeshima, K; Makhoul, K; Maki, T; Maksimovic, P; Malde, S; Malik, S; Manca, G; Manousakis, A; Margaroli, F; Marginean, R; Marino, C; Marino, C P; Martin, A; Martin, M; Martin, V; Martínez, M; Martínez-Ballarín, R; Maruyama, T; Mastrandrea, P; Masubuchi, T; Matsunaga, H; Mattson, M E; Mazini, R; Mazzanti, P; McFarland, K S; McIntyre, P; McNulty, R; Mehta, A; Mehtala, P; Menzemer, S; Menzione, A; Merkel, P; Mesropian, C; Messina, A; Miao, T; Miladinovic, N; Miles, J; Miller, R; Mills, C; Milnik, M; Mitra, A; Mitselmakher, G; Miyamoto, A; Moed, S; Moggi, N; Mohr, B; Moon, C S; Moore, R; Morello, M; Movilla Fernandez, P; Mülmenstädt, J; Mukherjee, A; Muller, Th; Mumford, R; Murat, P; Mussini, M; Nachtman, J; Nagano, A; Naganoma, J; Nakamura, K; Nakano, I; Napier, A; Necula, V; Neu, C; Neubauer, M S; Nielsen, J; Nodulman, L; Norniella, O; Nurse, E; Oh, S H; Oh, Y D; Oksuzian, I; Okusawa, T; Oldeman, R; Orava, R; Osterberg, K; Pagliarone, C; Palencia, E; Papadimitriou, V; Papaikonomou, A; Paramonov, A A; Parks, B; Pashapour, S; Patrick, J; Pauletta, G; Paulini, M; Paus, C; Pellett, D E; Penzo, A; Phillips, T J; Piacentino, G; Piedra, J; Pinera, L; Pitts, K; Plager, C; Pondrom, L; Portell, X; Poukhov, O; Pounder, N; Prakoshyn, F; Pronko, A; Proudfoot, J; Ptohos, F; Punzi, G; Pursley, J; Rademacker, J; Rahaman, A; Ramakrishnan, V; Ranjan, N; Redondo, I; Reisert, B; Rekovic, V; Renton, P; Rescigno, M; Richter, S; Rimondi, F; Ristori, L; Robson, A; Rodrigo, T; Rogers, E; Rolli, S; Roser, R; Rossi, M; Rossin, R; Roy, P; Ruiz, A; Russ, J; Rusu, V; Saarikko, H; Safonov, A; Sakumoto, W K; Salamanna, G; Saltó, O; Santi, L; Sarkar, S; Sartori, L; Sato, K; Savard, P; Savoy-Navarro, A; Scheidle, T; Schlabach, P; Schmidt, E E; Schmidt, M P; Schmitt, M; Schwarz, T; Scodellaro, L; Scott, A L; Scribano, A; Scuri, F; Sedov, A; Seidel, S; Seiya, Y; Semenov, A; Sexton-Kennedy, L; Sfyrla, A; Shalhout, S Z; Shapiro, M D; Shears, T; Shepard, P F; Sherman, D; Shimojima, M; Shochet, M; Shon, Y; Shreyber, I; Sidoti, A; Sinervo, P; Sisakyan, A; Slaughter, A J; Slaunwhite, J; Sliwa, K; Smith, J R; Snider, F D; Snihur, R; Soderberg, M; Soha, A; Somalwar, S; Sorin, V; Spalding, J; Spinella, F; Spreitzer, T; Squillacioti, P; Stanitzki, M; Staveris-Polykalas, A; St Denis, R; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Stuart, D; Suh, J S; Sukhanov, A; Sun, H; Suslov, I; Suzuki, T; Taffard, A; Takashima, R; Takeuchi, Y; Tanaka, R; Tecchio, M; Teng, P K; Terashi, K; Thom, J; Thompson, A S; Thomson, E; Tipton, P; Tiwari, V; Tkaczyk, S; Toback, D; Tokar, S; Tollefson, K; Tomura, T; Tonelli, D; Torre, S; Torretta, D; Tourneur, S; Trischuk, W; Tsuno, S; Tu, Y; Turini, N; Ukegawa, F; Uozumi, S; Vallecorsa, S; van Remortel, N; Varganov, A; Vataga, E; Vazquez, F; Velev, G; Vellidis, C; Veramendi, G; Veszpremi, V; Vidal, M; Vidal, R; Vila, I; Vilar, R; Vine, T; Vogel, M; Vollrath, I; Volobouev, I; Volpi, G; Würthwein, F; Wagner, P; Wagner, R G; Wagner, R L; Wagner, J; Wagner, W; Wallny, R; Wang, S M; Warburton, A; Waters, D; Weinberger, M; Wester, W C; Whitehouse, B; Whiteson, D; Wicklund, A B; Wicklund, E; Williams, G; Williams, H H; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, C; Wright, T; Wu, X; Wynne, S M; Yagil, A; Yamamoto, K; Yamaoka, J; Yamashita, T; Yang, C; Yang, U K; Yang, Y C; Yao, W M; Yeh, G P; Yoh, J; Yorita, K; Yoshida, T; Yu, G B; Yu, I; Yu, S S; Yun, J C; Zanello, L; Zanetti, A; Zaw, I; Zhang, X; Zhou, J; Zucchelli, S

    2007-11-02

    We present a measurement of the mass of the top quark from proton-antiproton collisions recorded at the CDF experiment in Run II of the Fermilab Tevatron. We analyze events from the single lepton plus jets final state (tt-->W(+)bW(-)b-->lnubqq'b). The top-quark mass is extracted using a direct calculation of the probability density that each event corresponds to the tt final state. The probability is a function of both the mass of the top quark and the energy scale of the calorimeter jets, which is constrained in situ by the hadronic W boson mass. Using 167 events observed in 955 pb(-1) of integrated luminosity, we achieve the single most precise measurement of the top-quark mass, 170.8+/-2.2(stat.)+/-1.4(syst.) GeV/c(2).

  9. Modular version of SIMCON, FPGA based, DSP integrated, LLRF control system for TESLA FEL part II: measurement of SIMCON 3.0 DSP daughterboard

    Science.gov (United States)

    Giergusiewicz, Wojciech; Koprek, Waldemar; Jalmuzna, Wojciech; Pozniak, Krzysztof T.; Romaniuk, Ryszard S.

    2006-02-01

    The paper describes design, construction and initial measurements of an eight channel electronic LLRF device predicted for building of the control system for the W-FEL accelerator at DESY (Hamburg). The device, referred in the paper to as the SIMCON 3.0 (from the SC cavity simulator and controller) consists of a 16 layer, VME size, PCB, a large FPGA chip (VirtexII-4000 by Xilinx), eight fast ADCs and four DACs (by Analog Devices). To our knowledge, the proposed device is the first of this kind for the accelerator technology in which there was achieved (the FPGA based) DSP latency below 200 ns. With the optimized data transmission system, the overall LLRF system latency can be as low as 500 ns. The SIMCON 3.0 sub-system was applied for initial tests with the ACCl module of the VUV FEL accelerator (eight channels) and with the CHECHIA test stand (single channel), both at the DESY. The promising results with the SIMCON 3.0. encouraged us to enter the design of SIMCON 3.1. possessing 10 measurement and control channels and some additional features to be reported in the next technical note. SIMCON 3.0. is a modular solution, while SIMCON 3.1. will be an integrated board of the all-in-one type. Two design approaches - modular and all-in-one - after branching off in this version of the Simcon, will be continued.

  10. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  11. Generalized measures and the Feynman path integral

    International Nuclear Information System (INIS)

    Maslov, V.P.; Chebotarev, A.M.

    1976-01-01

    Generalizations are obtained for the earlier results by the authors concerning the inclusion of the Feynmann path integral in the momentum representation into the general integration theory. Feynmann path integrals are considered which do not represent T-products. Generalized Feynmann measure in the configuration representation is introduced

  12. An emerging role of pectic rhamnogalacturonanII for cell wall integrity

    OpenAIRE

    Reboul, Rebecca; Tenhaken, Raimund

    2012-01-01

    The plant cell wall is a complex network of different polysaccharides and glycoproteins, showing high diversity in nature. The essential components, tethering cell wall are under debate, as novel mutants challenge established models. The mutant ugd2,3 with a reduced supply of the important wall precursor UDP-glucuronic acid reveals the critical role of the pectic compound rhamnogalacturonanII for cell wall stability. This polymer seems to be more important for cell wall integrity than the pre...

  13. Exogenous 24-Epibrassinolide Interacts with Light to Regulate Anthocyanin and Proanthocyanidin Biosynthesis in Cabernet Sauvignon (Vitis vinifera L.).

    Science.gov (United States)

    Zhou, Yali; Yuan, Chunlong; Ruan, Shicheng; Zhang, Zhenwen; Meng, Jiangfei; Xi, Zhumei

    2018-01-09

    Anthocyanins and proanthocyanidins (PAs) are crucial factors that affect the quality of grapes and the making of wine, which were stimulated by various stimuli and environment factors (sugar, hormones, light, and temperature). The aim of the study was to investigate the influence of exogenous 24-Epibrassinolide (EBR) and light on the mechanism of anthocyanins and PAs accumulation in grape berries. Grape clusters were sprayed with EBR (0.4 mg/L) under light and darkness conditions (EBR + L, EBR + D), or sprayed with deionized water under light and darkness conditions as controls (L, D), at the onset of veraison. A large amount of anthocyanins accumulated in the grape skins and was measured under EBR + L and L treatments, whereas EBR + D and D treatments severely suppressed anthocyanin accumulation. This indicated that EBR treatment could produce overlay effects under light, in comparison to that in dark. Real-time quantitative PCR analysis indicated that EBR application up-regulated the expression of genes ( VvCHI1 , VvCHS2 , VvCHS3 , VvDFR , VvLDOX , VvMYBA1 ) under light conditions. Under darkness conditions, only early biosynthetic genes of anthocyanin biosynthesis responded to EBR. Furthermore, we also analyzed the expression levels of the BR-regulated transcription factor VvBZR1 (Brassinazole-resistant 1) and light-regulated transcription factor VvHY5 (Elongated hypocotyl 5). Our results suggested that EBR and light had synergistic effects on the expression of genes in the anthocyanin biosynthesis pathway.

  14. Exogenous 24-Epibrassinolide Interacts with Light to Regulate Anthocyanin and Proanthocyanidin Biosynthesis in Cabernet Sauvignon (Vitis vinifera L.

    Directory of Open Access Journals (Sweden)

    Yali Zhou

    2018-01-01

    Full Text Available Anthocyanins and proanthocyanidins (PAs are crucial factors that affect the quality of grapes and the making of wine, which were stimulated by various stimuli and environment factors (sugar, hormones, light, and temperature. The aim of the study was to investigate the influence of exogenous 24-Epibrassinolide (EBR and light on the mechanism of anthocyanins and PAs accumulation in grape berries. Grape clusters were sprayed with EBR (0.4 mg/L under light and darkness conditions (EBR + L, EBR + D, or sprayed with deionized water under light and darkness conditions as controls (L, D, at the onset of veraison. A large amount of anthocyanins accumulated in the grape skins and was measured under EBR + L and L treatments, whereas EBR + D and D treatments severely suppressed anthocyanin accumulation. This indicated that EBR treatment could produce overlay effects under light, in comparison to that in dark. Real-time quantitative PCR analysis indicated that EBR application up-regulated the expression of genes (VvCHI1, VvCHS2, VvCHS3, VvDFR, VvLDOX, VvMYBA1 under light conditions. Under darkness conditions, only early biosynthetic genes of anthocyanin biosynthesis responded to EBR. Furthermore, we also analyzed the expression levels of the BR-regulated transcription factor VvBZR1 (Brassinazole-resistant 1 and light-regulated transcription factor VvHY5 (Elongated hypocotyl 5. Our results suggested that EBR and light had synergistic effects on the expression of genes in the anthocyanin biosynthesis pathway.

  15. An overview af SAGE I and II ozone measurements

    Science.gov (United States)

    Mccormick, M. P.; Zawodny, J. M.; Veiga, R. E.; Larsen, J. C.; Wang, P. H.

    1989-01-01

    The stratospheric Aerosol and Gas Experiments (SAGE) I and II measure Mie, Rayleigh, and gaseous extinction profiles using the solar occultation technique. These global measurements yield ozone profiles with a vertical resolution of 1 km which have been routinely obtained for the periods from February 1979 to November 1981 (SAGE I) and October 1984 to the present (SAGE II). The long-term periodic behavior of the measured ozone is presented as well as case studies of the observed short-term spatial and temporal variability. A linear regression shows annual, semiannual, and quasi-biennial oscillation features at various altitudes and latitudes which, in general, agree with past work. Also, ozone, aerosol, and water vapor data are described for the Antarctic springtime, showing large variation relative to the vortex. Cross-sections in latitude and altitude and polar plots at various altitudes clearly delineate the ozone hole vertically and areally.

  16. $W$ boson polarization measurement in the $t\\bar{t}$ dilepton channel using the CDF II Detector

    Energy Technology Data Exchange (ETDEWEB)

    Aaltonen, T.; /Helsinki Inst. of Phys.; Alvarez Gonzalez, B.; /Oviedo U. /Cantabria Inst. of Phys.; Amerio, S.; /INFN, Padua; Amidei, D.; /Michigan U.; Anastassov, A.; /Northwestern U. /Fermilab; Annovi, A.; /Frascati; Antos, J.; /Comenius U.; Apollinari, G.; /Fermilab; Appel, J.A.; /Fermilab; Arisawa, T.; /Waseda U.; Artikov, A.; /Dubna, JINR /Texas A-M

    2012-05-01

    We present a measurement of W boson polarization in top-quark decays in t{bar t} events with decays to dilepton final states using 5.1 fb{sup -1} integrated luminosity in p{bar p} collisions collected by the CDF II detector at the Tevatron. A simultaneous measurement of the fractions of longitudinal (f{sub 0}) and right-handed (f{sub +}) W bosons yields the results f{sub 0} = 0.71{sub -0.17}{sup +0.18}(stat) {+-} 0.06(syst) and f{sub +} = -0.07 {+-} 0.09(stat) {+-} 0.03(syst). Combining this measurement with our previous result based on single lepton final states, we obtain f{sub 0} = 0.84 {+-} 0.09(stat) {+-} 0.05(syst) and f{sub +} = -0.16 {+-} 0.05(stat) {+-} 0.04(syst). The results are consistent with standard model expectation.

  17. Speciation of Co(II) and Ni(II) in anaerobic bioreactors measured by competitive ligand exchange - adsorptive stripping voltammetry

    NARCIS (Netherlands)

    Jansen, S.; Steffen, F.; Threels, W.F.; Leeuwen, van H.P.

    2005-01-01

    Competitive ligand exchange-adsorptive stripping voltammetry is applied to speciation analysis of dissolved Ni(II) and Co(II) in an anaerobic bioreactor and similar batch media. Co and Ni speciation in these media can be measured down to concentration levels of ca. 1 nM. Sulfide interference is

  18. Instrument for Airborne Measurement of Carbonyl Sulfide, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase II SBIR program, Southwest Sciences will continue the development of small, low power instrumentation for real-time direct measurement of carbonyl...

  19. Safety aspects of the IFR pyroprocess fuel cycle

    International Nuclear Information System (INIS)

    Forrester, R.J.; Lineberry, M.J.; Charak, I.; Tessier, J.H.; Solbrig, C.W.; Gabor, J.D.

    1989-01-01

    This paper addresses the important safety considerations related to the unique Integral Fast Reactor (IFR) fuel cycle technology, the pyroprocess. Argonne has been developing the IFR since 1984. It is a liquid metal cooled reactor, with a unique metal alloy fuel, and it utilizes a radically new fuel cycle. An existing facility, the Hot Fuel Examination Facility-South (HFEF/S) is being modified and equipped to provide a complete demonstration of the fuel cycle. This paper will concentrate on safety aspects of the future HFEF/S operation, slated to begin late next year. HFEF/S is part of Argonne's complex of reactor test facilities located on the Idaho National Engineering Laboratory. HFEF/S was originally put into operation in 1964 as the EBR-II Fuel Cycle Facility (FCF) (Stevenson, 1987). From 1964--69 FCF operated to demonstrate an earlier and incomplete form of today's pyroprocess, recycling some 400 fuel assemblies back to EBR-II. The FCF mission was then changed to one of an irradiated fuels and materials examination facility, hence the name change to HFEF/S. The modifications consist of activities to bring the facility into conformance with today's much more stringent safety standards, and, of course, providing the new process equipment. The pyroprocess and the modifications themselves are described more fully elsewhere (Lineberry, 1987; Chang, 1987). 18 refs., 5 figs., 2 tabs

  20. Paired emitter-detector light emitting diodes for the measurement of lead(II) and cadmium(II)

    International Nuclear Information System (INIS)

    Lau, K.-T.; McHugh, Eimear; Baldwin, Susan; Diamond, Dermot

    2006-01-01

    A transmittance mode optical device based on using a reverse biased light emitting diode (LED) as light detector has been developed for colorimetric analysis. This new optical device was validated with bromocresol green dye for absorbance measurements before being employed for detecting cadmium(II) and lead(II) in water. Results show that the performance of this LED-based device is comparable to much more expensive bench top UV-vis instruments, but with the advantages of being low cost, low power and simple to operate

  1. Paired emitter-detector light emitting diodes for the measurement of lead(II) and cadmium(II)

    Energy Technology Data Exchange (ETDEWEB)

    Lau, K.-T. [Adaptive Sensors Group, National Centre for Sensor Research, School of Chemical Sciences, Dublin City University, Dublin 9 (Ireland)]. E-mail: kim.lau@dcu.ie; McHugh, Eimear [Adaptive Sensors Group, National Centre for Sensor Research, School of Chemical Sciences, Dublin City University, Dublin 9 (Ireland); Baldwin, Susan [Adaptive Sensors Group, National Centre for Sensor Research, School of Chemical Sciences, Dublin City University, Dublin 9 (Ireland); Diamond, Dermot [Adaptive Sensors Group, National Centre for Sensor Research, School of Chemical Sciences, Dublin City University, Dublin 9 (Ireland)]. E-mail: Dermot.diamond@dcu.ie

    2006-05-31

    A transmittance mode optical device based on using a reverse biased light emitting diode (LED) as light detector has been developed for colorimetric analysis. This new optical device was validated with bromocresol green dye for absorbance measurements before being employed for detecting cadmium(II) and lead(II) in water. Results show that the performance of this LED-based device is comparable to much more expensive bench top UV-vis instruments, but with the advantages of being low cost, low power and simple to operate.

  2. Measurement of the t (bar t) cross section at the Run II Tevatron using Support Vector Machines

    International Nuclear Information System (INIS)

    Whitehouse, Benjamin Eric

    2010-01-01

    This dissertation measures the t(bar t) production cross section at the Run II CDF detector using data from early 2001 through March 2007. The Tevatron at Fermilab is a p(bar p) collider with center of mass energy √s = 1.96 TeV. This data composes a sample with a time-integrated luminosity measured at 2.2 ± 0.1 fb -1 . A system of learning machines is developed to recognize t(bar t) events in the 'lepton plus jets' decay channel. Support Vector Machines are described, and their ability to cope with a multi-class discrimination problem is provided. The t(bar t) production cross section is then measured in this framework, and found to be σ t# bar t# = 7.14 ± 0.25 (stat) -0.86 +0.61 (sys) pb.

  3. Measurements of Narrow Mg II Associated Absorption Doublets with ...

    Indian Academy of Sciences (India)

    2016-01-27

    Jan 27, 2016 ... The measurement of the variations of absorption lines over time is a good method to study the physical conditions of absorbers. In this paper, we measure the variations of the line strength of 36 narrow Mg II2796, 2803 associated absorption doublets, which are imprinted on 31 quasar spectra with two ...

  4. Technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check''

    International Nuclear Information System (INIS)

    Serkiz, A.W.

    1988-04-01

    This report contains the technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check.'' An evaluation of the containment isolation history from 1965 to 1983 reveals that (except for a small number of events) containment integrity has been maintained and that the majority of reported events have been events related to exceeding Technical Specification limits (or 0.6 of the allowable leakage level). In addition, more recent risk analyses have shown that allowable leakage rates even if increased by a factor of 10 would not significantly increase risk. Potential methods of continuous monitoring are identified and evaluated. Therefore, these technical findings and risk evaluations support closure of Generic Safety Issue II.E.4.3

  5. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  6. Introduction to Measure Theory and Integration

    CERN Document Server

    Ambrosio, Luigi; Mennucci, Andrea

    2011-01-01

    This textbook collects the notes for an introductory course in measure theory and integration. The course was taught by the authors to undergraduate students of the Scuola Normale Superiore, in the years 2000-2011. The goal of the course was to present, in a quick but rigorous way, the modern point of view on measure theory and integration, putting Lebesgue's Euclidean space theory into a more general context and presenting the basic applications to Fourier series, calculus and real analysis. The text can also pave the way to more advanced courses in probability, stochastic processes or geomet

  7. Time and temperature dependent breakdown characteristics of ZnS:Mn films obtained by rf-magnetron sputtering

    Science.gov (United States)

    Zhigal'Skii, A. A.; Mukhachev, V. A.; Troyan, P. E.

    1994-04-01

    Breakdown delay times (tdel) for films of managanese-doped zinc sulfide (ZnS:Mn) were measured in the range 10-6-10-1 s. The maximum value was tdel=10-3-10-2 s. The electrical strength (Ebr) was found to increase as the voltage pulse duration was reduced, the more so the thinner the ZnS:Mn film. The temperature dependence of Ebr exhibited a weak reduction in Ebr as the temperature was raised to roughly 80°C and a sharp reduction in Ebr for T>130°C. A maximum in Ebr was observed at T≈130°C which is presumably explained by a structural modification of the ZnS:Mn film. The experimental results obtained are explained in terms of a combined electronic and thermal breakdown mechanism.

  8. A gateless charge integrator for Borexino energy measurement

    International Nuclear Information System (INIS)

    Lagomarsino, V.; Testera, G.

    1999-01-01

    A gateless charge integrator designed for the energy measurement in the Borexino experiment is described and the results of various tests performed on prototypes are shown. The circuit integrates always its input taking advantage of the AC coupling to the photomultipliers that equalizes to zero the charge associated with each event signal. A double sampling of the integrator output allows to perform a charge measurement in principle without dead time, avoiding the use of gate signals and baseline restoration networks and permitting a precise detection of the fast correlated events (delayed coincidences due to the decay of a nuclide producing a daughter having a lifetime in the tens of ns time range). The precision of the charge measurement is discussed together with the performances of the front end Borexino board where the integrator is mounted

  9. How type II CRISPR-Cas establish immunity through Cas1-Cas2-mediated spacer integration.

    Science.gov (United States)

    Xiao, Yibei; Ng, Sherwin; Nam, Ki Hyun; Ke, Ailong

    2017-10-05

    CRISPR (clustered regularly interspaced short palindromic repeats) and the nearby Cas (CRISPR-associated) operon establish an RNA-based adaptive immunity system in prokaryotes. Molecular memory is created when a short foreign DNA-derived prespacer is integrated into the CRISPR array as a new spacer. Whereas the RNA-guided CRISPR interference mechanism varies widely among CRISPR-Cas systems, the spacer integration mechanism is essentially identical. The conserved Cas1 and Cas2 proteins form an integrase complex consisting of two distal Cas1 dimers bridged by a Cas2 dimer. The prespacer is bound by Cas1-Cas2 as a dual-forked DNA, and the terminal 3'-OH of each 3' overhang serves as an attacking nucleophile during integration. The prespacer is preferentially integrated into the leader-proximal region of the CRISPR array, guided by the leader sequence and a pair of inverted repeats inside the CRISPR repeat. Spacer integration in the well-studied Escherichia coli type I-E CRISPR system also relies on the bacterial integration host factor. In type II-A CRISPR, however, Cas1-Cas2 alone integrates spacers efficiently in vitro; other Cas proteins (such as Cas9 and Csn2) have accessory roles in the biogenesis phase of prespacers. Here we present four structural snapshots from the type II-A system of Enterococcus faecalis Cas1 and Cas2 during spacer integration. Enterococcus faecalis Cas1-Cas2 selectively binds to a splayed 30-base-pair prespacer bearing 4-nucleotide 3' overhangs. Three molecular events take place upon encountering a target: first, the Cas1-Cas2-prespacer complex searches for half-sites stochastically, then it preferentially interacts with the leader-side CRISPR repeat, and finally, it catalyses a nucleophilic attack that connects one strand of the leader-proximal repeat to the prespacer 3' overhang. Recognition of the spacer half-site requires DNA bending and leads to full integration. We derive a mechanistic framework to explain the stepwise spacer

  10. Integrated Spreadsheets as a Paradigm of Type II Technology Applications in Mathematics Teacher Education

    Science.gov (United States)

    Abramovich, Sergei

    2016-01-01

    The paper presents the use of spreadsheets integrated with digital tools capable of symbolic computations and graphic constructions in a master's level capstone course for secondary mathematics teachers. Such use of spreadsheets is congruent with the Type II technology applications framework aimed at the development of conceptual knowledge in the…

  11. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    International Nuclear Information System (INIS)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B.

    2013-01-01

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs

  12. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.

  13. Measuring cross-border regional STI integration

    Energy Technology Data Exchange (ETDEWEB)

    Makkonen, T

    2016-07-01

    Earlier quantitative studies on cross-border regional integration processes have commonly neglected science, technology and innovation (STI) indicators: even the most notable example of a composite indicator approach to measuring cross-border regional integration, i.e. the Oresund index, lacks a sub-category for STI. Consequently, by ignoring cross-border innovation and knowledge flows, the Oresund integration index fails to take into account one of the most important drivers of economic growth in cross-border regions. Therefore, a new composite STI indicator (sub-category) was introduced to strengthen the Oresund integration index. This was compiled from patent, publication and collaborative R&D project data. The findings show that this index performs reasonably well in depicting STI integration, while at the same time remaining simple and straightforward enough to be adopted in other crossborder regions. (Author)

  14. Core and shielding analysis of the SCM-100

    International Nuclear Information System (INIS)

    Olson, A. P.

    2002-01-01

    It is widely accepted that an intense neutron source can be produced in a suitable target by spallation neutrons generated by a high-current high-energy proton beam. Typical beam energy for such an accelerator is 400 to 2000 MeV. A conventional critical reactor can readily be replaced by a ''sub-critical reactor'' driven by this source. A 5 MW proton beam at 600 MeV can drive a sub-critical reactor to 100 MWt. The accelerator and the associated plant support equipment at these design specifications are complex systems, but they are well within recent technology. The purpose of this study was to examine core design and shielding design issues for a 100 MWt sodium-cooled fast-spectrum Sub-Critical Multiplier (SCM-100) based on LMFBR technology, but driven by an intense neutron source created by spallation reactions. SCM-100 is a component of the Accelerator Driven Test Facility. In this report we provide an overview of the SCM-100 concept. Two designs were investigated: (1) a vertical entry for the beam on the axial centerline; and (2) an inclined entry design where the core is ''C'' shaped and the beam enters the side of the target at an angle of 32 degrees. A brief overview of relevant shielding design data from EBR-II is also provided. The key result of this report is that the inclined entry design cannot achieve design objectives for radial power peaking. Consequently it cannot achieve design objectives for peak neutron flux. Axial power peaking factors are controlled by the axial fuel height and the axial reflector properties. These dimensions and compositions are very similar in SCM-100 to those of EBR-II. EBR-II had an axial power peaking factor of 1.093, and a radial power peaking factor of about 1.46. The radial power peaking of SCM-100 with the inclined entry is too extreme at 2.15, and cannot be made acceptable by modifying the size and detailed shape of the ''C'' shaped core and reflector. The axial power peaking of SCM-100 is very close to that of EBR-II

  15. Reirradiation of mixed-oxide fuel pins at increased temperatures

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, E.T.

    1976-05-01

    Mixed-oxide fuel pins from EBR-II irradiations were reirradiated in the General Electric Test Reactor (GETR) at higher temperatures than experienced in EBR-II to study effects of the increased operating temperatures on thermal/mechanical and chemical behavior. The response of a mixed-oxide fuel pin to a power increase after having operated at a lower power for a significant portion of its life-time is an area of performance evaluation where little information currently exists. Results show that the cladding diameter changes resulting from the reirradiation are strongly dependent upon both prior burnup level and the magnitude of the temperature increase. Results provide the initial rough outlines of boundaries within which mixed-oxide fuel pins can or cannot tolerate power increases after substantial prior burnup at lower powers

  16. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  17. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  18. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  19. Precise measurement of the top quark mass in the lepton+jets topology at CDF II

    Energy Technology Data Exchange (ETDEWEB)

    Abulencia, A.; /Illinois U., Urbana; Adelman, J.; /Chicago U.; Affolder, T.; /UC, Santa Barbara; Akimoto, T.; /Tsukuba U.; Albrow, M.G.; /Fermilab; Amerio, S.; /Padua U.; Amidei, D.; /Michigan U.; Anastassov, A.; /Rutgers U., Piscataway; Anikeev, K.; /Fermilab; Annovi, A.; /Frascati; Antos, J.; /Comenius U. /Tsukuba U.

    2007-03-01

    The authors present a measurement of the mass of the top quark from proton-antiproton collisions recorded at the CDF experiment in Run II of the Fermilab Tevatron. They analyze events from the single lepton plus jets final state (t{bar t} {yields} W{sup +}bW{sup -}{bar b} {yields} lvbq{bar q}{bar b}). The top quark mass is extracted using a direct calculation of the probability density that each event corresponds to the t{bar t} final state. The probability is a function of both the mass of the top quark and the energy scale of the calorimeter jets, which is constrained in situ by the hadronic W boson mass. Using 167 events observed in 955 pb{sup -1} of integrated luminosity, they achieve the single most precise measurement of the top quark mass, 170.8 {+-} 2.2(stat.) {+-} 1.4(syst.) GeV/c{sup 2}.

  20. Implementation of multivariable control techniques with application to Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Berkan, R.C.

    1990-06-01

    After several successful applications to aerospace industry, the modern control theory methods have recently attracted many control engineers from other engineering disciplines. For advanced nuclear reactors, the modern control theory may provide major advantages in safety, availability, and economic aspects. This report is intended to illustrate the feasibility of applying the linear quadratic Gaussian (LQG) compensator in nuclear reactor applications. The LQG design is compared with the existing classical control schemes. Both approaches are tested using the Experimental Breeder Reactor 2 (EBR-2) as the system. The experiments are performed using a mathematical model of the EBR-2 plant. Despite the fact that the controller and plant models do not include all known physical constraints, the results are encouraging. This preliminary study provides an informative, introductory picture for future considerations of using modern control theory methods in nuclear industry. 10 refs., 25 figs

  1. First results from SAGE II

    International Nuclear Information System (INIS)

    Abdurashitov, J.N.; Faizov, E.L.; Gavrin, V.N.

    1994-01-01

    The Russian-American Gallium solar neutrino Experiment (SAGE) began the second phase of operation (SAGE II) in September of 1992. Monthly measurements of the integral flux of solar neutrinos have been made with 55 tonnes of gallium. The K-peak results of the first five runs of SAGE II give a capture rate of 76 -18 +21 (stat) -7 +5 (sys) SNU. combined with the SAGE I result, the capture rate is 74 -12 +13 (stat) -7 +5 (sys) SNU. This represents only 56%--60% of the capture rate predicted by different Standard Solar Models

  2. Status of gamma-ray heating characterization in LMFBR

    International Nuclear Information System (INIS)

    Gold, R.

    1975-11-01

    Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR

  3. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  4. The sodium process facility at Argonne National Laboratory - West

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1997-01-01

    Argonne National Laboratory - West (ANL-W) has approximately 680,000 liters (180,000 gallons) of raw sodium stored in facilities on site. As mandated by the State of Idaho and the United States Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The SPF was designed to react elemental sodium to sodium carbonate through two-stages involving caustic process and carbonate process steps. The sodium is first reacted to sodium hydroxide in the caustic process step. The caustic process step involves the injection of sodium into a nickel reaction vessel filled with a 50 wt% solution of sodium hydroxide. Water is also injected, controlling the boiling point of the solution. In the carbonate process, the sodium hydroxide is reacted with carbon dioxide to form sodium carbonate. This dry powder, similar in consistency to baking soda, is a waste form acceptable for burial in the State of Idaho as a non-hazardous, radioactive waste. The caustic process was originally designed and built in the 1980s for reacting the 290,000 liters (77,000 gallons) of primary sodium from the Fermi-1 Reactor to sodium hydroxide. The hydroxide was slated to be used to neutralize acid products from the PUREX process at the Hanford site. However, changes in the DOE mission precluded the need for hydroxide and the caustic process was never operated. With the shutdown of the Experimental Breeder Reactor-II (EBR-II), the necessity for a facility to react sodium was identified. In order to comply with Resource Conservation and Recovery Act (RCRA) requirements, the sodium had to be converted into a waste form acceptable for disposal in a Sub-Title D low-level radioactive waste disposal facility. Sodium hydroxide is a RCRA

  5. Indicators and Measurement Tools for Health Systems Integration: A Knowledge Synthesis

    Directory of Open Access Journals (Sweden)

    Esther Suter

    2017-11-01

    Full Text Available Background: Despite far reaching support for integrated care, conceptualizing and measuring integrated care remains challenging. This knowledge synthesis aimed to identify indicator domains and tools to measure progress towards integrated care. Methods: We used an established framework and a Delphi survey with integration experts to identify relevant measurement domains. For each domain, we searched and reviewed the literature for relevant tools. Findings: From 7,133 abstracts, we retrieved 114 unique tools. We found many quality tools to measure care coordination, patient engagement and team effectiveness/performance. In contrast, there were few tools in the domains of performance measurement and information systems, alignment of organizational goals and resource allocation. The search yielded 12 tools that measure overall integration or three or more indicator domains. Discussion: Our findings highlight a continued gap in tools to measure foundational components that support integrated care. In the absence of such targeted tools, “overall integration” tools may be useful for a broad assessment of the overall state of a system. Conclusions: Continued progress towards integrated care depends on our ability to evaluate the success of strategies across different levels and context. This study has identified 114 tools that measure integrated care across 16 domains, supporting efforts towards a unified measurement framework.

  6. Indicators and Measurement Tools for Health Systems Integration: A Knowledge Synthesis

    Science.gov (United States)

    Oelke, Nelly D.; da Silva Lima, Maria Alice Dias; Stiphout, Michelle; Janke, Robert; Witt, Regina Rigatto; Van Vliet-Brown, Cheryl; Schill, Kaela; Rostami, Mahnoush; Hepp, Shelanne; Birney, Arden; Al-Roubaiai, Fatima; Marques, Giselda Quintana

    2017-01-01

    Background: Despite far reaching support for integrated care, conceptualizing and measuring integrated care remains challenging. This knowledge synthesis aimed to identify indicator domains and tools to measure progress towards integrated care. Methods: We used an established framework and a Delphi survey with integration experts to identify relevant measurement domains. For each domain, we searched and reviewed the literature for relevant tools. Findings: From 7,133 abstracts, we retrieved 114 unique tools. We found many quality tools to measure care coordination, patient engagement and team effectiveness/performance. In contrast, there were few tools in the domains of performance measurement and information systems, alignment of organizational goals and resource allocation. The search yielded 12 tools that measure overall integration or three or more indicator domains. Discussion: Our findings highlight a continued gap in tools to measure foundational components that support integrated care. In the absence of such targeted tools, “overall integration” tools may be useful for a broad assessment of the overall state of a system. Conclusions: Continued progress towards integrated care depends on our ability to evaluate the success of strategies across different levels and context. This study has identified 114 tools that measure integrated care across 16 domains, supporting efforts towards a unified measurement framework. PMID:29588637

  7. Doppler coherence imaging of ion dynamics in VINETA.II and ASDEX-upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Gradic, Dorothea; Ford, Oliver; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald (Germany); Lunt, Tilmann [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2016-07-01

    In magnetically confining plasma experiments, diagnosis of ion flows is of great importance to measure the plasma response to the magnetic field or the exhaust particle flows in the divertor areas. Doppler coherence imaging spectroscopy (CIS) is a relatively new technique for the observation of plasma bulk ion dynamics. It is a passive optical diagnostic enabling line-integrated measurements to obtain 2D images of the ion flow and ion temperature. The general principle is similar to traditional Doppler spectroscopy, however CIS uses an imaging interferometer to perform narrow-bandwidth Fourier spectroscopy. A major advantage of the coherence imaging technique is the large amount of spatial information recovered. This allows tomographic inversion of the line-integrated measurements. With existing CIS setups, scrape-off-layer and high field side edge impurity flows could be observed in the MAST, core and edge poloidal He II flows in the WEGA stellarator and divertor impurity flows in DIII-D. The main objective of this study is the research of ion dynamics in the small linear plasma experiment VINETA.II and ASDEX-Upgrade. First Doppler CIS measurements from Ar-II plasma discharges in VINETA.II and He-II, C-III divertor flows in ASDEX-Upgrade and their preliminary interpretation will be presented.

  8. Measurements of the exclusive production of a real photon with the ZEUS detector at HERA II

    International Nuclear Information System (INIS)

    Santamarta Martinez, R.

    2007-06-01

    The study of exclusive processes is one of the most promising tools to obtain information on the nucleon via generalized parton distributions. The simplest of these reactions is the Deeply Virtual Compton Scattering process in which a real photon is produced via diffractive exchange. Using the interference term between Deeply Virtual Compton Scattering and the Bethe-Heitler processes it is possible to extract the generalized parton distributions. The measurement presented in this thesis is a study of Deeply Virtual Compton Scattering and Bethe-Heitler processes recorder with the ZEUS detector. The data analyzed were taken with HERA II in the years 2003 to 2005, corresponding to an integrated luminosity of 41 pb -1 for e + p and 136 pb -1 for e - p scattering data. The cross section of the elastic Bethe-Heitler process measured in the kinematic region 230 2 2 and vertical stroke t vertical stroke 2 is presented. The feasibility of measurement of the Deeply Virtual Compton Scattering process is also reported. (orig.)

  9. Towards an interpretive measurement framework to assess the levels of integrated and integrative thinking within organisations

    Directory of Open Access Journals (Sweden)

    Daniel Schörger

    2015-07-01

    Full Text Available This research study is located within the context of corporate reporting and is relevant for the agenda of sustainability and sustainable development. The specific context for this study is the South African mining industry, within which three units in the form of three companies, were chosen to provide a coherent case for this study. The sample for the analysis is based on the integrated reports of these companies for the years 2012 and 2013. This gives this research a total sample size of six reports. Based on the research findings an initial interpretive measurement framework to assess the levels of capital integration has been theorised which enables the various stakeholders of an organisation to assess the integrated and integrative thinking capabilities. The level of integration is represented as a maturity scale on which integrated thinking is associated with the lower levels, while integrative thinking is attributed to higher levels of maturity. In the elaborated framework, integrated thinking is perceived as being a prerequisite for integrative thinking. The practical implication of this study is that it provides a potential measurement framework for various organisational stakeholders, including investors, to assess the thinking capabilities that are more likely to lead to long term financial stability and sustainability. The value of this research study is that it provides an initial step towards measuring the level of integrated and integrative thinking capabilities within organisations where no such measurement framework currently exists. The limitations and implications of this research study are that the interpretive measurement framework represents merely an initial step and an ongoing working hypothesis which requires further research to develop its maturity and usefulness.

  10. Confirmatory Factor Analysis of the Beck Depression Inventory-II in Bariatric Surgery Candidates

    Science.gov (United States)

    Hall, Brian J.; Hood, Megan M.; Nackers, Lisa M.; Azarbad, Leila; Ivan, Iulia; Corsica, Joyce

    2013-01-01

    Screening for depression is an integral part of psychological evaluations conducted prior to bariatric surgery. The Beck Depression Inventory-II (BDI-II) is the most commonly used measure of depression in these treatment evaluations. The reliability and validity of the BDI-II has not yet been evaluated within bariatric surgery-seeking samples,…

  11. Commercial disposal of High Integrity Containers (HICs) containing EPICOR-II prefilters from Three Mile Island

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Lynch, R.J.; Tyacke, M.J.

    1985-09-01

    This report describes the processes of loading, transporting, and commercially disposing of 45 High Integrity Containers (HICs), each containing an EPICOR-II prefilter. Also described are the improvements that were applied in the disposition of the 45 commercial EPICOR-II prefilters at the Idaho National Engineering Laboratory (INEL), versus those used for the demonstration unit. The significance of this effort was that the commercial disposal campaign involved the first-of-a-kind production use of a reinforced concrete HIC at the US Ecology, Inc. facility in the State of Washington. This allowed for safe disposal of high-specific-activity ion exchange material in EPICOR-II prefilters generated during the cleanup of the Unit-2 Auxiliary and Fuel Handling Building of the Three Mile Island Nuclear Power Station. 26 figs

  12. Measurement of the t$\\bar{t}$ cross section at the Run II Tevatron using Support Vector Machines

    Energy Technology Data Exchange (ETDEWEB)

    Whitehouse, Benjamin Eric [Tufts Univ., Medford, MA (United States)

    2010-08-01

    This dissertation measures the t$\\bar{t}$ production cross section at the Run II CDF detector using data from early 2001 through March 2007. The Tevatron at Fermilab is a p$\\bar{p}$ collider with center of mass energy √s = 1.96 TeV. This data composes a sample with a time-integrated luminosity measured at 2.2 ± 0.1 fb-1. A system of learning machines is developed to recognize t$\\bar{t}$ events in the 'lepton plus jets' decay channel. Support Vector Machines are described, and their ability to cope with a multi-class discrimination problem is provided. The t$\\bar{t}$ production cross section is then measured in this framework, and found to be σt$\\bar{t}$ = 7.14 ± 0.25 (stat)-0.86+0.61(sys) pb.

  13. Real analysis measure theory, integration, and Hilbert spaces

    CERN Document Server

    Stein, Elias M

    2005-01-01

    Real Analysis is the third volume in the Princeton Lectures in Analysis, a series of four textbooks that aim to present, in an integrated manner, the core areas of analysis. Here the focus is on the development of measure and integration theory, differentiation and integration, Hilbert spaces, and Hausdorff measure and fractals. This book reflects the objective of the series as a whole: to make plain the organic unity that exists between the various parts of the subject, and to illustrate the wide applicability of ideas of analysis to other fields of mathematics and science. After

  14. Financial integration in the European Union. Measurement and determination

    NARCIS (Netherlands)

    Lemmen, J.J.G.

    1996-01-01

    The first part of this study addresses the measurement of financial integration in the European Union (EU). First, we present empirical evidence on the degree of financial integration as measured with interest parity conditions. Second, the study applies an error-correction model of

  15. Design and Measurement of the NSLS II Quadrupole Prototypes

    Energy Technology Data Exchange (ETDEWEB)

    Rehak,M.; Jain, A. K.; Skaritka, J.; Spataro, C.

    2009-05-04

    The design and measurement of the NSLS-II ring quadrupoles prototypes are presented. These magnets are part of a larger prototype program described in [1]. Advances in software, hardware, and manufacturing have led to some new level of insight in the quest for the perfect magnet design. Three geometric features are used to minimize the first three allowed harmonics by way of optimization. Validations through measurement and confidence levels in calculations are established.

  16. First measurement of the W-boson mass in run II of the Tevatron.

    Science.gov (United States)

    Aaltonen, T; Abulencia, A; Adelman, J; Affolder, T; Akimoto, T; Albrow, M G; Amerio, S; Amidei, D; Anastassov, A; Anikeev, K; Annovi, A; Antos, J; Aoki, M; Apollinari, G; Arisawa, T; Artikov, A; Ashmanskas, W; Attal, A; Aurisano, A; Azfar, F; Azzi-Bacchetta, P; Azzurri, P; Bacchetta, N; Badgett, W; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Baroiant, S; Bartsch, V; Bauer, G; Beauchemin, P-H; Bedeschi, F; Behari, S; Bellettini, G; Bellinger, J; Belloni, A; Benjamin, D; Beretvas, A; Beringer, J; Berry, T; Bhatti, A; Binkley, M; Bisello, D; Bizjak, I; Blair, R E; Blocker, C; Blumenfeld, B; Bocci, A; Bodek, A; Boisvert, V; Bolla, G; Bolshov, A; Bortoletto, D; Boudreau, J; Boveia, A; Brau, B; Brigliadori, L; Bromberg, C; Brubaker, E; Budagov, J; Budd, H S; Budd, S; Burkett, K; Busetto, G; Bussey, P; Buzatu, A; Byrum, K L; Cabrera, S; Campanelli, M; Campbell, M; Canelli, F; Canepa, A; Carrillo, S; Carlsmith, D; Carosi, R; Carron, S; Casal, B; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chang, S H; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Chlebana, F; Cho, I; Cho, K; Chokheli, D; Chou, J P; Choudalakis, G; Chuang, S H; Chung, K; Chung, W H; Chung, Y S; Cilijak, M; Ciobanu, C I; Ciocci, M A; Clark, A; Clark, D; Coca, M; Compostella, G; Convery, M E; Conway, J; Cooper, B; Copic, K; Cordelli, M; Cortiana, G; Crescioli, F; Almenar, C Cuenca; Cuevas, J; Culbertson, R; Cully, J C; Daronco, S; Datta, M; D'Auria, S; Davies, T; Dagenhart, D; de Barbaro, P; De Cecco, S; Deisher, A; De Lentdecker, G; De Lorenzo, G; Dell'orso, M; Delli Paoli, F; Demortier, L; Deng, J; Deninno, M; De Pedis, D; Derwent, P F; Di Giovanni, G P; Dionisi, C; Di Ruzza, B; Dittmann, J R; D'Onofrio, M; Dörr, C; Donati, S; Dong, P; Donini, J; Dorigo, T; Dube, S; Efron, J; Erbacher, R; Errede, D; Errede, S; Eusebi, R; Fang, H C; Farrington, S; Fedorko, I; Fedorko, W T; Feild, R G; Feindt, M; Fernandez, J P; Field, R; Flanagan, G; Forrest, R; Forrester, S; Franklin, M; Freeman, J C; Furic, I; Gallinaro, M; Galyardt, J; Garcia, J E; Garberson, F; Garfinkel, A F; Gay, C; Gerberich, H; Gerdes, D; Giagu, S; Giannetti, P; Gibson, K; Gimmell, J L; Ginsburg, C; Giokaris, N; Giordani, M; Giromini, P; Giunta, M; Giurgiu, G; Glagolev, V; Glenzinski, D; Gold, M; Goldschmidt, N; Goldstein, J; Golossanov, A; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González, O; Gorelov, I; Goshaw, A T; Goulianos, K; Gresele, A; Grinstein, S; Grosso-Pilcher, C; Grundler, U; da Costa, J Guimaraes; Gunay-Unalan, Z; Haber, C; Hahn, K; Hahn, S R; Halkiadakis, E; Hamilton, A; Han, B-Y; Han, J Y; Handler, R; Happacher, F; Hara, K; Hare, D; Hare, M; Harper, S; Harr, R F; Harris, R M; Hartz, M; Hatakeyama, K; Hauser, J; Hays, C; Heck, M; Heijboer, A; Heinemann, B; Heinrich, J; Henderson, C; Herndon, M; Heuser, J; Hidas, D; Hill, C S; Hirschbuehl, D; Hocker, A; Holloway, A; Hou, S; Houlden, M; Hsu, S-C; Huffman, B T; Hughes, R E; Husemann, U; Huston, J; Incandela, J; Introzzi, G; Iori, M; Ivanov, A; Iyutin, B; James, E; Jang, D; Jayatilaka, B; Jeans, D; Jeon, E J; Jindariani, S; Johnson, W; Jones, M; Joo, K K; Jun, S Y; Jung, J E; Junk, T R; Kamon, T; Karchin, P E; Kato, Y; Kemp, Y; Kephart, R; Kerzel, U; Khotilovich, V; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, S B; Kim, S H; Kim, Y K; Kimura, N; Kirsch, L; Klimenko, S; Klute, M; Knuteson, B; Ko, B R; Kondo, K; Kong, D J; Konigsberg, J; Korytov, A; Kotwal, A V; Kraan, A C; Kraus, J; Kreps, M; Kroll, J; Krumnack, N; Kruse, M; Krutelyov, V; Kubo, T; Kuhlmann, S E; Kuhr, T; Kulkarni, N P; Kusakabe, Y; Kwang, S; Laasanen, A T; Lai, S; Lami, S; Lammel, S; Lancaster, M; Lander, R L; Lannon, K; Lath, A; Latino, G; Lazzizzera, I; Lecompte, T; Lee, J; Lee, J; Lee, Y J; Lee, S W; Lefèvre, R; Leonardo, N; Leone, S; Levy, S; Lewis, J D; Lin, C; Lin, C S; Lindgren, M; Lipeles, E; Liss, T M; Lister, A; Litvintsev, D O; Liu, T; Lockyer, N S; Loginov, A; Loreti, M; Lu, R-S; Lucchesi, D; Lujan, P; Lukens, P; Lungu, G; Lyons, L; Lys, J; Lysak, R; Lytken, E; Mack, P; Macqueen, D; Madrak, R; Maeshima, K; Makhoul, K; Maki, T; Maksimovic, P; Malde, S; Malik, S; Manca, G; Manousakis, A; Margaroli, F; Marginean, R; Marino, C; Marino, C P; Martin, A; Martin, M; Martin, V; Martínez, M; Martínez-Ballarín, R; Maruyama, T; Mastrandrea, P; Masubuchi, T; Matsunaga, H; Mattson, M E; Mazini, R; Mazzanti, P; McFarland, K S; McIntyre, P; McNulty, R; Mehta, A; Mehtala, P; Menzemer, S; Menzione, A; Merkel, P; Mesropian, C; Messina, A; Miao, T; Miladinovic, N; Miles, J; Miller, R; Mills, C; Milnik, M; Mitra, A; Mitselmakher, G; Miyamoto, A; Moed, S; Moggi, N; Mohr, B; Moon, C S; Moore, R; Morello, M; Fernandez, P Movilla; Mülmenstädt, J; Mukherjee, A; Muller, Th; Mumford, R; Murat, P; Mussini, M; Nachtman, J; Nagano, A; Naganoma, J; Nakamura, K; Nakano, I; Napier, A; Necula, V; Neu, C; Neubauer, M S; Nielsen, J; Nodulman, L; Norniella, O; Nurse, E; Oh, S H; Oh, Y D; Oksuzian, I; Okusawa, T; Oldeman, R; Orava, R; Osterberg, K; Pagliarone, C; Palencia, E; Papadimitriou, V; Papaikonomou, A; Paramonov, A A; Parks, B; Pashapour, S; Patrick, J; Pauletta, G; Paulini, M; Paus, C; Pellett, D E; Penzo, A; Phillips, T J; Piacentino, G; Piedra, J; Pinera, L; Pitts, K; Plager, C; Pondrom, L; Portell, X; Poukhov, O; Pounder, N; Prakoshyn, F; Pronko, A; Proudfoot, J; Ptohos, F; Punzi, G; Pursley, J; Rademacker, J; Rahaman, A; Ramakrishnan, V; Ranjan, N; Redondo, I; Reisert, B; Rekovic, V; Renton, P; Rescigno, M; Richter, S; Rimondi, F; Ristori, L; Robson, A; Rodrigo, T; Rogers, E; Rolli, S; Roser, R; Rossi, M; Rossin, R; Roy, P; Ruiz, A; Russ, J; Rusu, V; Saarikko, H; Safonov, A; Sakumoto, W K; Salamanna, G; Saltó, O; Santi, L; Sarkar, S; Sartori, L; Sato, K; Savard, P; Savoy-Navarro, A; Scheidle, T; Schlabach, P; Schmidt, E E; Schmidt, M P; Schmitt, M; Schwarz, T; Scodellaro, L; Scott, A L; Scribano, A; Scuri, F; Sedov, A; Seidel, S; Seiya, Y; Semenov, A; Sexton-Kennedy, L; Sfyrla, A; Shalhout, S Z; Shapiro, M D; Shears, T; Shepard, P F; Sherman, D; Shimojima, M; Shochet, M; Shon, Y; Shreyber, I; Sidoti, A; Sinervo, P; Sisakyan, A; Slaughter, A J; Slaunwhite, J; Sliwa, K; Smith, J R; Snider, F D; Snihur, R; Soderberg, M; Soha, A; Somalwar, S; Sorin, V; Spalding, J; Spinella, F; Spreitzer, T; Squillacioti, P; Stanitzki, M; Staveris-Polykalas, A; Denis, R St; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Stuart, D; Suh, J S; Sukhanov, A; Sun, H; Suslov, I; Suzuki, T; Taffard, A; Takashima, R; Takeuchi, Y; Tanaka, R; Tecchio, M; Teng, P K; Terashi, K; Thom, J; Thompson, A S; Thomson, E; Tipton, P; Tiwari, V; Tkaczyk, S; Toback, D; Tokar, S; Tollefson, K; Tomura, T; Tonelli, D; Torre, S; Torretta, D; Tourneur, S; Trischuk, W; Tsuno, S; Tu, Y; Turini, N; Ukegawa, F; Uozumi, S; Vallecorsa, S; van Remortel, N; Varganov, A; Vataga, E; Vazquez, F; Velev, G; Vellidis, C; Veramendi, G; Veszpremi, V; Vidal, M; Vidal, R; Vila, I; Vilar, R; Vine, T; Vogel, M; Vollrath, I; Volobouev, I; Volpi, G; Würthwein, F; Wagner, P; Wagner, R G; Wagner, R L; Wagner, J; Wagner, W; Wallny, R; Wang, S M; Warburton, A; Waters, D; Weinberger, M; Wester, W C; Whitehouse, B; Whiteson, D; Wicklund, A B; Wicklund, E; Williams, G; Williams, H H; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, C; Wright, T; Wu, X; Wynne, S M; Yagil, A; Yamamoto, K; Yamaoka, J; Yamashita, T; Yang, C; Yang, U K; Yang, Y C; Yao, W M; Yeh, G P; Yoh, J; Yorita, K; Yoshida, T; Yu, G B; Yu, I; Yu, S S; Yun, J C; Zanello, L; Zanetti, A; Zaw, I; Zhang, X; Zhou, J; Zucchelli, S

    2007-10-12

    We present a measurement of the W-boson mass using 200 pb{-1} of data collected in pp[over ] collisions at sqrt[s]=1.96 TeV by the CDF II detector at run II of the Fermilab Tevatron. With a sample of 63 964 W-->enu candidates and 51 128 W-->munu candidates, we measure M_{W}=80 413+/-34{stat}+/-34{syst}=80,413+/-48 MeV/c;{2}. This is the most precise single measurement of the W-boson mass to date.

  17. Measurement of the cross section for prompt isolated diphoton production using the full CDF run II data sample.

    Science.gov (United States)

    Aaltonen, T; Amerio, S; Amidei, D; Anastassov, A; Annovi, A; Antos, J; Apollinari, G; Appel, J A; Arisawa, T; Artikov, A; Asaadi, J; Ashmanskas, W; Auerbach, B; Aurisano, A; Azfar, F; Badgett, W; Bae, T; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Barria, P; Bartos, P; Bauce, M; Bedeschi, F; Behari, S; Bellettini, G; Bellinger, J; Benjamin, D; Beretvas, A; Bhatti, A; Bland, K R; Blumenfeld, B; Bocci, A; Bodek, A; Bortoletto, D; Boudreau, J; Boveia, A; Brigliadori, L; Bromberg, C; Brucken, E; Budagov, J; Budd, H S; Burkett, K; Busetto, G; Bussey, P; Butti, P; Buzatu, A; Calamba, A; Camarda, S; Campanelli, M; Canelli, F; Carls, B; Carlsmith, D; Carosi, R; Carrillo, S; Casal, B; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavaliere, V; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Cho, K; Chokheli, D; Ciocci, M A; Clark, A; Clarke, C; Convery, M E; Conway, J; Corbo, M; Cordelli, M; Cox, C A; Cox, D J; Cremonesi, M; Cruz, D; Cuevas, J; Culbertson, R; d'Ascenzo, N; Datta, M; De Barbaro, P; Demortier, L; Deninno, M; Devoto, F; d'Errico, M; Di Canto, A; Di Ruzza, B; Dittmann, J R; D'Onofrio, M; Donati, S; Dorigo, M; Driutti, A; Ebina, K; Edgar, R; Elagin, A; Erbacher, R; Errede, S; Esham, B; Eusebi, R; Farrington, S; Fernández Ramos, J P; Field, R; Flanagan, G; Forrest, R; Franklin, M; Freeman, J C; Frisch, H; Funakoshi, Y; Garfinkel, A F; Garosi, P; Gerberich, H; Gerchtein, E; Giagu, S; Giakoumopoulou, V; Gibson, K; Ginsburg, C M; Giokaris, N; Giromini, P; Giurgiu, G; Glagolev, V; Glenzinski, D; Gold, M; Goldin, D; Golossanov, A; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González López, O; Gorelov, I; Goshaw, A T; Goulianos, K; Gramellini, E; Grinstein, S; Grosso-Pilcher, C; Group, R C; Guimaraes da Costa, J; Hahn, S R; Han, J Y; Happacher, F; Hara, K; Hare, M; Harr, R F; Harrington-Taber, T; Hatakeyama, K; Hays, C; Heinrich, J; Herndon, M; Hocker, A; Hong, Z; Hopkins, W; Hou, S; Hughes, R E; Husemann, U; Huston, J; Introzzi, G; Iori, M; Ivanov, A; James, E; Jang, D; Jayatilaka, B; Jeon, E J; Jindariani, S; Jones, M; Joo, K K; Jun, S Y; Junk, T R; Kambeitz, M; Kamon, T; Karchin, P E; Kasmi, A; Kato, Y; Ketchum, W; Keung, J; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, S B; Kim, S H; Kim, Y K; Kim, Y J; Kimura, N; Kirby, M; Knoepfel, K; Kondo, K; Kong, D J; Konigsberg, J; Kotwal, A V; Kreps, M; Kroll, J; Kruse, M; Kuhr, T; Kurata, M; Laasanen, A T; Lammel, S; Lancaster, M; Lannon, K; Latino, G; Lee, H S; Lee, J S; Leo, S; Leone, S; Lewis, J D; Limosani, A; Lipeles, E; Liu, H; Liu, Q; Liu, T; Lockwitz, S; Loginov, A; Lucchesi, D; Lueck, J; Lujan, P; Lukens, P; Lungu, G; Lys, J; Lysak, R; Madrak, R; Maestro, P; Malik, S; Manca, G; Manousakis-Katsikakis, A; Margaroli, F; Marino, P; Martínez, M; Matera, K; Mattson, M E; Mazzacane, A; Mazzanti, P; McNulty, R; Mehta, A; Mehtala, P; Mesropian, C; Miao, T; Mietlicki, D; Mitra, A; Miyake, H; Moed, S; Moggi, N; Moon, C S; Moore, R; Morello, M J; Mukherjee, A; Muller, Th; Murat, P; Mussini, M; Nachtman, J; Nagai, Y; Naganoma, J; Nakano, I; Napier, A; Nett, J; Neu, C; Nigmanov, T; Nodulman, L; Noh, S Y; Norniella, O; Oakes, L; Oh, S H; Oh, Y D; Oksuzian, I; Okusawa, T; Orava, R; Ortolan, L; Pagliarone, C; Palencia, E; Palni, P; Papadimitriou, V; Parker, W; Pauletta, G; Paulini, M; Paus, C; Phillips, T J; Piacentino, G; Pianori, E; Pilot, J; Pitts, K; Plager, C; Pondrom, L; Poprocki, S; Potamianos, K; Prokoshin, F; Pranko, A; Ptohos, F; Punzi, G; Ranjan, N; Redondo Fernández, I; Renton, P; Rescigno, M; Riddick, T; Rimondi, F; Ristori, L; Robson, A; Rodriguez, T; Rolli, S; Ronzani, M; Roser, R; Rosner, J L; Ruffini, F; Ruiz, A; Russ, J; Rusu, V; Safonov, A; Sakumoto, W K; Sakurai, Y; Santi, L; Sato, K; Saveliev, V; Savoy-Navarro, A; Schlabach, P; Schmidt, E E; Schwarz, T; Scodellaro, L; Scuri, F; Seidel, S; Seiya, Y; Semenov, A; Sforza, F; Shalhout, S Z; Shears, T; Shepard, P F; Shimojima, M; Shochet, M; Shreyber-Tecker, I; Simonenko, A; Sinervo, P; Sliwa, K; Smith, J R; Snider, F D; Sorin, V; Song, H; Stancari, M; St Denis, R; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Sudo, Y; Sukhanov, A; Suslov, I; Takemasa, K; Takeuchi, Y; Tang, J; Tecchio, M; Teng, P K; Thom, J; Thomson, E; Thukral, V; Toback, D; Tokar, S; Tollefson, K; Tomura, T; Tonelli, D; Torre, S; Torretta, D; Totaro, P; Trovato, M; Ukegawa, F; Uozumi, S; Vázquez, F; Velev, G; Vellidis, C; Vernieri, C; Vidal, M; Vilar, R; Vizán, J; Vogel, M; Volpi, G; Wagner, P; Wallny, R; Wang, S M; Warburton, A; Waters, D; Wester, W C; Whiteson, D; Wicklund, A B; Wilbur, S; Williams, H H; Wilson, J S; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, H; Wright, T; Wu, X; Wu, Z; Yamamoto, K; Yamato, D; Yang, T; Yang, U K; Yang, Y C; Yao, W-M; Yeh, G P; Yi, K; Yoh, J; Yorita, K; Yoshida, T; Yu, G B; Yu, I; Zanetti, A M; Zeng, Y; Zhou, C; Zucchelli, S

    2013-03-08

    This Letter reports a measurement of the cross section for producing pairs of central prompt isolated photons in proton-antiproton collisions at a total energy sqrt[s] = 1.96 TeV using data corresponding to 9.5 fb(-1) integrated luminosity collected with the CDF II detector at the Fermilab Tevatron. The measured differential cross section is compared to three calculations derived from the theory of strong interactions. These include a prediction based on a leading order matrix element calculation merged with a parton shower model, a next-to-leading order calculation, and a next-to-next-to-leading order calculation. The first and last calculations reproduce most aspects of the data, thus showing the importance of higher-order contributions for understanding the theory of strong interaction and improving measurements of the Higgs boson and searches for new phenomena in diphoton final states.

  18. The black hole mass of NGC 4151. II. Stellar dynamical measurement from near-infrared integral field spectroscopy

    International Nuclear Information System (INIS)

    Onken, Christopher A.; Ferrarese, Laura; Valluri, Monica; Brown, Jonathan S.; McGregor, Peter J.; Peterson, Bradley M.; Pogge, Richard W.; Bentz, Misty C.; Vestergaard, Marianne; Storchi-Bergmann, Thaisa; Riffel, Rogemar A.

    2014-01-01

    We present a revised measurement of the mass of the central black hole (M BH ) in the Seyfert 1 galaxy NGC 4151. The new stellar dynamical mass measurement is derived by applying an axisymmetric orbit-superposition code to near-infrared integral field data obtained using adaptive optics with the Gemini Near-infrared Integral Field Spectrograph (NIFS). When our models attempt to fit both the NIFS kinematics and additional low spatial resolution kinematics, our results depend sensitively on how χ 2 is computed—probably a consequence of complex bar kinematics that manifest immediately outside the nuclear region. The most robust results are obtained when only the high spatial resolution kinematic constraints in the nuclear region are included in the fit. Our best estimates for the black hole mass and H-band mass-to-light ratio are M BH ∼ 3.76 ± 1.15 × 10 7 M ☉ (1σ error) and Y H ∼ 0.34 ± 0.03 M ☉ /L ☉ (3σ error), respectively (the quoted errors reflect the model uncertainties). Our black hole mass measurement is consistent with estimates from both reverberation mapping (3.57 −0.37 +0.45 ×10 7 M ⊙ ) and gas kinematics (3.0 −2.2 +0.75 ×10 7 M ⊙ ; 1σ errors), and our best-fit mass-to-light ratio is consistent with the photometric estimate of Y H = 0.4 ± 0.2 M ☉ /L ☉ . The NIFS kinematics give a central bulge velocity dispersion σ c = 116 ± 3 km s –1 , bringing this object slightly closer to the M BH -σ relation for quiescent galaxies. Although NGC 4151 is one of only a few Seyfert 1 galaxies in which it is possible to obtain a direct dynamical black hole mass measurement—and thus, an independent calibration of the reverberation mapping mass scale—the complex bar kinematics makes it less than ideally suited for this purpose.

  19. Structural Integrity in Measures of Self Concept.

    Science.gov (United States)

    Stenner, A. Jackson; Katzenmeyer, W.G.

    Structural integrity of a measure is defined in terms of its replicability, constancy, invariance, and stability. Work completed in the development and validation of the Self Observation Scales (SOS) Primary Level (Stenner and Katzenmeyer, 1973) serves to illustrate one method of establishing structural integrity. The name of each scale of the SOS…

  20. Results from SAGE II

    International Nuclear Information System (INIS)

    Nico, J.S.

    1994-01-01

    The Russian-American Gallium solar neutrino Experiment (SAGE) began the second phase of operation (SAGE II) in September of 1992. Monthly measurements of the integral flux of solar neutrinos have been made with 55 tonnes of gallium. The K-peak results of the first nine runs of SAGE II give a capture rate of 66 -13 +18 (stat) -7 +5 (sys) SNU. Combined with the SAGE I result of 73 -16 +18 (stat) -7 5 (sys) SNU, the capture rate is 69 -11 +11 (stat) -7 +5 (sys) SNU. This represents only 52%--56% of the capture rate predicted by different Standard Solar Models

  1. La perduración del poder en un espacio arquitectónico simbólico. La torre T-3 del asentamiento protohistórico de L’Assut (Tivenys, Baix Ebre, Tarragona)

    OpenAIRE

    Diloli Fons, Jordi

    2009-01-01

    Not available

    Los trabajos arqueológicos efectuados en el asentamiento protohistórico de L’Assut (Tivenys, Baix Ebre, Tarragona) han permitido localizar una torre fortificada de planta circular (T3), construida durante el siglo VI a.n.e. Este tipo de estructuras se ha relacionado en áreas vecinas con una intencionalidad elitista ejercida por un grupo de poder emergente mediante el simbolismo arquitectónico, desapareciendo el modelo a inicios del Ibérico Pleno. En el caso de...

  2. Cross section measurement and integral test for several activation reactions using T + d and thick-Li + d sources

    International Nuclear Information System (INIS)

    Dumais, J.R.; Tanaka, S.; Odano, N.; Iwasaki, S.; Sugiyama, K.

    1988-01-01

    Recent activities on the area of the cross section measurement for several activation reactions at Department of Nucl. Eng., Tohoku Univ. are described. The first subject is the cross section measurement for (n,2n) reaction on aluminum using the RTNS-II neutron source. Cross sections with rather small error band were obtained for the incident neutron energies from 14 to 14.7 MeV. The second one is the status of the program for the integral experiments on several reactions using the thick Li + d source at Tohoku Fast Neutron Lab. The experimental results showed the usefullness of the source as a tool for the cross section assessment. (author)

  3. NEUTRON-INDUCED SWELLING OF Fe-Cr BINARY ALLOYS IN FFTF AT ∼400 DEGREES C

    International Nuclear Information System (INIS)

    Garner, Francis A.; Greenwood, Lawrence R.; Okita, Taira; Sekimura, Naoto; Wolfer, W. G.

    2002-01-01

    The purpose of this effort is to determine the influence of dpa rate, He/dpa ratio and composition on the void swelling of simple binary Fe-Cr alloys. Contrary to the behavior of swelling of model fcc Fe-Cr-Ni alloys irradiated in the same FFTF-MOTA experiment, model bcc Fe-Cr alloys do not exhibit a dependence of swelling on dpa rate at approximately 400 degrees C. This is surprising in that an apparent flux-sensitivity was observed in an earlier comparative irradiation of Fe-Cr binaries conducted in EBR-II and FFTF. The difference in behavior is ascribed to the higher helium generation rates of Fe-Cr alloys in EBR-II compared to that of FFTF, and also the fact that lower dpa rates in FFTF are accompanied by progressively lower helium generation rates.

  4. Integrating institutional bribery and behavioral measures of bribery

    OpenAIRE

    Rose, Richard; Peiffer, Caryn

    2016-01-01

    Bribery involves individuals exchanging material benefits for a service of a public institution. To understand the process of bribery we need to integrate measures of individual behaviour and institutional attributes rather than rely exclusively on surveys of individual perceptions and experience or macro-level corruption indexes of national institutions. This paper integrates institutional and behavioural measures to show that where you live and who you are have independent influence on whet...

  5. Biologically active new Fe(II, Co(II, Ni(II, Cu(II, Zn(II and Cd(II complexes of N-(2-thienylmethylenemethanamine

    Directory of Open Access Journals (Sweden)

    C. SPÎNU

    2008-04-01

    Full Text Available Iron(II, cobalt(II, nickel (II, copper (II, zinc(II and cadmium(II complexes of the type ML2Cl2, where M is a metal and L is the Schiff base N-(2-thienylmethylenemethanamine (TNAM formed by the condensation of 2-thiophenecarboxaldehyde and methylamine, were prepared and characterized by elemental analysis as well as magnetic and spectroscopic measurements. The elemental analyses suggest the stoichiometry to be 1:2 (metal:ligand. Magnetic susceptibility data coupled with electronic, ESR and Mössbauer spectra suggest a distorted octahedral structure for the Fe(II, Co(II and Ni(II complexes, a square-planar geometry for the Cu(II compound and a tetrahedral geometry for the Zn(II and Cd(II complexes. The infrared and NMR spectra of the complexes agree with co-ordination to the central metal atom through nitrogen and sulphur atoms. Conductance measurements suggest the non-electrolytic nature of the complexes, except for the Cu(II, Zn(II and Cd(II complexes, which are 1:2 electrolytes. The Schiff base and its metal chelates were screened for their biological activity against Escherichia coli, Staphylococcus aureus and Pseudomonas aeruginosa and the metal chelates were found to possess better antibacterial activity than that of the uncomplexed Schiff base.

  6. Joint Integration Test Facility (JITF) Engineering II Performance Measurement Plans

    National Research Council Canada - National Science Library

    Boucher, Joanne

    2001-01-01

    ..., effectiveness, and accountability in federal programs and spending. The plan establishes six separate performance measurements, which correlate directly to customer satisfaction, Intelligence Mission Application (IMA...

  7. First results from SAGE II

    International Nuclear Information System (INIS)

    Aburashitov, J.N.; Faizov, E.L.; Gavrin, V.N.; Gusev, A.O.; Kalikhov, A.V.; Knodel, T.V.; Knyshenko, I.I.; Kornoukhov, V.N.; Mirmov, I.N.; Pshukov, A.M.; Shalagin, A.M.; Shikhin, A.A.; Timofeyev, P.V.; Veretenkin, E.P.; Vermul, V.M.; Zatsepin, G.T.; Bowles, T.J.; Nico, J.S.; Teasdale, W.A.; Wark, D.L.; Wilkerson, J.F.; Cleveland, B.T.; Daily, T.; Davis, R. Jr.; Lande, K.; Lee, C.K.; Wildenhain, P.W.; Elliott, S.R.; Cherry, M.L.

    1995-01-01

    The Russian-American Gallium solar neutrino Experiment (SAGE) began the second phase of operation (SAGE II) in September of 1992. Monthly measurements of the integral flux of solar neutrinos have been made with 55 tonnes of gallium. The K-peak results of the first five runs of SAGE II give a capture rate of 76 +21 -18 (stat) +5 -7 (sys) SNU. Combined with the SAGE I result, the capture rate is 74 +13 -12 (stat) +5 -7 (sys) SNU. This represents only 56%--60% of the capture rate predicted by different Standard Solar Models. copyright 1995 American Institute of Physics

  8. Energy Levels, wavelengths and hyperfine structure measurements of Sc II

    Science.gov (United States)

    Hala, Fnu; Nave, Gillian

    2018-01-01

    Lines of singly ionized Scandium (Sc II) along with other Iron group elements have been observed [1] in the region surrounding the massive star Eta Carinae [2,3] called the strontium filament (SrF). The last extensive analysis of Sc II was the four-decade old work of Johansson & Litzen [4], using low-resolution grating spectroscopy. To update and extend the Sc II spectra, we have made observation of Sc/Ar, Sc/Ne and Sc/Ge/Ar hollow cathode emission spectrum on the NIST high resolution FT700 UV/Vis and 2 m UV/Vis/IR Fourier transform spectrometers (FTS). More than 850 Sc II lines have been measured in the wavelength range of 187 nm to 3.2 μm. connecting a total of 152 energy levels. The present work also focuses to resolve hyperfine structure (HFS) in Sc II lines. We aim to obtain accurate transition wavelengths, improved energy levels and HFS constants of Sc II. The latest results from work in progress will be presented.Reference[1] Hartman H, Gull T, Johansson S and Smith N 2004 Astron. Astrophys. 419 215[2] Smith N, Morse J A and Gull T R 2004 Astrophys. J. 605 405[3] Davidson K and Humphreys R M 1997 Annu. Rev. Astron. Astrophys. 35[4] Johansson S and Litzén U 1980 Phys. Scr. 22 49

  9. Constituting fully integrated visual analysis system for Cu(II) on TiO₂/cellulose paper.

    Science.gov (United States)

    Li, Shun-Xing; Lin, Xiaofeng; Zheng, Feng-Ying; Liang, Wenjie; Zhong, Yanxue; Cai, Jiabai

    2014-07-15

    As a cheap and abundant porous material, cellulose filter paper was used to immobilize nano-TiO2 and denoted as TiO2/cellulose paper (TCP). With high adsorption capacity for Cu(II) (more than 1.65 mg), TCP was used as an adsorbent, photocatalyst, and colorimetric sensor at the same time. Under the optimum adsorption conditions, i.e., pH 6.5 and 25 °C, the adsorption ratio of Cu(II) was higher than 96.1%. Humic substances from the matrix could be enriched onto TCP but the interference of their colors on colorimetric detection could be eliminated by the photodegradation. In the presence of hydroxylamine, neocuproine, as a selective indicator, was added onto TCP, and a visual color change from white to orange was generated. The concentration of Cu(II) was quantified by the color intensity images using image processing software. This fully integrated visual analysis system was successfully applied for the detection of Cu(II) in 10.0 L of drinking water and seawater with a preconcentration factor of 10(4). The log-linear calibration curve for Cu(II) was in the range of 0.5-50.0 μg L(-1) with a determination coefficient (R(2)) of 0.985 and its detection limit was 0.073 μg L(-1).

  10. Synergistic enhancement of supercapacitance upon integration of nickel (II) octa[(3,5-biscarboxylate)-phenoxy] phthalocyanine with SWCNT-phenylamine

    CSIR Research Space (South Africa)

    Agboola, BO

    2010-06-01

    Full Text Available Supercapacitive behaviour of a novel functional material, nickel (II) octa [(3,5-biscarboxylate)-phenoxy] phthalocyanine (NiOBCPPc) upon covalent integration with phenylamine functionalized single-walled carbon nanotubes (SWCNT...

  11. Combined measurements on stationary flow of helium II

    International Nuclear Information System (INIS)

    Ijsselstein, R.R.

    1979-01-01

    Transport phenomena in helium II can in principle be described by a two fluid model. One of the fluids, the superfluid component, carries no entropy and has no viscosity while its velocity field is curl free. The other, the normal component, behaves like an ordinary fluid and carries the entropy of the whole liquid. In measuring flow two different methods are required because of the two independent velocity fields. This thesis describes an experiment where both techniques are applied to flow through a capillary of 0.62 mm, enabling direct comparison. The apparatus is described, and details of the measuring techniques are reported. An extended treatment of second-sound phenomena in a Helmholtz resonator is given. The results of the measurements are reported and discussed. (Auth.)

  12. KIT/KPS of Qinshan phase-II and a discussion on integrated information management and automatic control

    International Nuclear Information System (INIS)

    Yan Changhui

    2001-01-01

    Centralized Data Processing and Safety Panel (KIT/KPS) of Qinshan Phase-II power project is described, and the necessity and engineering scheme is presented of integrated information management and automatic control that would achieve in power plant according to the technology scheme and technology trait of KIT/KPS

  13. Path integral measure for first-order and metric gravities

    International Nuclear Information System (INIS)

    Aros, Rodrigo; Contreras, Mauricio; Zanelli, Jorge

    2003-01-01

    The equivalence between the path integrals for first-order gravity and the standard torsion-free, metric gravity in 3 + 1 dimensions is analysed. Starting with the path integral for first-order gravity, the correct measure for the path integral of the metric theory is obtained

  14. The black hole mass of NGC 4151. II. Stellar dynamical measurement from near-infrared integral field spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Onken, Christopher A.; Ferrarese, Laura [Herzberg Institute of Astrophysics, National Research Council of Canada, 5071 West Saanich Road, Victoria, BC V9E 2E7 (Canada); Valluri, Monica; Brown, Jonathan S. [Department of Astronomy, University of Michigan, 500 Church Street, Ann Arbor, MI 48109-1042 (United States); McGregor, Peter J. [Research School of Astronomy and Astrophysics, The Australian National University, Canberra, ACT 2611 (Australia); Peterson, Bradley M.; Pogge, Richard W. [Department of Astronomy, The Ohio State University, 140 West 18th Avenue, Columbus, OH 43210 (United States); Bentz, Misty C. [Department of Physics and Astronomy, Georgia State University, 25 Park Place, Office 610, Atlanta, GA 30303 (United States); Vestergaard, Marianne [Dark Cosmology Centre, The Niels Bohr Institute, Copenhagen University, Juliane Maries Vej 30, DK-2100 Copenhagen Ø (Denmark); Storchi-Bergmann, Thaisa [Universidade Federal do Rio Grande do Sul, Instituto de Física, CP 15051, Porto Alegre 91501-970, RS (Brazil); Riffel, Rogemar A., E-mail: christopher.onken@anu.edu.au, E-mail: mvalluri@umich.edu [Departamento de Física, Centro de Ciências Naturais e Exatas, Universidade Federal de Santa Maria, 97105-900 Santa Maria, RS (Brazil)

    2014-08-10

    We present a revised measurement of the mass of the central black hole (M{sub BH} ) in the Seyfert 1 galaxy NGC 4151. The new stellar dynamical mass measurement is derived by applying an axisymmetric orbit-superposition code to near-infrared integral field data obtained using adaptive optics with the Gemini Near-infrared Integral Field Spectrograph (NIFS). When our models attempt to fit both the NIFS kinematics and additional low spatial resolution kinematics, our results depend sensitively on how χ{sup 2} is computed—probably a consequence of complex bar kinematics that manifest immediately outside the nuclear region. The most robust results are obtained when only the high spatial resolution kinematic constraints in the nuclear region are included in the fit. Our best estimates for the black hole mass and H-band mass-to-light ratio are M{sub BH} ∼ 3.76 ± 1.15 × 10{sup 7} M{sub ☉} (1σ error) and Y{sub H} ∼ 0.34 ± 0.03 M{sub ☉}/L{sub ☉} (3σ error), respectively (the quoted errors reflect the model uncertainties). Our black hole mass measurement is consistent with estimates from both reverberation mapping (3.57{sub −0.37}{sup +0.45}×10{sup 7} M{sub ⊙}) and gas kinematics (3.0{sub −2.2}{sup +0.75}×10{sup 7} M{sub ⊙}; 1σ errors), and our best-fit mass-to-light ratio is consistent with the photometric estimate of Y{sub H} = 0.4 ± 0.2 M{sub ☉}/L{sub ☉}. The NIFS kinematics give a central bulge velocity dispersion σ{sub c} = 116 ± 3 km s{sup –1}, bringing this object slightly closer to the M{sub BH}-σ relation for quiescent galaxies. Although NGC 4151 is one of only a few Seyfert 1 galaxies in which it is possible to obtain a direct dynamical black hole mass measurement—and thus, an independent calibration of the reverberation mapping mass scale—the complex bar kinematics makes it less than ideally suited for this purpose.

  15. Synthesis and spectroscopic studies of biologically active tetraazamacrocyclic complexes of Mn(II, Co(II, Ni(II, Pd(II and Pt(II

    Directory of Open Access Journals (Sweden)

    Monika Tyagi

    2014-01-01

    Full Text Available Complexes of Mn(II, Co(II, Ni(II, Pd(II and Pt(II were synthesized with the macrocyclic ligand, i.e., 2,3,9,10-tetraketo-1,4,8,11-tetraazacycoletradecane. The ligand was prepared by the [2 + 2] condensation of diethyloxalate and 1,3-diamino propane and characterized by elemental analysis, mass, IR and 1H NMR spectral studies. All the complexes were characterized by elemental analysis, molar conductance, magnetic susceptibility measurements, IR, electronic and electron paramagnetic resonance spectral studies. The molar conductance measurements of Mn(II, Co(II and Ni(II complexes in DMF correspond to non electrolyte nature, whereas Pd(II and Pt(II complexes are 1:2 electrolyte. On the basis of spectral studies an octahedral geometry has been assigned for Mn(II, Co(II and Ni(II complexes, whereas square planar geometry assigned for Pd(II and Pt(II. In vitro the ligand and its metal complexes were evaluated against plant pathogenic fungi (Fusarium odum, Aspergillus niger and Rhizoctonia bataticola and some compounds found to be more active as commercially available fungicide like Chlorothalonil.

  16. Overview of gamma-ray energy deposition and spectra in fast reactor environments

    International Nuclear Information System (INIS)

    Gold, R.

    1977-01-01

    Efforts to define gamma-ray heating in Breeder Reactor (BR) environments are reviewed. This critique is restricted to programmatic activities in the United States, as best exemplified by current practice for the Experimental Breeder Reactor II (EBR-II). Future needs are also addressed in terms of requirements for the Fast Flux Test Facility (FFTF). Experimental efforts and theoretical analyses are surveyed for both high and low power environments. Special emphasis is placed on experimental techniques for calorimetry, temperature measurement, dosimetry and spectrometry. The relation between neutron and gamma-ray calculations is stressed with particular attention given to contrasting analytical techniques and basic nuclear data requirements. Wherever possible comparisons between theory and experiment are cited

  17. Transmittance measurements at DIRT-II

    Science.gov (United States)

    Curcio, J. A.; Haught, K. M.; Woytko, M. A.

    1980-07-01

    This is a report on the NRL experiments at the DIRT-II tests sponsored by the Atmospheric Sciences Laboratory at the White Sands Missile Range in July 1970. The NRL experiment was designed to measure spectral transmittance through smoke and dust clouds generated by detonations of various explosive charges and also by impact of artillery rounds. Spectral transmission data as a function of time for 0.55 micrometers, 1.06 micrometers, and 10.37 micrometers were obtained for 63 events comprised of static detonations and artillery rounds. Transmission data for 1.06 micrometers, in most cases were similar and equal to 0.55 micrometers. In dry soil conditions the 10.37 micrometers channel showed higher transmittance values than the visible channel. There are indications that 10.37 micrometers transmittance in wet soil events is lower than visible presumably because of strong liquid water absorption at the IR wavelength.

  18. Synthesis and structural characterization of nickel(II), cobalt(II), Zinc(II), manganese(II), cadmium(II) and uranium(VI) complexes of α-oximinoacetoacet-o/p-anisidide thiosemicarbazone

    International Nuclear Information System (INIS)

    Patel, P.S.; Patel, M.M.; Ray, R.M.

    1993-01-01

    A few metal complexes of α-oximinoacetoacet-o/p-anisidide thiosemicarbazones (OAOATS)/(OAPATS) with Ni(II), Co(II), Zn(II), Mn(II), Hg(II), Cd(II) and UO 2 (II) have been prepared and characterized by elemental analyses, conductivity, differential scanning calorimetry study, thermogravimetric analyses and infrared and electronic spectral measurements in conjunction with magnetic susceptibility measurements at room temperature. They have also been tested for their antimicrobial activities. (author). 24 refs., 2 tabs

  19. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  20. On the Efficiency of Connection Charges---Part II: Integration of Distributed Energy Resources

    OpenAIRE

    Munoz-Alvarez, Daniel; Garcia-Franco, Juan F.; Tong, Lang

    2017-01-01

    This two-part paper addresses the design of retail electricity tariffs for distribution systems with distributed energy resources (DERs). Part I presents a framework to optimize an ex-ante two-part tariff for a regulated monopolistic retailer who faces stochastic wholesale prices on the one hand and stochastic demand on the other. In Part II, the integration of DERs is addressed by analyzing their endogenous effect on the optimal two-part tariff and the induced welfare gains. Two DER integrat...

  1. Calculus II For Dummies

    CERN Document Server

    Zegarelli, Mark

    2012-01-01

    An easy-to-understand primer on advanced calculus topics Calculus II is a prerequisite for many popular college majors, including pre-med, engineering, and physics. Calculus II For Dummies offers expert instruction, advice, and tips to help second semester calculus students get a handle on the subject and ace their exams. It covers intermediate calculus topics in plain English, featuring in-depth coverage of integration, including substitution, integration techniques and when to use them, approximate integration, and improper integrals. This hands-on guide also covers sequences and series, wit

  2. Measuring Evapotranspiration of five Alley Cropping systems in Germany using the Eddy-Covariance- and Bowen-Ratio Energy-Balance methods

    Science.gov (United States)

    Markwitz, Christian; Knohl, Alexander; Siebicke, Lukas

    2017-04-01

    The inclusion of trees into the agricultural landscape of Europe is gaining popularity as a source for energy production. Fast growing tree species such as poplar or willow are included as short rotation coppice or alley cropping systems, which consist of tree alleys interleaved by annual rotating crops or perennial grasslands. Estimating turbulent fluxes of those systems using the eddy-covariance- (ECEB) and bowen-ratio energy-balance (BREB) method is challenging due to the methods limitation to horizontally homogeneous terrain and steady state conditions. As the conditions are not fulfilled for those systems the energy-balance is commonly not fully closed, with the non-closure being site specific. An underestimation of measured heat fluxes leads to an overestimation of the latent heat fluxes inferred from the ECEB method. The aim of our study is to 1) quantify the site specific non-closure of the energy-balance and 2) characterize the performance of both methods, compared to direct eddy-covariance measurements using a high frequency infra-red gas analyzer (LI-7200, Licor Inc.). To assess continuous evapotranspiration (ET) rates on a 30-minute time scale we installed a combined ECEB and BREB system at five alley cropping and five agricultural reference sites across Germany. For time periods of four weeks we performed direct eddy covariance flux measurements for H2O and CO2 over one crop- and one grassland alley cropping- and their respective reference systems during the growing season of 2016. We found a non-closure between 21 and 26 % for all sites, considering all day- and night-time data. The residual energy was highest during the morning and lowest in the afternoon. Related to that the energy-balance ratio (EBR), i.e. the ratio between the turbulent heat fluxes and available energy, was below one in the morning hours and increased slightly during the day up to 1.8, until the EBR decreased sharply after sunset. The EBR correlated to the daily cycle of solar

  3. Identification and management of plant aging and life extension issues for a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    King, R.W.; Perry, W.H.

    1991-06-01

    Experimental Breeder Reactor 2 (EBR-2) is a pool-type sodium-cooled fast reactor that supports extensive experimental, test and demonstration programs while providing electrical power to the local grid. EBR-2 is a US Department of Energy Facility located at the Idaho National Engineering Laboratory and operated by Argonne National Laboratory (ANL). The current mission of EBR-2 is to serve as the operational prototype for the Integral Fast Reactor demonstration program. This mission and other programs require EBR-2 to operate reliability to a 40-year lifetime, a significant extension beyond the five to ten year life originally planned for the facility. The benefits of operating EBR-2 in the extended-life mode are important for providing long-term operational performance data for a sodium-cooled fast reactor that is not available elsewhere. Identification and preliminary assessment of potential life-limiting factors indicate that, with appropriate consideration given in the design phase, the sodium-cooled plant has potential for a very long operational lifetime. Achievement of a 40-year lifetime with high reliability is important not only for achieving the near-term goals of the EBR-2/IFR programs, but for the advancement of the liquid-metal-cooled reactor concept to the demonstration/commercialization phase. Key features make extended-life operation feasible based on the use of sodium as the primary coolant: low-pressure, high thermal capacity primary system and a low-pressure secondary system requiring no active valves; and limited corrosion of components. 2 refs

  4. The Role Of Grenelle II In Corporate Social Responsibility Integrated Reporting

    Directory of Open Access Journals (Sweden)

    Radoslaw Wolniak

    2013-12-01

    Full Text Available There is in the paper, the analysis of the Grenelle II act from Corporate Social Responsibility reporting point of view. France has long been an important global champion of corporate sustainability reporting. While this trend has advanced the integration of CSR into business practices, the increasing number of regulations has also made it more complicated and costly for international companies to create reports that fulfill the differing requirements of each foreign law. The Grenelle 2 law adopted in July 2010, implements the Grenelle 1 goals. It sets specific objectives, strategic, incentives and regulations. This Act aims at forcing companies to progress in reporting their environmental and social information: there is no sanction (such as fines, requirements are on a “comply or explain” basis and the implementation is progressive. The main advantage of such an Grenelle II Act is a beginning and will have to be rewritten and improved, taking into account the international movement – is to shift minds regarding CSR. CSR and extra-financial information has nothing to do with communication anymore: it is clearly a strategic and management process, facilitated by the reporting process.

  5. Kr II laser-induced fluorescence for measuring plasma acceleration.

    Science.gov (United States)

    Hargus, W A; Azarnia, G M; Nakles, M R

    2012-10-01

    We present the application of laser-induced fluorescence of singly ionized krypton as a diagnostic technique for quantifying the electrostatic acceleration within the discharge of a laboratory cross-field plasma accelerator also known as a Hall effect thruster, which has heritage as spacecraft propulsion. The 728.98 nm Kr II transition from the metastable 5d(4)D(7/2) to the 5p(4)P(5/2)(∘) state was used for the measurement of laser-induced fluorescence within the plasma discharge. From these measurements, it is possible to measure velocity as krypton ions are accelerated from near rest to approximately 21 km/s (190 eV). Ion temperature and the ion velocity distributions may also be extracted from the fluorescence data since available hyperfine splitting data allow for the Kr II 5d(4)D(7/2)-5p(4)P(5/2)(∘) transition lineshape to be modeled. From the analysis, the fluorescence lineshape appears to be a reasonable estimate for the relatively broad ion velocity distributions. However, due to an apparent overlap of the ion creation and acceleration regions within the discharge, the distributed velocity distributions increase ion temperature determination uncertainty significantly. Using the most probable ion velocity as a representative, or characteristic, measure of the ion acceleration, overall propellant energy deposition, and effective electric fields may be calculated. With this diagnostic technique, it is possible to nonintrusively characterize the ion acceleration both within the discharge and in the plume.

  6. Photosystem II electron flow as a measure for phytoplankton gross primary production = [Fotosysteem II elektronentransport als een maat voor de bruto primaire produktie van fytoplankton

    NARCIS (Netherlands)

    Geel, C.

    1997-01-01

    Saturating pulse fluorescence measurements, well known from studies of higher plants for determination of photosystem II (PS II) characteristics, were applied to cultures of the green alga Dunaliella teitiolecta (Chapter 2). The actual efficiency of PS IIPS

  7. The flow measurement methods for the primary system of integral reactors

    International Nuclear Information System (INIS)

    Lee, J.; Seo, J. K.; Lee, D. J.

    2001-01-01

    It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

  8. AmeriFlux Measurement Network: Science Team Research

    Energy Technology Data Exchange (ETDEWEB)

    Law, B E

    2012-12-12

    Research involves analysis and field direction of AmeriFlux operations, and the PI provides scientific leadership of the AmeriFlux network. Activities include the coordination and quality assurance of measurements across AmeriFlux network sites, synthesis of results across the network, organizing and supporting the annual Science Team Meeting, and communicating AmeriFlux results to the scientific community and other users. Objectives of measurement research include (i) coordination of flux and biometric measurement protocols (ii) timely data delivery to the Carbon Dioxide Information and Analysis Center (CDIAC); and (iii) assurance of data quality of flux and ecosystem measurements contributed by AmeriFlux sites. Objectives of integration and synthesis activities include (i) integration of site data into network-wide synthesis products; and (ii) participation in the analysis, modeling and interpretation of network data products. Communications objectives include (i) organizing an annual meeting of AmeriFlux investigators for reporting annual flux measurements and exchanging scientific information on ecosystem carbon budgets; (ii) developing focused topics for analysis and publication; and (iii) developing data reporting protocols in support of AmeriFlux network goals.

  9. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  10. Numerical method for solving integral equations of neutron transport. II

    International Nuclear Information System (INIS)

    Loyalka, S.K.; Tsai, R.W.

    1975-01-01

    In a recent paper it was pointed out that the weakly singular integral equations of neutron transport can be quite conveniently solved by a method based on subtraction of singularity. This previous paper was devoted entirely to the consideration of simple one-dimensional isotropic-scattering and one-group problems. The present paper constitutes interesting extensions of the previous work in that in addition to a typical two-group anisotropic-scattering albedo problem in the slab geometry, the method is also applied to an isotropic-scattering problem in the x-y geometry. These results are compared with discrete S/sub N/ (ANISN or TWOTRAN-II) results, and for the problems considered here, the proposed method is found to be quite effective. Thus, the method appears to hold considerable potential for future applications. (auth)

  11. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

  12. Man/machine interface algorithm for advanced delayed-neutron signal characterization system

    International Nuclear Information System (INIS)

    Gross, K.C.

    1985-01-01

    The present failed-element rupture detector (FERD) at Experimental Breeder Reactor II (EBR-II) consists of a single bank of delayed-neutron (DN) detectors at a fixed transit time from the core. Plans are currently under way to upgrade the FERD in 1986 and provide advanced DN signal characterization capability that is embodied in an equivalent-recoil-area (ERA) meter. The new configuration will make available to the operator a wealth of quantitative diagnostic information related to the condition and dynamic evolution of a fuel breach. The diagnostic parameters will include a continuous reading of the ERA value for the breach; the transit time, T/sub tr/, for DN emitters traveling from the core to the FERD; and the isotopic holdup time, T/sub h/, for the source. To enhance the processing, interpretation, and display of these parameters to the reactor operator, a man/machine interface (MMI) algorithm has been developed to run in the background on EBR-II's data acquisition system (DAS). The purpose of this paper is to describe the features and implementation of this newly developed MMI algorithm

  13. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  14. SAGE II measurements of early Pinatubo aerosols

    Science.gov (United States)

    Mccormick, M. P.; Veiga, R. E.

    1992-01-01

    SAGE II satellite measurements of the Mt. Pinatubo eruption cloud in the stratosphere during June, July, and early August 1991 show that aerosols in the tropics reached as high as 29 km altitude with most of the cloud between 20 and 25 km. The most optically thick portions of the cloud covered latitudes from 10 deg S to 30 deg N during the early part of this period. By late July, high stratospheric optical depths were observed to at least 70 deg N, with the high values north of about 30 deg N from layers below 20 km. High pressure systems in both hemispheres were observed to be correlated with the movement of volcanic material at 21 km into the westerly jet stream at high southern latitudes and similarly to high northern latitudes at 16 km. By August, the entire Southern Hemisphere had experienced a 10-fold increase in optical depth relative to early July due to layers above 20 km. Initial mass calculations using SAGE II data place the aerosol produced from this eruption at 20 to 30 megatons, well above the 12 megatons produced by El Chichon.

  15. Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-12-01

    The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

  16. Medical information, communication, and archiving system (MICAS): Phase II integration and acceptance testing

    Science.gov (United States)

    Smith, Edward M.; Wandtke, John; Robinson, Arvin E.

    1999-07-01

    The Medical Information, Communication and Archive System (MICAS) is a multi-modality integrated image management system that is seamlessly integrated with the Radiology Information System (RIS). This project was initiated in the summer of 1995 with the first phase being installed during the first half of 1997 and the second phase installed during the summer of 1998. Phase II enhancements include a permanent archive, automated workflow including modality worklist, study caches, NT diagnostic workstations with all components adhering to Digital Imaging and Communications in Medicine (DICOM) standards. This multi-vendor phased approach to PACS implementation is designed as an enterprise-wide PACS to provide images and reports throughout our healthcare network. MICAS demonstrates that aa multi-vendor open system phased approach to PACS is feasible, cost-effective, and has significant advantages over a single vendor implementation.

  17. Actinide recycle potential in the IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    Rising concern about the greenhouse effect reinforces the need to reexamine the question of a next-generation reactor concept that can contribute significantly toward substitution for fossil-based energy generation. Even with only the nuclear capacity on-line today, world-wide reasonably assured uranium resources would last for only about 50 years. If nuclear is to make a significant contribution, breeding is a fundamental requirement. Uranium resources can then be extended by two orders of magnitude, making nuclear essentially a renewable energy source. The key technical elements of the IFR concept are metallic fuel and fuel cycle technology based on pyroprocessing. Pyroprocessing is radically different from the conventional PUREX reprocessing developed for the LWR oxide fuel. Chemical feasibility of pyroprocessing has been demonstrated. The next major step in the IFR development program will be the full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. IFR fuel cycle closure based on pyroprocessing can also have a dramatic impact on the waste management options, and in particular on the actinide recycling. 6 figs

  18. Benchmark Analysis of EBR-II Shutdown Heat Removal Tests

    International Nuclear Information System (INIS)

    2017-08-01

    This publication presents the results and main achievements of an IAEA coordinated research project to verify and validate system and safety codes used in the analyses of liquid metal thermal hydraulics and neutronics phenomena in sodium cooled fast reactors. The publication will be of use to the researchers and professionals currently working on relevant fast reactors programmes. In addition, it is intended to support the training of the next generation of analysts and designers through international benchmark exercises

  19. Gamma heated subassembly for sodium boiling experiments

    International Nuclear Information System (INIS)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed

  20. Gamma heated subassembly for sodium boiling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed.

  1. Measurements of the neutral current e{sup {+-}}p cross sections using longitudinally polarised lepton beams at HERA II

    Energy Technology Data Exchange (ETDEWEB)

    Nikiforov, A.

    2007-01-18

    This thesis presents inclusive e{sup {+-}}p single and double differential cross sections for neutral current deep inelastic scattering measured as functions of the four-momentum transfer squared Q{sup 2} and the Bjorken variable x in interactions of longitudinally polarised leptons with unpolarised protons using the H1 detector at HERA II. An overview of the phenomenology of deep inelastic scattering is given and the experimental apparatus as well as the measurement and analysis procedures are described. The analysis is based on e{sup +}p data taken in 2003-04 and e{sup -}p data taken in 2005 at a centre-of-mass energy of {radical}(s)=318 GeV, with integrated luminosities of 47.6 pb{sup -1} and 98.4 pb{sup -1} for the e{sup +}p and e{sup -}p samples, respectively. The cross sections are measured in the range of 200measurements are used to study polarisation e ects in neutral current interactions and to determine the structure function xF{sub 3}. The new HERA II data are combined together with previously published data from HERA I to determine the structure function xF{sub 3} with improved precision. Furthermore, this measurement is combined with the corresponding ZEUS measurement to provide the most accurate measurement of the interference structure function xF{sup {gamma}}{sup Z}{sub 3}, which is sensitive to the valence quark distributions down to low values of x. The data on polarised cross section asymmetries A{sup {+-}} are also combined with the ZEUS data. This leads to the first observation of parity violation in neutral current e{sup {+-}}p scattering at distances down to 10{sup -18} m. The data are well described by the Standard Model predictions. (orig.)

  2. Cr(III,Mn(II,Fe(III,Co(II,Ni(II,Cu(II and Zn(II Complexes with Diisobutyldithiocarbamato Ligand

    Directory of Open Access Journals (Sweden)

    Mohammad Tarique

    2011-01-01

    Full Text Available The synthesis of sulphur and nitrogen containing dithiocarbamato ligand derived from diisobutylamine as well as its coordination compounds with 3d series transition metals is presented. These synthesized compounds were characterized on the basis of elemental analysis, conductometric measurements and IR spectral studies. The analytical data showed the stoichiometry 1:2 and 1:3 for the compounds of the types ML2 {M=Mn(II, Co(II, Ni(II, Cu(II and Zn(II} and M'L3{M'=Cr(III and Fe(III} respectively. The conductometric measurements proved the non-electrolytic behaviour of all the compounds. The bidentate nature of dithiocarbamato moiety was confirmed on the basis of IR spectral data.

  3. Limits of reliability for the measurement of integral count

    International Nuclear Information System (INIS)

    Erbeszkorn, L.

    1979-01-01

    A method is presented for exact and approximate calculation of reliability limits of measured nuclear integral count. The formulae are applicable in measuring conditions which assure the Poisson distribution of the counts. The coefficients of the approximate formulae for 90, 95, 98 and 99 per cent reliability levels are given. The exact reliability limits for 90 per cent reliability level are calculated up to 80 integral counts. (R.J.)

  4. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

  5. Precision lifetime measurements of Ar II 4p doublet levels

    International Nuclear Information System (INIS)

    Marger, D.; Schmoranzer, H.

    1990-01-01

    The lifetimes of the Ar II 4p doublet fine-structure levels 4p 2 D 0 5/2 , 4p' 2 F 0 5/2 and 4p' 2 F 0 7/2 were measured by beam-dye laser spectroscopy. The experimental uncertainty was reduced to below 1%. (orig.)

  6. Calibration of cellulose nitrate film for measurement of time-integrated concentration of radon-222 in air

    International Nuclear Information System (INIS)

    Jha, G.; Raghavayya, M.

    1986-01-01

    Measurement of time-integrated concentration of 222 Rn in air by using solid-state nuclear track detectors (SSNTD) is finding increasing application in such diverse fields as radiation protection, exploration of radioactive minerals, prediction of earthquakes etc. While there are several types of SSNTDs sensitive to specific types of particulate radiation, films made from cellulose nitrate (CN) are found to be the best suited for quick and quantitative measurement of alpha radiation. This is because CN films are available in small thicknesses, of the order of 10-12 μm, which can be suitably evaluated by spark counting technique. This report describes the use of a sensitive thin film of CN (Kodak LR 115, Type II) for quantitative estimation of 222 Rn. The film (along with the base is exposed in a cylindrical plastic cup closed at one end with a special rubber membrane which permits discrimination of 222 Rn against 220 Rn, which is also present in air to varying extents. The calibration procedure, including etching and evaluation of track registration efficiency, are described in detail. The mean track registration efficiency works out to 41.9 per cent and the sensitivity of the system is found to be 58.92 tracks cm -2 per (Bq.litre -1 ) day. The report lists the advantages and limitations of the system for measurement of time-integrated concentration of 222 Rn in air. (author)

  7. Electronics/avionics integrity - Definition, measurement and improvement

    Science.gov (United States)

    Kolarik, W.; Rasty, J.; Chen, M.; Kim, Y.

    The authors report on the results obtained from an extensive, three-fold research project: (1) to search the open quality and reliability literature for documented information relative to electronics/avionics integrity; (2) to interpret and evaluate the literature as to significant concepts, strategies, and tools appropriate for use in electronics/avionics product and process integrity efforts; and (3) to develop a list of critical findings and recommendations that will lead to significant progress in product integrity definition, measurement, modeling, and improvements. The research consisted of examining a broad range of trade journals, scientific journals, and technical reports, as well as face-to-face discussions with reliability professionals. Ten significant recommendations have been supported by the research work.

  8. IFR fuel cycle--pyroprocess development

    International Nuclear Information System (INIS)

    Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

    1992-01-01

    The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

  9. cobalt (ii), nickel (ii)

    African Journals Online (AJOL)

    DR. AMINU

    Department of Chemistry Bayero University, P. M. B. 3011, Kano, Nigeria. E-mail: hnuhu2000@yahoo.com. ABSTRACT. The manganese (II), cobalt (II), nickel (II) and .... water and common organic solvents, but are readily soluble in acetone. The molar conductance measurement [Table 3] of the complex compounds in.

  10. Optimal distribution of integration time for intensity measurements in Stokes polarimetry.

    Science.gov (United States)

    Li, Xiaobo; Liu, Tiegen; Huang, Bingjing; Song, Zhanjie; Hu, Haofeng

    2015-10-19

    We consider the typical Stokes polarimetry system, which performs four intensity measurements to estimate a Stokes vector. We show that if the total integration time of intensity measurements is fixed, the variance of the Stokes vector estimator depends on the distribution of the integration time at four intensity measurements. Therefore, by optimizing the distribution of integration time, the variance of the Stokes vector estimator can be decreased. In this paper, we obtain the closed-form solution of the optimal distribution of integration time by employing Lagrange multiplier method. According to the theoretical analysis and real-world experiment, it is shown that the total variance of the Stokes vector estimator can be significantly decreased about 40% in the case discussed in this paper. The method proposed in this paper can effectively decrease the measurement variance and thus statistically improves the measurement accuracy of the polarimetric system.

  11. Preliminary Guideline for the High Temperature Structure Integrity Assessment Procedure Part II. High Temperature Structural Integrity Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Kim, J. B.; Lee, H. Y.; Park, C. G.; Joo, Y. S.; Koo, G. H.; Kim, S. H

    2007-02-15

    A high temperature structural integrity assessment belongs to the Part II of a whole preliminary guideline for the high temperature structure. The main contents of this guideline are the evaluation procedures of the creep-fatigue crack initiation and growth in high temperature condition, the high temperature LBB evaluation procedure, and the inelastic evaluations of the welded joints in SFR structures. The methodologies for the proper inelastic analysis of an SFR structures in high temperatures are explained and the guidelines of inelastic analysis options using ANSYS and ABAQUS are suggested. In addition, user guidelines for the developed NONSTA code are included. This guidelines need to be continuously revised to improve the applicability to the design and analysis of the SFR structures.

  12. [FeII] as a tracer supernova rate

    International Nuclear Information System (INIS)

    Rosenberg, M J F; Van der Werf, P P; Israel, F P

    2012-01-01

    Supernovae play an integral role in the feedback of processed material into the ISM of galaxies and are responsible for most of the chemical enrichment of the universe. The rate of supernovae can also reveal the star formation histories. Supernova rate is usually measured through the non-thermal radio continuum luminosity, but in this paper we establish a quantitative relationship between the [FeII] 1.26 luminosity and supernova rate in a sample of 11 near-by starburst galaxies. SINFONI data cubes are used to perform a pixel pixel analysis of this correlation. Using Bry equivalent width and luminosity as the only observational inputs into Starburst 99, the supernova rate is derived at each pixel and a map of supernova rate is created. This is then compared morphologically and quantitatively to [FeII] 1.26 luminosity map. We find a strong linear and morphological correlation between supernova rate and [FeII] 1.26 on a pixel-pixel basis. The Starburst 99 derived supernova rates are also in good agreement with the radio derived supernova rates, which further demonstrates the strength of [FeII] as a tracer of supernova rate. With the strong correlation found in this sample of galaxies, we now qualitatively use [FeII] 1.26 to derive supernova rate on either a pixel-pixel or integrated galactic basis.

  13. Integrated safety analysis of rolapitant with coadministered drugs from phase II/III trials

    DEFF Research Database (Denmark)

    Barbour, S; Smit, T.; Wang, X

    2017-01-01

    adverse events by use versus non-use of drug substrates of CYP2D6 or BCRP. Patients and methods: Patients were randomized to receive either 180 mg oral rolapitant or placebo approximately 1-2 hours before chemotherapy in combination with a 5-hydroxytryptamine type 3 RA and dexamethasone. Data...... cytochrome P450 (CYP) 3A4, but it does inhibit CYP2D6 and breast cancer resistance protein (BCRP). To analyze potential drug-drug interactions between rolapitant and concomitant medications, this integrated safety analysis of four double-blind, randomized phase II or III studies of rolapitant examined...... for treatment-emergent adverse events (TEAEs) and treatment-emergent serious adverse events (TESAEs) during cycle 1 were pooled across the four studies and summarized in the overall population and by concomitant use/non-use of CYP2D6 or BCRP substrate drugs. Results: In the integrated safety population, 828...

  14. Towards integrating environmental performance in divisional performance measurement

    Directory of Open Access Journals (Sweden)

    Collins C Ngwakwe

    2014-08-01

    Full Text Available This paper suggests an integration of environmental performance measurement (EPM into conventional divisional financial performance measures as a catalyst to enhance managers’ drive toward cleaner production and sustainable development. The approach is conceptual and normative; and using a hypothetical firm, it suggests a model to integrate environmental performance measure as an ancillary to conventional divisional financial performance measures. Vroom’s motivation theory and other literature evidence indicate that corporate goals are achievable in an environment where managers’ efforts are recognised and thus rewarded. Consequently the paper suggests that environmentally motivated managers are important to propel corporate sustainability strategy toward desired corporate environmental governance and sustainable economic development. Thus this suggested approach modestly adds to existing environmental management accounting (EMA theory and literature. It is hoped that this paper may provide an agenda for further research toward a practical application of the suggested method in a firm.

  15. Liquid metal reactor cover gas purification and analysis in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Allen, K J [Argonne National Laboratory, EBR-II Division, Idaho Falls, ID (United States); Meadows, G E; Schuck, W J [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

  16. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1987-01-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  17. Measuring cross-border regional integration with composite indicators

    DEFF Research Database (Denmark)

    Makkonen, Teemu

    2016-01-01

    Earlier quantitative studies on cross-border regional integration processes have commonly neglected science, technology and innovation (STI) indicators: even the most notable example of a composite indicator approach to measuring cross-border regional integration, i.e. the Oresund index, lacks...... a sub-category for STI. Consequently, by ignoring cross-border innovation and knowledge flows, the Oresund integration index fails to take into account one of the most important drivers of economic growth in cross-border regions. Therefore, a new composite STI indicator (sub-category) was introduced......-border regions....

  18. 41Ar-concentration measurements at the Finnish TRIGA Mark II

    International Nuclear Information System (INIS)

    Tamminen, A.

    1970-01-01

    41 Ar-concentrations in the reactor hall and in the Ar-ventilation duct of FiR-1 (IRIGA Mark II, 250 kW) have been measured during normal operation conditions. Additional measurements have been performed varying Ar-ventilation and hall ventilation flow. A γ- and background compensated thin window proportional detector and a β-Geiger detector for 41 Ar-measurements are described. The hall concentration has also been estimated from background measurements with an unshielded Ge(Li)-detector. The saturation concentration in the reactor hall with and without the hall ventilation is about 1 x 10 -7 μCi/cm 3 and 3 x 10 -7 μCi/cm 3 , respectively. (author)

  19. Building integrated care systems: a case study of Bidasoa Integrated Health Organisation

    Directory of Open Access Journals (Sweden)

    Nuria Toro Polanco

    2015-06-01

    Full Text Available Introduction: This paper analyses the implementation of integrated care policies in the Basque Country through the deployment of an Integrated Health Organisation in Bidasoa area during the period 2011–2014. Structural, functional and clinical integration policies have been employed with the aim to deliver integrated and person-centred care for patients, especially for those living with chronic conditions.Methods: This organisational case study used multiple data sources and methods in a pragmatic and reflexive manner to build a picture of the organisational development over a 4-year period. In order to measure the progress of integration three concepts have been measured: (i readiness for chronicity measured with Assessment of Readiness for Chronicity in Healthcare Organisations tool; (ii collaboration between clinicians from different care levels measured with the D'Amour Questionnaire, and (iii overall impact of integration through several indicators based on the Triple Aim Framework.Results: The measurement of organisational readiness for chronicity showed improvements in five of the six areas under evaluation. Similarly the collaboration between professionals of different care levels showed a steady improvement in each of the 10 items. Furthermore, the Triple Aim-based indicators showed a better experience of care in terms of patients’ perceptions of care coordination; a reduction in hospital utilisation, particularly for patients with complex chronic conditions; and cost-containment in terms of per capita expenditure.Conclusion: There is a significant amount of data that shows that Bidasoa Integrated Health Organisation has progressed in terms of delivering integrated care for chronic conditions with a positive impact on several Triple Aim outcomes.

  20. Building integrated care systems: a case study of Bidasoa Integrated Health Organisation

    Directory of Open Access Journals (Sweden)

    Nuria Toro Polanco

    2015-06-01

    Full Text Available Introduction: This paper analyses the implementation of integrated care policies in the Basque Country through the deployment of an Integrated Health Organisation in Bidasoa area during the period 2011–2014. Structural, functional and clinical integration policies have been employed with the aim to deliver integrated and person-centred care for patients, especially for those living with chronic conditions. Methods: This organisational case study used multiple data sources and methods in a pragmatic and reflexive manner to build a picture of the organisational development over a 4-year period. In order to measure the progress of integration three concepts have been measured: (i readiness for chronicity measured with Assessment of Readiness for Chronicity in Healthcare Organisations tool; (ii collaboration between clinicians from different care levels measured with the D'Amour Questionnaire, and (iii overall impact of integration through several indicators based on the Triple Aim Framework. Results: The measurement of organisational readiness for chronicity showed improvements in five of the six areas under evaluation. Similarly the collaboration between professionals of different care levels showed a steady improvement in each of the 10 items. Furthermore, the Triple Aim-based indicators showed a better experience of care in terms of patients’ perceptions of care coordination; a reduction in hospital utilisation, particularly for patients with complex chronic conditions; and cost-containment in terms of per capita expenditure. Conclusion: There is a significant amount of data that shows that Bidasoa Integrated Health Organisation has progressed in terms of delivering integrated care for chronic conditions with a positive impact on several Triple Aim outcomes.