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

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

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

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

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.

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

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

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

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

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

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

EBR-II Primary Tank Wash-Water Alternatives Evaluation

International Nuclear Information System (INIS)

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

2009-01-01

to six potentially viable alternatives at the end of the workshop. The six alternatives developed in the workshop were put into a scoring matrix that weighed each alternative against five important criteria that were identified during the workshop as keys to the success of the favored sodium cleanup alternative. The results of the scoring matrix exercise identified the Optimized Baseline Approach (OBA) as the favored alternative by the wash water project team. This reaffirmed the individual recommendations of the sodium cleanup specialists in a panel discussion held at the conclusion of the workshop. The OBA consists of leaving all major components of the primary system in place, and breaching certain components with a penetrating device to allow circulation of steam and reduced wash water through the components. The Optimized Baseline Approach was the alternative method recommended by the panel of sodium cleanup specialists at the workshop, and confirmed as the best alternative by the weighted scoring matrix analysis performed by the wash water project team. This alternative was chosen because it scored the highest among the common project criteria in the matrix, and most importantly it reduced the quantity of waste that would be generated for disposal. As discussed in Path Forward section of the report, actual progress towards clean closure of EBR-II will require elevating this activity to a priority project within DOE-EM. This will require transferring ownership of the EBR-II facilities from DOE-NE to DOE-EM. In addition, an acceptable de minimus level of sodium left in the primary system must be negotiated with DEQ, as it is practically impossible to remove every last gram of sodium from the system. (authors)

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

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

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)

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

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

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

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

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

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.)

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

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

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

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

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

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

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

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

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

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

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

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

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

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.

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

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

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

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

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

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.

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

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

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

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

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

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

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

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

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

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

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

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)

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

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

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

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

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)

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

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

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

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

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)

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

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

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

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

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

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.

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.

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

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).

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).

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

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

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.%

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

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

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

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

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

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

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

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

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

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

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

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.)

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)

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

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.

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

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

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

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

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

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

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

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

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.

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.)

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

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

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

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.

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

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

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

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)

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

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

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

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

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.

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

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

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

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

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

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)

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

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

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.

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

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

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.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.)

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}.

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

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

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.

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.

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

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

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

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

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.)

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

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

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

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

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

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

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.)

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.

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.

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

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

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

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.

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.)

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)

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.

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

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

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

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.

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.

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

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

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

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

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

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

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)

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

Sodium component reliability data collection at CREDO

International Nuclear Information System (INIS)

Bott, T.F.; Haas, P.M.; Manning, J.J.

1979-01-01

The Centralized Reliability Data Organization (CREDO) has been established at Oak Ridge National Laboratory (ORNL) by the Department of Energy to provide a national center for collection, evaluation and dissemination of reliability data for advanced reactors. While the system is being developed and continuous data collection at the two U.S. reactor sites (EBR-II and FFTF) is being established, data on advanced reactor components which have been in use at U.S. test loops and experimental reactors have been collected and analyzed. Engineering, operating and event data on sodium valves, pumps, flow meters, rupture discs, heat exchangers and cold traps have been collected from more than a dozen sites. The results of analyses of the data performed to date are presented

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.

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

Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

International Nuclear Information System (INIS)

Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

1988-01-01

The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

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

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.

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

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

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume II

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

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

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.

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

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.

Patch tests with fragrance mix II and its components.

Science.gov (United States)

Pónyai, Györgyi; Németh, Ilona; Altmayer, Anita; Nagy, Gabriella; Irinyi, Beatrix; Battyáni, Zita; Temesvári, Erzsébet

2012-01-01

Fragrance mix II (FM II) was initiated to detect contact hypersenstitivity (CH) to fragrances that could not have been identified previously. The aim of this multicenter study was to map the frequency of CH to FM II and its components in Hungary. Six centers participated in the survey from 2009 to 2010. A total off 565 patients (434 women and 131 men) with former skin symptoms provoked by scented products were patch tested. The tests were performed with Brial GmbH D-Greven allergens. In the environmental patch test series, FM II, FM I, Myroxylon pereirae, colophonium, wood-tar mix, propolis, and sesquiterpene lactone mix were tested as fragrance allergens. The FM II components (citral, farnesol, coumarin, citronellol, α-hexyl-cinnamaldehyde, and hydroxy-isohexyl-3-cyclohexene-carboxaldehyde [Lyral]) were also tested. Contact hypersenstitivity to any fragrances was detected in 28.8%, to FM II in 17.2% of the patients. Contact hypersenstitivity to hydroxy-isohexyl-3-cyclohexene-carboxaldehyde was observed in 7.3%, to coumarin in 5.1%, to α-hexyl-cinnamaldehyde in 3.5%, to citral in 3.4%, to farnesol in 2.5%, and to citronellol in 1.2%. Of the FM II-positive cases, 48.4% showed isolated CH reaction. The frequency of CH to FM II is 17.2% in the tested, selected Hungarian population. The CH to FM II and its components could not have been revealed without the present test materials.

Application of damage functions to CTR component fluence limit predictions

International Nuclear Information System (INIS)

Simons, R.L.; Doran, D.G.

1975-01-01

Material behavior observed under irradiation conditions in test reactors is not always directly applicable to the design of reactor components such as CTR first wall because of differences in the damage effectiveness of test reactor and service spectra. The interpolation and, under some conditions, extrapolation of material property change data from test conditions to different neutron spectra in service conditions can be accomplished using semi-empirical damage functions. The derivation and application of damage functions to CTR conditions is reviewed. Since limited amounts of data are available for applications to CTR design spectra, considerable attention is placed on the effectiveness of various available and proposed neutron sources in determining a damage function and subsequent fluence limit prediction. Neutron sources included in this study were EBR-II, HIFR, LAMPF (Be and Cu targets), high energy deuterons incident on Be (D-Be), and 14 MeV neutrons (D-T)

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

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

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

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

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

A display to support knowledge based behavior

International Nuclear Information System (INIS)

Lindsay, R.W.

1990-01-01

A computerized display has been created for the Experimental Breeder Reactor II (EBR-II) that incorporates information from plant sensors in a thermodynamic model display. The display is designed to provide an operator with an overall view of the plant process as a heat engine. The thermodynamics of the plant are depicted through the use of ionic figures, animated by plant signals, that are related to the major plant components and systems such as the reactor, intermediate heat exchanger, secondary system, evaporators, superheaters, steam system, steam drum, and turbine-generator. This display supports knowledge based reasoning for the operator as well as providing the traditional rule and skill based behavior, and includes side benefits such a inherent signal validation

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

Combining two major ATLAS inner detector components

CERN Multimedia

Maximilien Brice

2006-01-01

The semiconductor tracker is inserted into the transition radiation tracker for the ATLAS experiment at the LHC. These make up two of the three major components of the inner detector. They will work together to measure the trajectories produced in the proton-proton collisions at the centre of the detector when the LHC is switched on in 2008.

Residual life assessment of major LWR components: NPAR approach and results

International Nuclear Information System (INIS)

Shah, V.N.; Weidenhamer, G.H.; Vora, J.P.

1991-01-01

The nuclear plant aging research (NPAR) program is systematically addressing the technical issues associated with understanding and managing aging of major LWR components. Twenty-one major components have been identified and prioritized according to their relevance to plant safety. Qualitative aging assessment has identified pertinent design features, materials, stressors, environments, aging mechanisms. and failure modes for each of the components. Emerging inspection, surveillance, and monitoring methods to characterize aging damage and mitigation methods to reduce the damage are currently being assessed. The results of all these assessments are used to develop life-assessment procedures for the components and are included in appropriate documents supporting the regulatory requirements for license renewal. (author)

Aging assessment and mitigation for major LWR [light water reactor] components

International Nuclear Information System (INIS)

Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

1989-01-01

This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

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

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

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

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

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

A Study on Major Components of Bee Venom Using Electrophoresis

Directory of Open Access Journals (Sweden)

Lee, Jin-Seon

2000-12-01

Full Text Available This study was designed to study on major components of various Bee Venom(Bee Venom by electrical stimulation in Korea; K-BV I, Bee Venom by Microwave stimulation in Korea; K -BV II, 0.5rng/ml, Fu Yu Pharmaceutical Factory, China; C-BV, 1mg /ml, Monmouth Pain Institute, Inc., U.S.A.; A-BV using Electrophoresis. The results were summarized as follows: 1. In 1:4000 Bee Venom solution rate, the band was not displayed distinctly usmg Electrophoresis. But in 1: 1000, the band showed clearly. 2. The results of Electrophoresis at solution rate 1:1000, K-BV I and K-BVII showed similar band. 3. The molecular weight of Phospholipase A2 was known as 19,000 but its band was seen at 17,000 in Electrophoresis. 4. Protein concentration of Bee Venom by Lowry method was different at solution rate 1:4000 ; C-BV was 250μg/ml, K-BV I was 190μg/ml, K-BV Ⅱ was 160μg/ml and C-BV was 45μg/ml. 5. Electrophoresis method was unuseful for analysis of Bee Venom when solution rate is above 1:4000 but Protein concentration of Bee Venom by Lowry method was possible. These data from the study can be applied to establish the standard measurement of Bee Venom and prevent pure bee venom from mixing of another components. I think it is desirable to study more about safety of Bee Venom as time goes by.

COMDES-II: A Component-Based Framework for Generative Development of Distributed Real-Time Control Systems

DEFF Research Database (Denmark)

Ke, Xu; Sierszecki, Krzysztof; Angelov, Christo K.

2007-01-01

The paper presents a generative development methodology and component models of COMDES-II, a component-based software framework for distributed embedded control systems with real-time constraints. The adopted methodology allows for rapid modeling and validation of control software at a higher lev...... methodology for COMDES-II from a general perspective, describes the component models in details and demonstrates their application through a DC-Motor control system case study.......The paper presents a generative development methodology and component models of COMDES-II, a component-based software framework for distributed embedded control systems with real-time constraints. The adopted methodology allows for rapid modeling and validation of control software at a higher level...

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

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

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

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

Effect of Demand-Driven advisory services components of Fadama II ...

African Journals Online (AJOL)

Effect of Demand-Driven advisory services components of Fadama II on fish farmers' level of ... Journal of Agriculture, Forestry and the Social Sciences ... In conclusion, provision of subsidized advisory services to fish farmers will help to boost ...

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

41 CFR 102-85.40 - What are the major components of the pricing policy?

Science.gov (United States)

2010-07-01

... components of the pricing policy? 102-85.40 Section 102-85.40 Public Contracts and Property Management...-PRICING POLICY FOR OCCUPANCY IN GSA SPACE Pricing Policy-General § 102-85.40 What are the major components of the pricing policy? The major components of the pricing policy are: (a) An OA between a customer...

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.

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.

An Instrument for Inspecting Aspheric Optical Surfaces and Components, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — This is a Phase II SBIR proposal to develop an extremely versatile optical inspection tool for aspheric optical components and optics that are not easily inspected...

Common and distinct genetic properties of ESCRT-II components in Drosophila.

Directory of Open Access Journals (Sweden)

Hans-Martin Herz

Full Text Available BACKGROUND: Genetic studies in yeast have identified class E vps genes that form the ESCRT complexes required for protein sorting at the early endosome. In Drosophila, mutations of the ESCRT-II component vps25 cause endosomal defects leading to accumulation of Notch protein and increased Notch pathway activity. These endosomal and signaling defects are thought to account for several phenotypes. Depending on the developmental context, two different types of overgrowth can be detected. Tissue predominantly mutant for vps25 displays neoplastic tumor characteristics. In contrast, vps25 mutant clones in a wild-type background trigger hyperplastic overgrowth in a non-autonomous manner. In addition, vps25 mutant clones also promote apoptotic resistance in a non-autonomous manner. PRINCIPAL FINDINGS: Here, we genetically characterize the remaining ESCRT-II components vps22 and vps36. Like vps25, mutants of vps22 and vps36 display endosomal defects, accumulate Notch protein and--when the tissue is predominantly mutant--show neoplastic tumor characteristics. However, despite these common phenotypes, they have distinct non-autonomous phenotypes. While vps22 mutations cause strong non-autonomous overgrowth, they do not affect apoptotic resistance. In contrast, vps36 mutations increase apoptotic resistance, but have little effect on non-autonomous proliferation. Further characterization reveals that although all ESCRT-II mutants accumulate Notch protein, only vps22 and vps25 mutations trigger Notch activity. CONCLUSIONS/SIGNIFICANCE: The ESCRT-II components vps22, vps25 and vps36 display common and distinct genetic properties. Our data redefine the role of Notch for hyperplastic and neoplastic overgrowth in these mutants. While Notch is required for hyperplastic growth, it appears to be dispensable for neoplastic transformation.

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

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

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

A major protein component of the Bacillus subtilis biofilm matrix.

Science.gov (United States)

Branda, Steven S; Chu, Frances; Kearns, Daniel B; Losick, Richard; Kolter, Roberto

2006-02-01

Microbes construct structurally complex multicellular communities (biofilms) through production of an extracellular matrix. Here we present evidence from scanning electron microscopy showing that a wild strain of the Gram positive bacterium Bacillus subtilis builds such a matrix. Genetic, biochemical and cytological evidence indicates that the matrix is composed predominantly of a protein component, TasA, and an exopolysaccharide component. The absence of TasA or the exopolysaccharide resulted in a residual matrix, while the absence of both components led to complete failure to form complex multicellular communities. Extracellular complementation experiments revealed that a functional matrix can be assembled even when TasA and the exopolysaccharide are produced by different cells, reinforcing the view that the components contribute to matrix formation in an extracellular manner. Having defined the major components of the biofilm matrix and the control of their synthesis by the global regulator SinR, we present a working model for how B. subtilis switches between nomadic and sedentary lifestyles.

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)

Data base formation for important components of reactor TRIGA MARK II

International Nuclear Information System (INIS)

Jordan, R.; Mavko, B.; Kozuh, M.

1992-01-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [sl

Failed fuel identification techniques for liquid-metal cooled reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

1995-01-01

The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

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.

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

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

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

Ultrasonic inspection of heavy section steel components: the PISC II final report

International Nuclear Information System (INIS)

Nichols, R.W.; Crutzen, S.

1988-01-01

This Symposium represented the end of the PISC (Programme for Inspection of Steel Components), II Round Robin Test Project, and the book is the final report. The contents are divided into three parts: part 1 contains contributions from the PISC Management Group, part II contains individual contributions, part III contains views of Licensing Authorities. All the twenty three papers presented in the three parts are selected for INIS and indexed separately. (author)

Tech Area II: A history

Energy Technology Data Exchange (ETDEWEB)

Ullrich, R. [Ktech Corp., Albuquerque, NM (United States)

1998-07-01

This report documents the history of the major buildings in Sandia National Laboratories` Technical Area II. It was prepared in support of the Department of Energy`s compliance with Section 106 of the National Historic Preservation Act. Technical Area II was designed and constructed in 1948 specifically for the final assembly of the non-nuclear components of nuclear weapons, and was the primary site conducting such assembly until 1952. Both the architecture and location of the oldest buildings in the area reflect their original purpose. Assembly activities continued in Area II from 1952 to 1957, but the major responsibility for this work shifted to other sites in the Atomic Energy Commission`s integrated contractor complex. Gradually, additional buildings were constructed and the original buildings were modified. After 1960, the Area`s primary purpose was the research and testing of high-explosive components for nuclear weapons. In 1994, Sandia constructed new facilities for work on high-explosive components outside of the original Area II diamond-shaped parcel. Most of the buildings in the area are vacant and Sandia has no plans to use them. They are proposed for decontamination and demolition as funding becomes available.

The TFIID components human TAF(II)140 and Drosophila BIP2 (TAF(II)155) are novel metazoan homologues of yeast TAF(II)47 containing a histone fold and a PHD finger.

Science.gov (United States)

Gangloff, Y G; Pointud, J C; Thuault, S; Carré, L; Romier, C; Muratoglu, S; Brand, M; Tora, L; Couderc, J L; Davidson, I

2001-08-01

The RNA polymerase II transcription factor TFIID comprises the TATA binding protein (TBP) and a set of TBP-associated factors (TAF(II)s). TFIID has been extensively characterized for yeast, Drosophila, and humans, demonstrating a high degree of conservation of both the amino acid sequences of the constituent TAF(II)s and overall molecular organization. In recent years, it has been assumed that all the metazoan TAF(II)s have been identified, yet no metazoan homologues of yeast TAF(II)47 (yTAF(II)47) and yTAF(II)65 are known. Both of these yTAF(II)s contain a histone fold domain (HFD) which selectively heterodimerizes with that of yTAF(II)25. We have cloned a novel mouse protein, TAF(II)140, containing an HFD and a plant homeodomain (PHD) finger, which we demonstrated by immunoprecipitation to be a mammalian TFIID component. TAF(II)140 shows extensive sequence similarity to Drosophila BIP2 (dBIP2) (dTAF(II)155), which we also show to be a component of Drosophila TFIID. These proteins are metazoan homologues of yTAF(II)47 as their HFDs selectively heterodimerize with dTAF(II)24 and human TAF(II)30, metazoan homologues of yTAF(II)25. We further show that yTAF(II)65 shares two domains with the Drosophila Prodos protein, a recently described potential dTAF(II). These conserved domains are critical for yTAF(II)65 function in vivo. Our results therefore identify metazoan homologues of yTAF(II)47 and yTAF(II)65.

Chemical evolution of two-component galaxies. II

International Nuclear Information System (INIS)

Caimmi, R.

1978-01-01

In order to confirm and refine the results obtained in a previous paper the chemical evolution of two-component (spheroid + disk) galaxies is derived rejecting the instantaneous recycling approximation, by means of numerical computations, accounting for (i) the collapse phase of the gas, assumed to be uniform in density and composition, and (ii) a birth-rate stellar function. Computations are performed relatively to the solar neighbourhood and to model galaxies which closely resemble the real morphological sequence: in both cases, numerical results are compared with analytical ones. The numerical models of this paper constitute a first-order approximation, while higher order approximations could be made by rejecting the hypothesis of uniform density and composition, and making use of detailed dynamical models. (Auth.)

Evaluation of relationship between metacognition components and dysfunctional attitudes in outpatients with bipolar mood disorder II

Directory of Open Access Journals (Sweden)

H Kazemi

2012-05-01

Full Text Available Introduction: The aim of this study was to evaluate the relationship between metacognitive components and dysfunctional attitudes in outpatients with bipolar mood disorder II. Methods: Thirty-six young adult outpatients with current diagnoses of BMD II(20 females and 16 males were recruited from Esfahan Counseling Center. Diagnoses were based on the Structured Clinical Interview for DSM-IV Axis I Disorder-Clinical Version(SCID-CV. A battery of questionnaires including Metacognition Questionnaire and Dysfunctional Attitude Scale(DAS were self-assessed by patients before medical therapy. Results: Pearson's correlation analysis showed that the components of metacognition and its relationship with dysfunctional attitudes is positive and significant(r= 28/0, p<0/05. Multiple regression analysis showed that two of the metacognitive components emerged as potentially useful in prediction of dysfunctional attitudes(negative beliefs about uncontrollability, danger and thoughts control. Also, results indicated that those two components have a significant positive relationship with vulnerability, perfectionism and effectivenessR²= 0/29, 0/35; p<0/05. Components of positive beliefs about worry and beliefs about cognitive self-consciousness related to cognitive confidence in predicting the criterion variable and its components showed no significant contribution. Conclusion: Study findings suggest that DSM-IV BMD II outpatients with metacognitive distortions have shown higher levels of dysfunctional attitudes.

Aging and life extension of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; MacDonald, P.E.

1993-01-01

An understanding of the aging degradation of the major pressurized and boiling water reactor structures and components is given. The design and fabrication of each structure or component is briefly described followed by information on the associated stressors. Interactions between the design, materials and various stressors that cause aging degradation are reviewed. In many cases, aging degradation problems have occurred, and the plant experience to date is analyzed. The discussion summarize the available aging-related information and are supported with extensive references, including references to US Nuclear Regulatory Commission (USNRC) documents, Electric Power Research Institute reports, US and international conference proceedings and other publications

The major results of the PISC II Round Robin test

International Nuclear Information System (INIS)

Crutzen, S.

1987-01-01

The objectives of the PISC-II programme were: - To evaluate the effectiveness of NDT techniques either in use or being developed for the inspection of reactor pressure vessel components, with regard mainly to detecting and charaterising service-induced flaws. - To identify techniques for acceptance tests, pre-service inspection (PSI) and in-service inspection (ISI) which could be generally accepted, and - to bring the conclusions of the programme to the attention of the Code, Standard and Regulatory Bodies concerned with ISI. (orig./HP)

Data base formation for important components of reactor TRIGA MARK II

Energy Technology Data Exchange (ETDEWEB)

Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

1992-07-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

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

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.

Evidence for multiple major histocompatibility class II X-box binding proteins.

OpenAIRE

Celada, A; Maki, R

1989-01-01

The X box is a loosely conserved DNA sequence that is located upstream of all major histocompatibility class II genes and is one of the cis-acting regulatory elements. Despite the similarity between all X-box sequences, each promoter-proximal X box in the mouse appears to bind a separate nuclear factor.

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

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

Aging management of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; Sinha, U.P.; Ware, A.G.

1992-01-01

Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

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

Evolution of major histocompatibility complex class I and class II genes in the brown bear

Directory of Open Access Journals (Sweden)

Kuduk Katarzyna

2012-10-01

Full Text Available Abstract Background Major histocompatibility complex (MHC proteins constitute an essential component of the vertebrate immune response, and are coded by the most polymorphic of the vertebrate genes. Here, we investigated sequence variation and evolution of MHC class I and class II DRB, DQA and DQB genes in the brown bear Ursus arctos to characterise the level of polymorphism, estimate the strength of positive selection acting on them, and assess the extent of gene orthology and trans-species polymorphism in Ursidae. Results We found 37 MHC class I, 16 MHC class II DRB, four DQB and two DQA alleles. We confirmed the expression of several loci: three MHC class I, two DRB, two DQB and one DQA. MHC class I also contained two clusters of non-expressed sequences. MHC class I and DRB allele frequencies differed between northern and southern populations of the Scandinavian brown bear. The rate of nonsynonymous substitutions (dN exceeded the rate of synonymous substitutions (dS at putative antigen binding sites of DRB and DQB loci and, marginally significantly, at MHC class I loci. Models of codon evolution supported positive selection at DRB and MHC class I loci. Both MHC class I and MHC class II sequences showed orthology to gene clusters found in the giant panda Ailuropoda melanoleuca. Conclusions Historical positive selection has acted on MHC class I, class II DRB and DQB, but not on the DQA locus. The signal of historical positive selection on the DRB locus was particularly strong, which may be a general feature of caniforms. The presence of MHC class I pseudogenes may indicate faster gene turnover in this class through the birth-and-death process. South–north population structure at MHC loci probably reflects origin of the populations from separate glacial refugia.

Evolution of major histocompatibility complex class I and class II genes in the brown bear.

Science.gov (United States)

Kuduk, Katarzyna; Babik, Wiesław; Bojarska, Katarzyna; Sliwińska, Ewa B; Kindberg, Jonas; Taberlet, Pierre; Swenson, Jon E; Radwan, Jacek

2012-10-02

Major histocompatibility complex (MHC) proteins constitute an essential component of the vertebrate immune response, and are coded by the most polymorphic of the vertebrate genes. Here, we investigated sequence variation and evolution of MHC class I and class II DRB, DQA and DQB genes in the brown bear Ursus arctos to characterise the level of polymorphism, estimate the strength of positive selection acting on them, and assess the extent of gene orthology and trans-species polymorphism in Ursidae. We found 37 MHC class I, 16 MHC class II DRB, four DQB and two DQA alleles. We confirmed the expression of several loci: three MHC class I, two DRB, two DQB and one DQA. MHC class I also contained two clusters of non-expressed sequences. MHC class I and DRB allele frequencies differed between northern and southern populations of the Scandinavian brown bear. The rate of nonsynonymous substitutions (dN) exceeded the rate of synonymous substitutions (dS) at putative antigen binding sites of DRB and DQB loci and, marginally significantly, at MHC class I loci. Models of codon evolution supported positive selection at DRB and MHC class I loci. Both MHC class I and MHC class II sequences showed orthology to gene clusters found in the giant panda Ailuropoda melanoleuca. Historical positive selection has acted on MHC class I, class II DRB and DQB, but not on the DQA locus. The signal of historical positive selection on the DRB locus was particularly strong, which may be a general feature of caniforms. The presence of MHC class I pseudogenes may indicate faster gene turnover in this class through the birth-and-death process. South-north population structure at MHC loci probably reflects origin of the populations from separate glacial refugia.

Evolution of major histocompatibility complex class I and class II genes in the brown bear

Science.gov (United States)

2012-01-01

Background Major histocompatibility complex (MHC) proteins constitute an essential component of the vertebrate immune response, and are coded by the most polymorphic of the vertebrate genes. Here, we investigated sequence variation and evolution of MHC class I and class II DRB, DQA and DQB genes in the brown bear Ursus arctos to characterise the level of polymorphism, estimate the strength of positive selection acting on them, and assess the extent of gene orthology and trans-species polymorphism in Ursidae. Results We found 37 MHC class I, 16 MHC class II DRB, four DQB and two DQA alleles. We confirmed the expression of several loci: three MHC class I, two DRB, two DQB and one DQA. MHC class I also contained two clusters of non-expressed sequences. MHC class I and DRB allele frequencies differed between northern and southern populations of the Scandinavian brown bear. The rate of nonsynonymous substitutions (dN) exceeded the rate of synonymous substitutions (dS) at putative antigen binding sites of DRB and DQB loci and, marginally significantly, at MHC class I loci. Models of codon evolution supported positive selection at DRB and MHC class I loci. Both MHC class I and MHC class II sequences showed orthology to gene clusters found in the giant panda Ailuropoda melanoleuca. Conclusions Historical positive selection has acted on MHC class I, class II DRB and DQB, but not on the DQA locus. The signal of historical positive selection on the DRB locus was particularly strong, which may be a general feature of caniforms. The presence of MHC class I pseudogenes may indicate faster gene turnover in this class through the birth-and-death process. South–north population structure at MHC loci probably reflects origin of the populations from separate glacial refugia. PMID:23031405

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

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.)

Monitoring aquatic environment pollution : a major component of environment management system part-II

International Nuclear Information System (INIS)

Khan, I.H.; Khan, M.H.; Sheikh, I.M.

1999-01-01

The quality of water suitable for simple disinfection and filtration is presented in this article. Aquatic monitoring requires sampling frequencies along the alignment of surface water bodies. It is necessary to control the industrial effluents discharge in to river and sewers. Unlike wastes from entirely domestic sources, however industrial effluents may contain a very large variety of unnatural components which necessitates greater considerations in setting suitable discharge limits and, perhaps closer surveillance to ensure that standards are met. Several suggestion and example of different types of effluent have been described. All the examples given are sufficiently convincing that Pakistan can learn a great deal from international experiences in environmental pollution to avoid catastrophes. (A.B.)

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.

Aging management of major LWR components with nondestructive evaluation

International Nuclear Information System (INIS)

Shah, V.N.; MacDonald, P.E.; Akers, D.W.; Sellers, C.; Murty, K.L.; Miraglia, P.Q.; Mathew, M.D.; Haggag, F.M.

1997-01-01

Nondestructive evaluation of material damage can contribute to continued safe, reliable, and economical operation of nuclear power plants through their current and renewed license period. The aging mechanisms active in the major light water reactor components are radiation embrittlement, thermal aging, stress corrosion cracking, flow-accelerated corrosion, and fatigue, which reduce fracture toughness, structural strength, or fatigue resistance of the components and challenge structural integrity of the pressure boundary. This paper reviews four nondestructive evaluation methods with the potential for in situ assessment of damage caused by these mechanisms: stress-strain microprobe for determining mechanical properties of reactor pressure vessel and cast stainless materials, magnetic methods for estimating thermal aging damage in cast stainless steel, positron annihilation measurements for estimating early fatigue damage in reactor coolant system piping, and ultrasonic guided wave technique for detecting cracks and wall thinning in tubes and pipes and corrosion damage to embedded portion of metal containments

Matrix effects in PIXE evaluation of the major components in thick homogeneous samples

International Nuclear Information System (INIS)

Oystaeyen, B. van; Demortier, G.

1983-01-01

Matrix effects on the major component X-ray yields in PIXE measurements are studied in general terms with the Au-Cu-Ag matrix as practical example. Postulating firstly that all the major components may be simultaneously detected through one well-isolated peak for each of them, and secondly that a known reference material is available which contains all the elements of the unknown sample, we propose a direct method to extract the true concentrations of the latter taking into accout the matrix effects. The geometrical parameters describing the target location with respect to ion beam direction and detector position are also studied and a first attempt is made to include them in the data treatment. (orig.)

Brucella abortus Inhibits Major Histocompatibility Complex Class II Expression and Antigen Processing through Interleukin-6 Secretion via Toll-Like Receptor 2▿

Science.gov (United States)

Barrionuevo, Paula; Cassataro, Juliana; Delpino, M. Victoria; Zwerdling, Astrid; Pasquevich, Karina A.; Samartino, Clara García; Wallach, Jorge C.; Fossati, Carlos A.; Giambartolomei, Guillermo H.

2008-01-01

The strategies that allow Brucella abortus to survive inside macrophages for prolonged periods and to avoid the immunological surveillance of major histocompatibility complex class II (MHC-II)-restricted gamma interferon (IFN-γ)-producing CD4+ T lymphocytes are poorly understood. We report here that infection of THP-1 cells with B. abortus inhibited expression of MHC-II molecules and antigen (Ag) processing. Heat-killed B. abortus (HKBA) also induced both these phenomena, indicating the independence of bacterial viability and involvement of a structural component of the bacterium. Accordingly, outer membrane protein 19 (Omp19), a prototypical B. abortus lipoprotein, inhibited both MHC-II expression and Ag processing to the same extent as HKBA. Moreover, a synthetic lipohexapeptide that mimics the structure of the protein lipid moiety also inhibited MHC-II expression, indicating that any Brucella lipoprotein could down-modulate MHC-II expression and Ag processing. Inhibition of MHC-II expression and Ag processing by either HKBA or lipidated Omp19 (L-Omp19) depended on Toll-like receptor 2 and was mediated by interleukin-6. HKBA or L-Omp19 also inhibited MHC-II expression and Ag processing of human monocytes. In addition, exposure to the synthetic lipohexapeptide inhibited Ag-specific T-cell proliferation and IFN-γ production of peripheral blood mononuclear cells from Brucella-infected patients. Together, these results indicate that there is a mechanism by which B. abortus may prevent recognition by T cells to evade host immunity and establish a chronic infection. PMID:17984211

Zone-refining of beryllium(II) acetylacetonate

International Nuclear Information System (INIS)

Yoshida, Isao; Kobayashi, Hiroshi; Ueno, Keihei

1975-01-01

Zone melting was applied to bis (acetylacetonato) beryllium(II), Be(AA) 2 , to remove trace metal constiutents. The effective distribution coefficient, k, of a minor component measured in a binary mixture with the majority of Be(AA) 2 , fell in the range of 0.4 to 0.8, while the value obtained in a multiple component mixture fell in the range of 0.8 to 0.9 except Zn(II)- and Ni(II) chelates. Cr(AA) 3 was concentrated upward in the direction of the zone travelling, and similar results were obtained with Cu(II)-, Co(III)- Fe(III)- and Al(III) acetylacetonates, while Zn(II)-, Ni(II)- and Mn(III) chelates were fractionated in the both ends of the column. On the other hand, when the zone was moved downward, the minor components were concentrated in the lower end of a column. Efficiency in zone refining was compared to each other between the upward- and downward zone travelling modes on the crude beryllium acetylacetonates obtained from a commercial reagent of beryllium nitrate, and the latter mode was found more effective in the concentration of trace components. Discussion was given on possible mechanisms resulting in the difference in the refining efficiency. (author)

Endothelial monocyte activating polypeptide-II modulates endothelial cell responses by degrading hypoxia-inducible factor-1alpha through interaction with PSMA7, a component of the proteasome

International Nuclear Information System (INIS)

Tandle, Anita T.; Calvani, Maura; Uranchimeg, Badarch; Zahavi, David; Melillo, Giovanni; Libutti, Steven K.

2009-01-01

The majority of human tumors are angiogenesis dependent. Understanding the specific mechanisms that contribute to angiogenesis may offer the best approach to develop therapies to inhibit angiogenesis in cancer. Endothelial monocyte activating polypeptide-II (EMAP-II) is an anti-angiogenic cytokine with potent effects on endothelial cells (ECs). It inhibits EC proliferation and cord formation, and it suppresses primary and metastatic tumor growth in-vivo. However, very little is known about the molecular mechanisms behind the anti-angiogenic activity of EMAP-II. In the present study, we explored the molecular mechanism behind the anti-angiogenic activity exerted by this protein on ECs. Our results demonstrate that EMAP-II binds to the cell surface α5β1 integrin receptor. The cell surface binding of EMAP-II results in its internalization into the cytoplasmic compartment where it interacts with its cytoplasmic partner PSMA7, a component of the proteasome degradation pathway. This interaction increases hypoxia-inducible factor 1-alpha (HIF-1α) degradation under hypoxic conditions. The degradation results in the inhibition of HIF-1α mediated transcriptional activity as well as HIF-1α mediated angiogenic sprouting of ECs. HIF-1α plays a critical role in angiogenesis by activating a variety of angiogenic growth factors. Our results suggest that one of the major anti-angiogenic functions of EMAP-II is exerted through its inhibition of the HIF-1α activities.

Endothelial monocyte activating polypeptide-II modulates endothelial cell responses by degrading hypoxia-inducible factor-1alpha through interaction with PSMA7, a component of the proteasome

Energy Technology Data Exchange (ETDEWEB)

Tandle, Anita T. [Tumor Angiogenesis Section, Surgery Branch, Center for Cancer Research, National Cancer Institute, Bethesda, Maryland 20892 (United States); Calvani, Maura; Uranchimeg, Badarch [DTP-Tumor Hypoxia Laboratory, SAIC Frederick, Inc., National Cancer Institute, Frederick, Maryland 21702 (United States); Zahavi, David [Tumor Angiogenesis Section, Surgery Branch, Center for Cancer Research, National Cancer Institute, Bethesda, Maryland 20892 (United States); Melillo, Giovanni [DTP-Tumor Hypoxia Laboratory, SAIC Frederick, Inc., National Cancer Institute, Frederick, Maryland 21702 (United States); Libutti, Steven K., E-mail: slibutti@montefiore.org [Department of Surgery, Montefiore-Einstein Center for Cancer Care, Albert Einstein College of Medicine, Greene Medical Arts Pavilion, 4th Floor 3400, Bainbridge Avenue, Bronx, New York 10467 (United States)

2009-07-01

The majority of human tumors are angiogenesis dependent. Understanding the specific mechanisms that contribute to angiogenesis may offer the best approach to develop therapies to inhibit angiogenesis in cancer. Endothelial monocyte activating polypeptide-II (EMAP-II) is an anti-angiogenic cytokine with potent effects on endothelial cells (ECs). It inhibits EC proliferation and cord formation, and it suppresses primary and metastatic tumor growth in-vivo. However, very little is known about the molecular mechanisms behind the anti-angiogenic activity of EMAP-II. In the present study, we explored the molecular mechanism behind the anti-angiogenic activity exerted by this protein on ECs. Our results demonstrate that EMAP-II binds to the cell surface {alpha}5{beta}1 integrin receptor. The cell surface binding of EMAP-II results in its internalization into the cytoplasmic compartment where it interacts with its cytoplasmic partner PSMA7, a component of the proteasome degradation pathway. This interaction increases hypoxia-inducible factor 1-alpha (HIF-1{alpha}) degradation under hypoxic conditions. The degradation results in the inhibition of HIF-1{alpha} mediated transcriptional activity as well as HIF-1{alpha} mediated angiogenic sprouting of ECs. HIF-1{alpha} plays a critical role in angiogenesis by activating a variety of angiogenic growth factors. Our results suggest that one of the major anti-angiogenic functions of EMAP-II is exerted through its inhibition of the HIF-1{alpha} activities.

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

A CK2 site is reversibly phosphorylated in the photosystem II subunit CP29

NARCIS (Netherlands)

Testi, Maria Grazia; Croce, Roberta; Polverino-De Laureto, Patrizia; Bassi, Roberto

1996-01-01

Protein phosphorylation is a major mechanism in the regulation of protein function. In chloroplast thylakoids several photosystem II subunits, including the major antenna light-harvesting complex II and several core complex components, are reversibly phosphorylated depending on the redox state of

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

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

Overview of US fast-neutron facilities and testing capabilities

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Jackson, R.J.

1982-01-01

Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

Association between single nucleotide polymorphism of apoVLDL-II ...

African Journals Online (AJOL)

ONOS

2010-07-05

Jul 5, 2010 ... Very low density apolipoprotein-II (apoVLDL-II) is a major polypeptide component of avian VLDL. The function of apoVLDL-II is the transport of neutral lipids (triacylglycerol) in the form of VLDL in the plasma. The apoVLDLII gene is dormant in embryos, chicks and roosters but can be activated by estrogen.

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

Chemical composition and antiprotozoal activities of Colombian Lippia spp essential oils and their major components

Directory of Open Access Journals (Sweden)

Patricia Escobar

2010-03-01

Full Text Available The chemical composition and biological activities of 19 essential oils and seven of their major components were tested against free and intracellular forms of Leishmania chagasi and Trypanosoma cruzi parasites as well as Vero and THP-1 mammalian cell lines. The essential oils were obtained from different species of Lippia, a widely distributed genus of Colombian plants. They were extracted by microwave radiation-assisted hydro-distillation and characterised by GC-FID and GC-MS. The major components were geranial, neral, limonene, nerol, carvacrol, p-cymene, γ-terpinene, carvone and thymol. The essential oil of Lippia alba exhibited the highest activity against T. cruzi epimastigotes and intracellular amastigotes with an IC50 of 5.5 μg/mL and 12.2 μg/mL, respectively. The essential oil of Lippia origanoides had an IC50 of 4.4 μg/mL in L. chagasi promastigotes and exhibited no toxicity in mammalian cells. Thymol (IC50 3.2 ± 0.4 μg/mL and S-carvone (IC50 6.1 ± 2.2 μg/mL, two of the major components of the active essential oils, were active on intracellular amastigotes of T. cruziinfected Vero cells, with a selective index greater than 10. None of the essential oils or major components tested in this study was active on amastigotes of L. chagasi infected THP-1 cells.

Advanced 3D Human Simulation Components with Thermal/Haptic Feedback and Tissue Deformation, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — In integrating the following three significant components for its research/research and development (R/R&D) effort, the power of this candidate Phase II project...

Reactions of green and black teas with Cu(II).

Science.gov (United States)

Goodman, B A; Ferreira Severino, J; Pirker, K F

2012-04-01

Electron paramagnetic resonance (EPR) measurements of the products of reactions between Cu(II) and samples of green and black teas showed spectral components from at least six different Cu(II) complexes with both tea types. Several of these complexes were common to both teas in spite of major differences in their polyphenol compositions. The pH range observed for complex formation, and the total signal intensity in the pH range 4-8, were greatly different from those for the reactions of Cu(II) with (-)-epigallocatechin gallate and gallic acid, the main polyphenols responsible for the free radical signals observed during oxidation of these beverages. Components with spectral parameters similar to those of Cu(II) complexes with theanine, the major amino acid in tea, may contribute to two of the spectra recorded under acidic conditions. However, the initial complexes formed at the lowest pH values investigated are still unidentified. EPR spectra with parameters consistent with Cu(II) polyphenol complexes were only observed under alkaline conditions, thus suggesting that components of tea other than polyphenols might be more important in reactions with copper, and possibly other transition metals, in solutions under physiological conditions. This journal is © The Royal Society of Chemistry 2012

Fluorescence lifetime components reveal kinetic intermediate states upon equilibrium denaturation of carbonic anhydrase II.

Science.gov (United States)

Nemtseva, Elena V; Lashchuk, Olesya O; Gerasimova, Marina A; Melnik, Tatiana N; Nagibina, Galina S; Melnik, Bogdan S

2017-12-21

In most cases, intermediate states of multistage folding proteins are not 'visible' under equilibrium conditions but are revealed in kinetic experiments. Time-resolved fluorescence spectroscopy was used in equilibrium denaturation studies. The technique allows for detecting changes in the conformation and environment of tryptophan residues in different structural elements of carbonic anhydrase II which in its turn has made it possible to study the intermediate states of carbonic anhydrase II under equilibrium conditions. The results of equilibrium and kinetic experiments using wild-type bovine carbonic anhydrase II and its mutant form with the substitution of leucine for alanine at position 139 (L139A) were compared. The obtained lifetime components of intrinsic tryptophan fluorescence allowed for revealing that, the same as in kinetic experiments, under equilibrium conditions the unfolding of carbonic anhydrase II ensues through formation of intermediate states.

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

Effects of Icariside II on Corpus Cavernosum and Major Pelvic Ganglion Neuropathy in Streptozotocin-Induced Diabetic Rats

Directory of Open Access Journals (Sweden)

Guang-Yi Bai

2014-12-01

Full Text Available Diabetic erectile dysfunction is associated with penile dorsal nerve bundle neuropathy in the corpus cavernosum and the mechanism is not well understood. We investigated the neuropathy changes in the corpus cavernosum of rats with streptozotocin-induced diabetes and the effects of Icariside II (ICA II on improving neuropathy. Thirty-six 8-week-old Sprague-Dawley rats were randomly distributed into normal control group, diabetic group and ICA-II treated group. Diabetes was induced by a one-time intraperitoneal injection of streptozotocin (60 mg/kg. Three days later, the diabetic rats were randomly divided into 2 groups including a saline treated placebo group and an ICA II-treated group (5 mg/kg/day, by intragastric administration daily. Twelve weeks later, erectile function was measured by cavernous nerve electrostimulation with real time intracorporal pressure assessment. The penis was harvested for the histological examination (immunofluorescence and immunohistochemical staining and transmission electron microscopy detecting. Diabetic animals exhibited a decreased density of dorsal nerve bundle in penis. The neurofilament of the dorsal nerve bundle was fragmented in the diabetic rats. There was a decreased expression of nNOS and NGF in the diabetic group. The ICA II group had higher density of dorsal nerve bundle, higher expression of NGF and nNOS in the penis. The pathological change of major pelvic nerve ganglion (including the microstructure by transmission electron microscope and the neurite outgrowth length of major pelvic nerve ganglion tissue cultured in vitro was greatly attenuated in the ICA II-treated group (p < 0.01. ICA II treatment attenuates the diabetes-related impairment of corpus cavernosum and major pelvic ganglion neuropathy in rats with Streptozotocin-Induced Diabetes.

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

Major histocompatibility complex class II compatibility, but not class I, predicts mate choice in a bird with highly developed olfaction.

Science.gov (United States)

Strandh, Maria; Westerdahl, Helena; Pontarp, Mikael; Canbäck, Björn; Dubois, Marie-Pierre; Miquel, Christian; Taberlet, Pierre; Bonadonna, Francesco

2012-11-07

Mate choice for major histocompatibility complex (MHC) compatibility has been found in several taxa, although rarely in birds. MHC is a crucial component in adaptive immunity and by choosing an MHC-dissimilar partner, heterozygosity and potentially broad pathogen resistance is maximized in the offspring. The MHC genotype influences odour cues and preferences in mammals and fish and hence olfactory-based mate choice can occur. We tested whether blue petrels, Halobaena caerulea, choose partners based on MHC compatibility. This bird is long-lived, monogamous and can discriminate between individual odours using olfaction, which makes it exceptionally well suited for this analysis. We screened MHC class I and II B alleles in blue petrels using 454-pyrosequencing and quantified the phylogenetic, functional and allele-sharing similarity between individuals. Partners were functionally more dissimilar at the MHC class II B loci than expected from random mating (p = 0.033), whereas there was no such difference at the MHC class I loci. Phylogenetic and non-sequence-based MHC allele-sharing measures detected no MHC dissimilarity between partners for either MHC class I or II B. Our study provides evidence of mate choice for MHC compatibility in a bird with a high dependency on odour cues, suggesting that MHC odour-mediated mate choice occurs in birds.

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

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

Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

International Nuclear Information System (INIS)

Werry, E.V.

1990-10-01

This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

The PECS II block as a major analgesic component for clavicle operations: A description of 7 case reports.

Science.gov (United States)

Schuitemaker R, J B; Sala-Blanch, X; Rodriguez-Pérez, C L; Mayoral R, J T; López-Pantaleon, L A; Sánchez-Cohen, A P

2018-01-01

Clavicle fractures correspond to 35% of traumatic fractures of the shoulder girdle. Regional anaesthesia has shown better analgesic results than systemic treatment for perioperative management. Innervation of the clavicle is complex, at present its knowledge raises controversy. The lateral pectoral nerve through the innervating musculature predominantly participates in the lateral and anterior part of the clavicle. The following report of 7 cases describes the effective postoperative analgesia of modified PEC II block in patients with middle third clavicle fracture or acromioclavicular dislocation who underwent a modified PEC II block for postoperative pain management, in the context of a multimodal analgesia. The potential advantage of this management over other analgesic procedures should be evaluated in specific clinical trials. Copyright © 2017 Sociedad Española de Anestesiología, Reanimación y Terapéutica del Dolor. Publicado por Elsevier España, S.L.U. All rights reserved.

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.

Analytical throughput-estimating methods for the Hot Fuel Examination Facility

International Nuclear Information System (INIS)

Keyes, R.W.; Phipps, R.D.

1983-01-01

The Hot Fuel Examination Facility (HFEF) supports the operation and experimental programs of the major Liquid Metal Fast Breeder Reactor (LMFBR) test facilities; specifically, the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor II (EBR-II), and the Transient Reactor Test (TREAT) Facility. Successful management of HFEF and of LMFBR safety and fuels and materials programs, therefore, requires reliable information regarding HFEF's capability to handle expected or proposed program work loads. This paper describes the 10-step method that has been developed to consider all variables which significantly affect the HFEF examination throughput and quickly provide the necessary planning information

Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

1995-01-01

The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described

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

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume I

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

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

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.

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

Analysis of sodium valve reliability data at CREDO

International Nuclear Information System (INIS)

Bott, T.F.; Haas, P.M.

1979-01-01

The Centralized Reliability Data Organization (CREDO) has been established at Oak Ridge National Laboratory (ORNL) by the Department of Energy to provide a centralized source of data for reliability/maintainabilty analysis of advanced reactor systems. The current schedule calls for develoment of the data system at a moderate pace, with the first major distribution of data in late FY-1980. Continuous long-term collection of engineering, operating, and event data has been initiated at EBR-II and FFTF

IFR fuel cycle

International Nuclear Information System (INIS)

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

1992-01-01

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 development of the electrorefining pyroprocess, the design and construction of the facility for the hot demonstration, the design and fabrication of the equipment, and the schedule and initial plan for its operation

Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

International Nuclear Information System (INIS)

2000-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

(R)-3-hydroxyhexan-2-one is a major pheromone component of Anelaphus inflaticollis (Coleoptera: Cerambycidae).

Science.gov (United States)

Ray, A M; Swift, I P; Moreira, J A; Millar, J G; Hanks, L M

2009-10-01

We report the identification and field bioassays of a major component of the male-produced aggregation pheromone of Anelaphus inflaticollis Chemsak, an uncommon desert cerambycine beetle. Male A. inflaticollis produced a sex-specific blend of components that included (R)-3-hydroxyhexan-2-one, (S)-2-hydroxyhexan-3-one, 2,3-hexanedione, and (2R,3R)- and (2R,3S)-2,3-hexanediols. Field trials with baited bucket traps determined that the reconstructed synthetic pheromone blend and (R)-3-hydroxyhexan-2-one alone attracted adult A. inflaticollis of both sexes, with significantly more beetles being attracted to the blend. We conclude that (R)-3-hydroxyhexan-2-one is a major pheromone component of A. inflaticollis, and our results suggest that one or more of the minor components may further increase attraction of conspecifics. Scanning electron microscopy showed that male A. inflaticollis have pores on the prothorax that are consistent in structure with sex-specific pheromone gland pores in related species. Males also displayed stereotyped calling behavior similar to that observed in other cerambycine species. This study represents the first report of volatile pheromones for a cerambycine species in the tribe Elaphidiini.

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

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

Expression of major histocompatibility complex class II and costimulatory molecules in oral carcinomas in vitro.

Science.gov (United States)

Villarroel-Dorrego, Mariana; Speight, Paul M; Barrett, A William

2005-01-01

Recognition in the 1980 s that keratinocytes can express class II molecules of the Major Histocompatibility Complex (MHC) first raised the possibility that these cells might have an immunological function, and may even act as antigen presenting cells (APC). For effective T lymphocyte activation, APC require, in addition to MHC II, appropriate costimulatory signals. The aim of this study was to determine the expression of MHC class II and the co-stimulatory molecules CD40, CD80 and CD86 in keratinocytes derived from healthy oral mucosa and oral carcinomas. Using flow cytometry, it was confirmed that oral keratinocytes, switch on, expression of MHC class II molecules after stimulation with IFNgamma in vitro. All keratinocyte lines expressed CD40 constitutively; by contrast, CD80 and CD86 were universally absent. Loss of CD80 and CD86 may be one means whereby tumours escape immunological surveillance.

Development and verification test of integral reactor major components

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

Development and verification test of integral reactor major components

International Nuclear Information System (INIS)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

Quercetin, not caffeine, is a major neuroprotective component in coffee.

Science.gov (United States)

Lee, Moonhee; McGeer, Edith G; McGeer, Patrick L

2016-10-01

Epidemiologic studies indicate that coffee consumption reduces the risk of Parkinson's disease and Alzheimer's disease. To determine the factors involved, we examined the protective effects of coffee components. The test involved prevention of neurotoxicity to SH-SY5Y cells that was induced by lipopolysaccharide plus interferon-γ or interferon-γ released from activated microglia and astrocytes. We found that quercetin, flavones, chlorogenic acid, and caffeine protected SH-SY5Y cells from these toxins. They also reduced the release of tumor necrosis factor-α and interleukin-6 from the activated microglia and astrocytes and attenuated the activation of proteins from P38 mitogen-activated protein kinase (MAPK) and nuclear factor kappa light chain enhancer of activated B cells (NFκB). After exposure to toxin containing glial-stimulated conditioned medium, we also found that quercetin reduced oxidative/nitrative damage to DNA, as well as to the lipids and proteins of SH-SY5Y cells. There was a resultant increase in [GSH]i in SH-SY5Y cells. The data indicate that quercetin is the major neuroprotective component in coffee against Parkinson's disease and Alzheimer's disease. Copyright © 2016 Elsevier Inc. All rights reserved.

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

The precipitation response of 20%-cold-worked type 316 stainless steel to simulated fusion irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1979-01-01

The precipitation response of 20%-cold-worked type 316 stainless steel has been examined after irradiation in HFIR at 380-600 0 C, after irradiation in EBR-II at 500 0 C, and after thermal aging at 600 to 750 0 C. Eta phase forms during exposure to all environments. It constitutes a major portion of the precipitation response, and is rich in Ni, Si and Mo relative to M 23 C 6 after thermal aging. It is not normally reported in 20%-cold-worked type 316 stainless steel. The eta, M 23 C 6 , Laves, sigma, and chi precipitate phases appear at similar temperatures after HFIR, EBR-II, or thermal exposure. There are, however, some differences in relative amounts, size, and distribution of phases among the various environments. Eta phase is the only carbide-type phase observed after irradiation in HFIR from 380-550 0 C. The large cavities associated with it at 380 0 C contribute significantly to swelling. Re-solution of fine M 23 C 6 , eta, and Laves particles and re-precipitation of massive particles of sigma, M 23 C 6 and chi are observed after recrystallization in HFIR. (orig.)

Clinical, immunological and genetic features in eleven Algerian patients with major histocompatibility complex class II expression deficiency

Directory of Open Access Journals (Sweden)

Djidjik Réda

2012-08-01

Full Text Available Abstract Presenting processed antigens to CD4+ lymphocytes during the immune response involves major histocompatibility complex class II molecules. MHC class II genes transcription is regulated by four transcription factors: CIITA, RFXANK, RFX5 and RFXAP. Defects in these factors result in major histocompatibility complex class II expression deficiency, a primary combined immunodeficiency frequent in North Africa. Autosomal recessive mutations in the RFXANK gene have been reported as being the principal defect found in North African patients with this disorder. In this paper, we describe clinical, immunological and genetic features of 11 unrelated Algerian patients whose monocytes display a total absence of MHC class II molecules. They shared mainly the same clinical picture which included protracted diarrhoea and respiratory tract recurrent infections. Genetic analysis revealed that 9 of the 11 patients had the same RFXANK founder mutation, a 26 bp deletion (named I5E6-25_I5E6+1, also known as 752delG26. Immunological and genetic findings in our series may facilitate genetic counselling implementation for Algerian consanguineous families. Further studies need to be conducted to determine 752delG26 heterozygous mutation frequency in Algerian population.

Brassinosteroids-Induced Systemic Stress Tolerance was Associated with Increased Transcripts of Several Defence-Related Genes in the Phloem in Cucumis sativus.

Directory of Open Access Journals (Sweden)

Pingfang Li

Full Text Available Brassinosteroids (BRs, a group of naturally occurring plant steroidal compounds, are essential for plant growth, development and stress tolerance. Recent studies showed that BRs could induce systemic tolerance to biotic and abiotic stresses; however, the molecular mechanisms by which BRs signals lead to responses in the whole plant are largely unknown. In this study, 24-epibrassinosteroid (EBR-induced systemic tolerance in Cucumis sativus L. cv. Jinyan No. 4 was analyzed through the assessment of symptoms of photooxidative stress by chlorophyll fluorescence imaging pulse amplitude modulation. Expression of defense/stress related genes were induced in both treated local leaves and untreated systemic leaves by local EBR application. With the suppressive subtractive hybridization (SSH library using cDNA from the phloem sap of EBR-treated plants as the tester and distilled water (DW-treated plants as the driver, 14 transcripts out of 260 clones were identified. Quantitative Real Time-Polymerase Chain Reaction (RT-qPCR validated the specific up-regulation of these transcripts. Of the differentially expressed transcripts with known functions, transcripts for the selected four cDNAs, which encode an auxin-responsive protein (IAA14, a putative ankyrin-repeat protein, an F-box protein (PP2, and a major latex, pathogenesis-related (MLP-like protein, were induced in local leaves, systemic leaves and roots after foliar application of EBR onto mature leaves. Our results demonstrated that EBR-induced systemic tolerance is accompanied with increased transcript of genes in the defense response in other organs. The potential role of phloem mRNAs as signaling components in mediating BR-regulated systemic resistance is discussed.

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

Isolation and characterization of major histocompatibility complex class II B genes in cranes.

Science.gov (United States)

Kohyama, Tetsuo I; Akiyama, Takuya; Nishida, Chizuko; Takami, Kazutoshi; Onuma, Manabu; Momose, Kunikazu; Masuda, Ryuichi

2015-11-01

In this study, we isolated and characterized the major histocompatibility complex (MHC) class II B genes in cranes. Genomic sequences spanning exons 1 to 4 were amplified and determined in 13 crane species and three other species closely related to cranes. In all, 55 unique sequences were identified, and at least two polymorphic MHC class II B loci were found in most species. An analysis of sequence polymorphisms showed the signature of positive selection and recombination. A phylogenetic reconstruction based on exon 2 sequences indicated that trans-species polymorphism has persisted for at least 10 million years, whereas phylogenetic analyses of the sequences flanking exon 2 revealed a pattern of concerted evolution. These results suggest that both balancing selection and recombination play important roles in the crane MHC evolution.

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

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

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

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

Dynamic simulator for nuclear power plants (DSNP): development, verification, and expansion of modules

International Nuclear Information System (INIS)

Larson, H.A.; Dean, E.M.; Koenig, J.F.; Gale, J.G.; Lehto, W.K.

1984-01-01

The DSNP Simulation Language facilitates whole reactor plant simulation and design. Verification includes DSNP dynamic modeling of Experimental Breeder Reactor No. 2 (EBR-II) plant experiments as well as comparisons with verified simulation programs. Great flexibility is allowed in expanding the DSNP language and accommodate other computer languages. The component modules of DSNP, contained in libraries, are continually updated with new, improved, and verified modules. The modules are used to simulate the dynamic response of LMFBR reactor systems to upset and transient conditions, with special emphasis on investigations of inherent shutdown mechanisms

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

International Nuclear Information System (INIS)

Shin, Andong; Choi, Yong Won

2016-01-01

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Technical meeting on decommissioning of fast reactors after sodium draining. Working material

International Nuclear Information System (INIS)

2005-01-01

The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

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

Major Components of the National TRU Waste System Optimization Project

International Nuclear Information System (INIS)

Moody, D.C.; Bennington, B.; Sharif, F.

2002-01-01

The National Transuranic (TRU) Program (NTP) is being optimized to allow for disposing of the legacy TRU waste at least 10 years earlier than originally planned. This acceleration will save the nation an estimated $713. The Department of Energy's (DOE'S) Carlsbad Field Office (CBFO) has initiated the National TRU Waste System Optimization Project to propose, and upon approvaI, implement activities that produce significant cost saving by improving efficiency, thereby accelerating the rate of TRU waste disposal without compromising safety. In its role as NTP agent of change, the National TRU Waste System Optimization Project (the Project) (1) interacts closely with all NTP activities. Three of the major components of the Project are the Central Characterization Project (CCP), the Central Confirmation Facility (CCF), and the MobiIe/Modular Deployment Program.

Evidence that Mediator is essential for Pol II transcription, but is not a required component of the preinitiation complex in vivo.

Science.gov (United States)

Petrenko, Natalia; Jin, Yi; Wong, Koon Ho; Struhl, Kevin

2017-07-12

The Mediator complex has been described as a general transcription factor, but it is unclear if it is essential for Pol II transcription and/or is a required component of the preinitiation complex (PIC) in vivo. Here, we show that depletion of individual subunits, even those essential for cell growth, causes a general but only modest decrease in transcription. In contrast, simultaneous depletion of all Mediator modules causes a drastic decrease in transcription. Depletion of head or middle subunits, but not tail subunits, causes a downstream shift in the Pol II occupancy profile, suggesting that Mediator at the core promoter inhibits promoter escape. Interestingly, a functional PIC and Pol II transcription can occur when Mediator is not detected at core promoters. These results provide strong evidence that Mediator is essential for Pol II transcription and stimulates PIC formation, but it is not a required component of the PIC in vivo.

Characterization of the major nutritional components of Caryocar brasiliense fruit pulp by NMR spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Miranda-Vilela, Ana Luisa; Grisolia, Cesar Koppe [Universidade de Brasilia (UnB), DF (Brazil). Inst. de Ciencias Biologicas. Dept. de Genetica e Morfologia; Resck, Ines Sabioni; Mendonca, Marcio Antonio [Universidade de Brasilia (UnB), DF (Brazil). Inst. de Quimica

2009-07-01

Pequi (Caryocar brasiliense Camb.), a typical fruit of Brazilian Cerrado, is well known in regional cookery and used in folk medicine to treat various illnesses. Mass spectrometry and chromatographic methods have identified the organic composition of pequi fruit pulp; however, NMR spectroscopy is used for the first time to characterize the nutritional components of organic and aqueous-ethanolic extracts. This spectroscopic technique determined the triacylglycerols in the pequi organic fraction, which is constituted mainly by oleate and palmitate esters, and detected the carbohydrate mixtures as the major components of aqueous and ethanolic fractions, respectively. In this study, presence of phenolic compounds was only evidenced in the ethanolic fraction. (author)

Characterization of the major nutritional components of Caryocar brasiliense fruit pulp by NMR spectroscopy

International Nuclear Information System (INIS)

Miranda-Vilela, Ana Luisa; Grisolia, Cesar Koppe; Resck, Ines Sabioni; Mendonca, Marcio Antonio

2009-01-01

Pequi (Caryocar brasiliense Camb.), a typical fruit of Brazilian Cerrado, is well known in regional cookery and used in folk medicine to treat various illnesses. Mass spectrometry and chromatographic methods have identified the organic composition of pequi fruit pulp; however, NMR spectroscopy is used for the first time to characterize the nutritional components of organic and aqueous-ethanolic extracts. This spectroscopic technique determined the triacylglycerols in the pequi organic fraction, which is constituted mainly by oleate and palmitate esters, and detected the carbohydrate mixtures as the major components of aqueous and ethanolic fractions, respectively. In this study, presence of phenolic compounds was only evidenced in the ethanolic fraction. (author)

CD1 and major histocompatibility complex II molecules follow a different course during dendritic cell maturation

NARCIS (Netherlands)

van der Wel, Nicole N.; Sugita, Masahiko; Fluitsma, Donna M.; Cao, Xaiochun; Schreibelt, Gerty; Brenner, Michael B.; Peters, Peter J.

2003-01-01

The maturation of dendritic cells is accompanied by the redistribution of major histocompatibility complex (MHC) class II molecules from the lysosomal MHC class IT compartment to the plasma membrane to mediate presentation of peptide antigens. Besides MHC molecules, dendritic cells also express CD1

Effects of lifestyle interventions that include a physical activity component in class II and III obese individuals: a systematic review and meta-analysis.

Directory of Open Access Journals (Sweden)

Aurélie Baillot

Full Text Available In class II and III obese individuals, lifestyle intervention is the first step to achieve weight loss and treat obesity-related comorbidities before considering bariatric surgery. A systematic review, meta-analysis, and meta-regression were performed to assess the impact of lifestyle interventions incorporating a physical activity (PA component on health outcomes of class II and III obese individuals.An electronic search was conducted in 4 databases (Medline, Scopus, CINAHL and Sportdiscus. Two independent investigators selected original studies assessing the impact of lifestyle interventions with PA components on anthropometric parameters, cardiometabolic risk factors (fat mass, blood pressure, lipid and glucose metabolism, behaviour modification (PA and nutritional changes, and quality of life in adults with body mass index (BMI ≥ 35 kg/m2. Estimates were pooled using a random-effect model (DerSimonian and Laird method. Heterogeneity between studies was assessed by the Cochran's chi-square test and quantified through an estimation of the I².Of the 3,170 identified articles, 56 met our eligibility criteria, with a large majority of uncontrolled studies (80%. The meta-analysis based on uncontrolled studies showed significant heterogeneity among all included studies. The pooled mean difference in weight loss was 8.9 kg (95% CI, 10.2-7.7; p < 0.01 and 2.8 kg/m² in BMI loss (95% CI, 3.4-2.2; p < 0.01. Long-term interventions produced superior weight loss (11.3 kg compared to short-term (7.2 kg and intermediate-term (8.0 kg interventions. A significant global effect of lifestyle intervention on fat mass, waist circumference, blood pressure, total cholesterol, LDL-C, triglycerides and fasting insulin was found (p<0.01, without significant effect on HDL-C and fasting blood glucose.Lifestyle interventions incorporating a PA component can improve weight and various cardiometabolic risk factors in class II and III obese individuals. However, further

Temporal variations of PM1 major components in an urban street canyon.

Science.gov (United States)

Yubero, E; Galindo, N; Nicolás, J F; Crespo, J; Calzolai, G; Lucarelli, F

2015-09-01

Seasonal changes in the levels of PM1 and its main components (organic carbon (OC), elemental carbon (EC), SO4 (2-), NO3 (-) and NH4 (+)) were studied in an urban street canyon in southeastern Spain. Although PM1 levels did not show an evident seasonal cycle, strong variations in the concentrations of its major components were observed. Ammonium sulfate, the main secondary inorganic compound, was found to be of regional origin. Its formation was favored during summer due to increased photochemical activity. In contrast, the concentrations of particulate ammonium nitrate, which is thermally unstable, were highest in winter. Although traffic emissions are the dominant source of EC in the city, variations in traffic intensity could not explain the seasonal cycle of this component. The higher EC concentrations during the cold months were attributed to the lower dispersion conditions and the increase in EC emissions. Special attention has been given to variations in organic carbon levels since it accounted for about one third of the total PM1 mass. The concentrations of both total OC and secondary OC (SOC) were maxima in winter. The observed seasonal variation in SOC levels is similar to that found in other southern European cities where the frequency of sunny days in winter is high enough to promote photochemical processes.

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

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.

Diagnosis of the Main Busbar II Panel Components Ageing of RSG-GAS Electrical System by Using Infrared Thermography

International Nuclear Information System (INIS)

Teguh Sulistyo; Kiswanto; Roziq Himawan; Ari Satmoko

2007-01-01

To support the operation of RSG-GAS safely, the diagnosis of the ageing of main busbar II BHD/BHE/BHF panel components of RSG-GAS electrical system have been done. By using infrared thermography type Thermo Tracer TH9100PM VI/PW VI. The results of the diagnosis showed that some of the components under degradation with various rate. It can cause the system failure. By understanding the components ageing degradation mechanism and performing the preventive and predictive maintenance and safety of RSG-GAS electrical system earlier, the possibility of accident can be avoided. (author)

Does temperamental instability support a continuity between bipolar II disorder and major depressive disorder?

Science.gov (United States)

Benazzi, F

2006-06-01

The current categorical split of mood disorders in bipolar disorders and depressive disorders has recently been questioned. Two highly unstable personality features, i.e. the cyclothymic temperament (CT) and borderline personality disorder (BPD), have been found to be more common in bipolar II (BP-II) disorder than in major depressive disorder (MDD). According to Kraepelin, temperamental instability was the "foundation" of his unitary view of mood disorders. The aim was to assess the distributions of the number of CT and borderline personality items between BP-II and MDD. Finding no bi-modal distribution (a "zone of rarity") of these items would support a continuity between the two disorders. an outpatient psychiatry private practice. Interviewer: A senior clinical and mood disorder research psychiatrist. A consecutive sample of 138 BP-II and 71 MDD remitted outpatients. Assessment instruments: The structured clinical interview for DSM-IV Axis I Disorders-Clinician Version (SCID-CV), the SCID-II Personality Questionnaire for self-assessing borderline personality traits (BPT) by patients, the TEMPS-A for self-assessing CT by patients. Interview methods: Patients were interviewed with the SCID-CV to diagnose BP-II and MDD, and then patients self-assessed the questions of the Personality Questionnaire relative to borderline personality, and the questions of the TEMPS-A relative to CT. As clinically significant distress or impairment of functioning is not assessed by the SCID-II Personality Questionnaire, a diagnosis of BPD could not be made, but BPT could be assessed (i.e. all BPD items but not the impairment criterion). The distribution of the number of CT and BPT items was studied by Kernel density estimate. CT and BPT items were significantly more common in BP-II versus MDD. The Kernel density estimate distributions of the number of CT and BPT items in the entire sample had a normal-like shape (i.e. no bi-modality). The expected finding, on the basis of previous

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.

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

International Nuclear Information System (INIS)

Won-Seok Kim; Young-Gyun Kim

2000-01-01

In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

RELAP5 based engineering simulator

International Nuclear Information System (INIS)

Charlton, T.R.; Laats, E.T.; Burtt, J.D.

1990-01-01

The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor II (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nuclear Plant Analyzer (NPA), have recently been improved significantly. This paper will present an overview of the INEL Engineering Simulation Center, and discuss the RELAP5/MOD3 and NPA/MOD1 codes, specifically how they are being used at the INEL Engineering Simulation Center. It will provide an update on the modifications to these two codes and their application to the ATR engineering simulator project, as well as, a discussion on the reactor system representation, control system modeling, two phase flow and heat transfer modeling. It will also discuss how these two codes are providing desktop, stand-alone reactor simulation. 12 refs., 2 figs

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

Identification of human cytochrome P450 and UGT enzymes involved in the metabolism of ferulic acid, a major bioactive component in traditional Chinese medicines.

Science.gov (United States)

Zhuang, Xiao-Mei; Chen, Lin; Tan, Yan; Yang, Hai-Ying; Lu, Chuang; Gao, Yue; Li, Hua

2017-09-01

Ferulic acid (FA) is an active component of herbal medicines. One of the best documented activities of FA is its antioxidant property. Moreover, FA exerts antiallergic, anti-inflammatory, and hepatoprotective effects. However, the metabolic pathways of FA in humans remain unclear. To identify whether human CYP or UGT enzymes are involved in the metabolism of FA, reaction phenotyping of FA was conducted using major CYP-selective chemical inhibitors together with individual CYP and UGT Supersomes. The CYP- and/or UGT-mediated metabolism kinetics were examined simultaneously or individually. Relative activity factor and total normalized rate approaches were used to assess the relative contributions of each major human CYPs towards the FA metabolism. Incubations of FA with human liver microsomes (HLM) displayed NADPH- and UDPGA-dependent metabolism with multiple CYP and UGT isoforms involved. CYPs and UGTs contributed equally to the metabolism of FA in HLM. Although CYP1A2 and CYP3A4 appeared to be the major contributors in the CYP-mediated clearance, their contributions to the overall clearance are still minor (medicines because multiple phase I and phase II enzymes are involved in its metabolism. Copyright © 2017 China Pharmaceutical University. Published by Elsevier B.V. All rights reserved.

Implementing partnerships in nonreactor facility safety analyses

International Nuclear Information System (INIS)

Courtney, J.C.; Perry, W.H.; Phipps, R.D.

1996-01-01

Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

Functional role of the cytoplasmic tail domain of the major envelope fusion protein of group II baculoviruses

NARCIS (Netherlands)

Long, G.; Pan, M.; Westenberg, M.; Vlak, J.M.

2006-01-01

F proteins from baculovirus nucleopolyhedrovirus (NPV) group II members are the major budded virus (BV) viral envelope fusion proteins. They undergo furin-like proteolysis processing in order to be functional. F proteins from different baculovirus species have a long cytoplasmic tail domain (CTD),

Methodology to obtain exchange properties of the calcite surface-Application to major and trace elements: Ca(II), HCO3-, and Zn(II)

International Nuclear Information System (INIS)

Tertre, E.; Beaucaire, C.; Juery, A.; Ly, J.; Tertre, E.; Beaucaire, C.; Juery, A.; Ly, J.

2010-01-01

Sorption of inorganic elements onto carbonate minerals has been intensively described in the literature by two reaction steps: (1) a first one rapid and completed within a few hours and (2) a second one slower, eventually irreversible, and occurring at a constant rate. The first step is often attributed to an ion-exchange process, but its reversibility is rarely investigated. Consequently, discrimination of the global sorption phenomenon into two different mechanisms is not always justified. In this study, we investigated, by batch experiments, both sorption and desorption of Ca(II), HCO 3 - , and Zn(II), radiolabeled with isotopes 45 Ca(II), H 14 CO 3 - , and 65 Zn(II), respectively, onto synthetic pure calcite. Solutions were pre-equilibrated with atmospheric p(CO 2 ) and saturated with respect to calcite. Therefore, our purpose was to: (1) obtain experimental distribution coefficients of major elements (Ca(II) and HCO 3 - ) and a trace element (Zn(II)) onto calcite from sorption and desorption experiments, (2) test the validity of a first-occurring ion-exchange process generally noted in the literature, by calculating distribution coefficients for the 'sole' exchange process, and (3) quantify the amounts of Ca(II), HCO 3 - , and Zn(II) sorbed on the calcite surface by the sole 'exchange process' and compare them with surface crystallochemical data. Ca(II) or HCO 3 - sorption experimental data suggest that a significant fraction of these two elements was sorbed irreversibly onto or in the calcite. By using a method based on isotopic ratios, the Ca(II) or HCO 3 - concentrations, which are reversibly adsorbed on the calcite, have been quantified. These concentrations are respectively estimated at 4. 0 ± 2. 0 * 10 -4 and 7. 0 ± 1. 5 * 10 -4 mol/kg. The obtained Ca(II) surface concentration value is one order of magnitude lower than the one obtained from isotopic measurement by former authors [Geochim. Cosmochim. Acta 55 (1991) 1549; Geochim. Cosmochim. Acta 51

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.)

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)

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

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.

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

Small Diameter Bomb Increment II (SDB II)

Science.gov (United States)

2015-12-01

Selected Acquisition Report (SAR) RCS: DD-A&T(Q&A)823-439 Small Diameter Bomb Increment II (SDB II) As of FY 2017 President’s Budget Defense... Bomb Increment II (SDB II) DoD Component Air Force Joint Participants Department of the Navy Responsible Office References SAR Baseline (Production...Mission and Description Small Diameter Bomb Increment II (SDB II) is a joint interest United States Air Force (USAF) and Department of the Navy

Leaching of cell wall components caused by acid deposition on fir needles and trees

Energy Technology Data Exchange (ETDEWEB)

Shigihara, Ado [Department of Material and Life Chemistry, Faculty of Engineering, Kanagawa University, 3-27-1, Rokkakubashi, Kanagawa-ku, Yokohama 221-8686 (Japan)], E-mail: r200670202@kanagawa-u.ac.jp; Matsumoto, Kiyoshi [Department of Material and Life Chemistry, Faculty of Engineering, Kanagawa University, 3-27-1, Rokkakubashi, Kanagawa-ku, Yokohama 221-8686 (Japan); Sakurai, Naoki [Faculty of Integrated Arts and Science, Hiroshima University, 1-7-1, Kagamiyama, Higashi-Hiroshima, 739-8521 (Japan); Igawa, Manabu [Department of Material and Life Chemistry, Faculty of Engineering, Kanagawa University, 3-27-1, Rokkakubashi, Kanagawa-ku, Yokohama 221-8686 (Japan)

2008-07-15

Virgin fir forests have been declining since the 1960s at Mt. Oyama, which is located at the eastern edge of the Tanzawa Mountains and adjacent to the Kanto plain in Japan. An acid fog frequently occurs in the mountains. We collected throughfall and stemflow under fir trees and rainfall every week during January-December 2004 at Mt. Oyama to clarify the influence of acid fog on the decline of fir (Abies firma) needles. In relation to throughfall and stemflow, D-mannose, D-galactose, and D-glucose are the major neutral sugar components; only D-glucose is a major component of rainfall. The correlation coefficient between the total neutral sugars and uronic acid (as D-galacturonic acid), which is a key component of the cross-linking between pectic polysaccharides, was high except for rainfall. The leached amount of calcium ion, neutral sugars, uronic acid, and boron is related to the nitrate ion concentration in throughfall. Results of a laboratory exposure experiment using artificial fog water simulating the average composition of fog water observed at Mt. Oyama (simulated acid fog: SAF) on the fir seedling needles also shows a large leaching of these components from the cell walls of fir needles. The leaching amount increased concomitantly with decreasing pH of the SAF solution. We also observed that a dimeric rhamnogalacturonan II-borate complex (dRG-II-B) that exists in the cell wall as pectic polysaccharide was converted to monomeric RG-II (mRG-II) by the leaching of calcium ion and boron. Results not only of field observations but also those of laboratory experiments indicate a large effect of acid depositions on fir needles.

A major genetic component of BSE susceptibility

Science.gov (United States)

Juling, Katrin; Schwarzenbacher, Hermann; Williams, John L; Fries, Ruedi

2006-01-01

Background Coding variants of the prion protein gene (PRNP) have been shown to be major determinants for the susceptibility to transmitted prion diseases in humans, mice and sheep. However, to date, the effects of polymorphisms in the coding and regulatory regions of bovine PRNP on bovine spongiform encephalopathy (BSE) susceptibility have been considered marginal or non-existent. Here we analysed two insertion/deletion (indel) polymorphisms in the regulatory region of bovine PRNP in BSE affected animals and controls of four independent cattle populations from UK and Germany. Results In the present report, we show that two previously reported 23- and 12-bp insertion/deletion (indel) polymorphisms in the regulatory region of bovine PRNP are strongly associated with BSE incidence in cattle. Genotyping of BSE-affected and control animals of UK Holstein, German Holstein, German Brown and German Fleckvieh breeds revealed a significant overrepresentation of the deletion alleles at both polymorphic sites in diseased animals (P = 2.01 × 10-3 and P = 8.66 × 10-5, respectively). The main effect on susceptibility is associated with the 12-bp indel polymorphism. Compared with non-carriers, heterozygous and homozygous carriers of the 12-bp deletion allele possess relatively higher risks of having BSE, ranging from 1.32 to 4.01 and 1.74 to 3.65 in the different breeds. These values correspond to population attributable risks ranging from 35% to 53%. Conclusion Our results demonstrate a substantial genetic PRNP associated component for BSE susceptibility in cattle. Although the BSE risk conferred by the deletion allele of the 12-bp indel in the regulatory region of PRNP is substantial, the main risk factor for BSE in cattle is environmental, i.e. exposure to feedstuffs contaminated with the infectious agent. PMID:17014722

A major genetic component of BSE susceptibility

Directory of Open Access Journals (Sweden)

Williams John L

2006-10-01

Full Text Available Abstract Background Coding variants of the prion protein gene (PRNP have been shown to be major determinants for the susceptibility to transmitted prion diseases in humans, mice and sheep. However, to date, the effects of polymorphisms in the coding and regulatory regions of bovine PRNP on bovine spongiform encephalopathy (BSE susceptibility have been considered marginal or non-existent. Here we analysed two insertion/deletion (indel polymorphisms in the regulatory region of bovine PRNP in BSE affected animals and controls of four independent cattle populations from UK and Germany. Results In the present report, we show that two previously reported 23- and 12-bp insertion/deletion (indel polymorphisms in the regulatory region of bovine PRNP are strongly associated with BSE incidence in cattle. Genotyping of BSE-affected and control animals of UK Holstein, German Holstein, German Brown and German Fleckvieh breeds revealed a significant overrepresentation of the deletion alleles at both polymorphic sites in diseased animals (P = 2.01 × 10-3 and P = 8.66 × 10-5, respectively. The main effect on susceptibility is associated with the 12-bp indel polymorphism. Compared with non-carriers, heterozygous and homozygous carriers of the 12-bp deletion allele possess relatively higher risks of having BSE, ranging from 1.32 to 4.01 and 1.74 to 3.65 in the different breeds. These values correspond to population attributable risks ranging from 35% to 53%. Conclusion Our results demonstrate a substantial genetic PRNP associated component for BSE susceptibility in cattle. Although the BSE risk conferred by the deletion allele of the 12-bp indel in the regulatory region of PRNP is substantial, the main risk factor for BSE in cattle is environmental, i.e. exposure to feedstuffs contaminated with the infectious agent.

The Mark II detector for the SLC

International Nuclear Information System (INIS)

Abrams, G.; Baden, A.R.; Boyer, J.; Butler, F.; Drell, P.S.; Fay, J.; Gidal, G.; Goldhaber, G.; Haggerty, J.; Harr, R.; Hearty, C.; Herrup, D.; Holmgren, S.O.; Jaffre, M.; Juricic, I.; Kadyk, J.A.; Kral, J.F.; Levi, M.E.; Lynch, G.R.; Richman, J.D.; Rouse, F.R.; Schaad, M.W.; Schmidke, W.B.; Schumm, B.A.; Trilling, G.H.; Wood, D.R.; Akerlof, C.; Bonvicini, G.; Chapman, J.; Chmeissani, M.; Frey, R.; Gero, E.; Hong, S.J.; Koska, W.; Nitz, D.; Petradza, M.; Thun, R.; Tschirhart, R.; Veltman, H.; Alexander, J.P.; Ballam, J.; Barklow, T.; Bartelt, J.; De Boer, W.; Boyarski, A.; Braune, K.; Bulos, F.; Burke, D.L.; Cords, D.; Coupal, D.P.; Destaebler, H.C.; Dorfan, J.M.; Feldman, G.J.; Fernandes, D.; Field, R.C.; Fordham, C.; Fujino, D.; Gan, K.K.; Glanzman, T.; Grosse-Wiesmann, P.; Hanson, G.; Hayes, K.; Himel, T.; Hutchinson, D.; Innes, W.R.; Jacobsen, R.G.; Jaros, J.A.; Jung, C.K.; Karlen, D.; Klein, S.R.; Koetke, D.; Komamiya, M.; Kowalski, L.A.; Kozanecki, W.; Lankford, A.J.; Larsen, R.R.; Lueth, V.; Mattison, T.; Moffeit, K.C.; Mueller, L.; Munger, C.T.; Nash, J.; Ong, R.A.; O'Shaughnessy, K.F.; Perl, J.; Perl, M.L.; Perrier, F.; Petersen, A.; Pitthan, R.; Riles, K.; Swartz, M.; Taylor, R.E.; Van Kooten, R.; Voruganti, P.; Weigend, A.; Woods, M.; Wormser, G.; Wright, R.; Alvarez, M.; Calvino, F.; Fernandez, E.; Ford, W.T.; Hinshaw, D.A.; Rankin, P.; Smith, J.G.; Wagner, S.R.; Weber, P.; White, S.L.; Averill, D.; Blockus, D.; Brabson, B.; Brom, J.M.; Murray, W.N.; Ogren, H.; Rust, D.R.; Snyder, A.; Yurko, M.; Barish, B.C.; Hawkes, C.M.; Hoenk, M.; Kuhlen, M.; Li, Z.; McKenna, J.A.; Milliken, B.D.; Nelson, M.E.; Peck, C.; Porter, F.C.; Soderstrom, E.; Stroynowski, R.; Weinstein, A.J.; Weir, A.J.; Wicklund, E.; Wolf, R.C.; Wu, D.Y.; Barnett, B.A.; Boswell, C.; Dauncey, P.; Drewer, D.C.; Harral, B.; Hylen, J.; Matthews, J.A.J.; Stoker, D.P.; Vejcik, S.; Breakstone, A.; Cence, R.J.; Gong, X.; Harris, F.A.; Koide, A.; Parker, S.I.; Green, A.; Lawrence Berkeley Lab., CA; California Univ., Berkeley

1989-01-01

The Mark II detector has been upgraded in preparation for its role as the first detector to take data at the Stanford Linear Collider. The new detector components include the central drift chamber, the time-of-flight system, the coil, the endcap electromagnetic calorimeters and the beam energy and luminosity measuring devices. There have also been improvements in detector hermeticity. All of the major components were installed for a test run at the PEP storage ring (√s=29 GeV) in 1985. This paper describes the upgraded detector, including its trigger and data acquisition systems, and gives performance figures for its components. Future improvements are also discussed. (orig.)

Melittin, a major peptide component of bee venom, and its conjugates in cancer therapy

OpenAIRE

Rady, Islam; Siddiqui, Imtiaz A.; Rady, Mohamad; Mukhtar, Hasan

2017-01-01

Melittin (MEL), a major peptide component of bee venom, is an attractive candidate for cancer therapy. This agent has shown a variety of anti-cancer effects in preclinical cell culture and animal model systems. Despite a convincing efficacy data against variety of cancers, its applicability to humans has met with challenges due to several issues including its non-specific cytotoxicity, degradation and hemolytic activity. Several optimization approaches including utilization of nanoparticle ba...

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

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

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

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Design and development of major balance of plant components in solid oxide fuel cell system

Energy Technology Data Exchange (ETDEWEB)

Hong, Wen-Tang; Huang, Cheng-Nan; Tan, Hsueh-I; Chao, Yu [Institute of Nuclear Energy Research Atomic Energy Council, Taoyuan County 32546 (Taiwan, Province of China); Yen, Tzu-Hsiang [Green Technology Research Institute, CPC Corporation, Chia-Yi City 60036 (Taiwan, Province of China)

2013-07-01

The balance of plant (BOP) of a Solid Oxide Fuel Cell (SOFC) system with a 2 kW stack and an electric efficiency of 40% is optimized using commercial GCTool software. The simulation results provide a detailed understanding of the optimal operating temperature, pressure and mass flow rate in all of the major BOP components, i.e., the gas distributor, the afterburner, the reformer and the heat exchanger. A series of experimental trials are performed to validate the simulation results. Overall, the results presented in this study not only indicate an appropriate set of operating conditions for the SOFC power system, but also suggest potential design improvements for several of the BOP components.

Chemical composition of the major components of PM in different sites at the Metropolitan Region of Chile

Science.gov (United States)

Reyes, F.; Castillo, M. A.; Rubio, M.; Gramsch, E.; Vasquez, Y.; Oyola, P.

2013-05-01

Santiago, Chile's capital is one of most polluted megacity (5.5 million of people) of the world. Currently, PM2.5 annual concentration is over 2.2 times the Chilean standard (20 μg/m3). Continuous measurements of non-refractory PM1.0 (sulfate, nitrate, chloride, ammonium and organics aerosols), black carbon, and PM2,5 mass concentration were determined using Aerosol Chemical Speciation Monitor (ACSM, Aerodyne Research, Inc), absorption coefficient monitor (SIMCA, Santiago University) and dustrack monitor (TSI Inc) in order to know the temporal variability of the major components of PM. The measurements were carried out at kerbside, urban background, industrial and mixed residential/industrial locations during year 2012 and -2013. Meteorological data (Relative Humidity, temperature, wind speed, wind direction and precipitations) were obtained from the air quality network operated by the environmental authority. The results show strong correlation with the metropolitan region major sources. Multiple regression analysis indicates that precipitations have a strong impact on PM1.0 soluble components; relative humidity has effects only on chloride, sulfate and black carbon. Chloride concentration decrease when temperature is increasing. The perceptual contribution of each component is similar among all sites. All sites shows that OA (Organics Aerosol) as the major constituent of PM1.0 (>50%), followed of nitrates (>13%). Sulfate could be used to differentiate the industrial site; due to there is a strong impact of SO2 emission. Combustion sources direct impact can be seen at BC contribution at industrial and kerbside site. Also, the OA/BC ratio shows slow value at kerbside (3.05) and industrial (3.26) site, and higher at urban background site (4.15). Aged organics aerosols are majority found at all sites (f43/f44 plot), indicating that regional background is strong in all results. These results will be compared with size distribution measurements available from previous

STELLAR ARCHAEOLOGY IN THE GALACTIC HALO WITH THE ULTRA-FAINT DWARFS. VI. URSA MAJOR II

Energy Technology Data Exchange (ETDEWEB)

Dall' Ora, M.; Ripepi, Vincenzo; Marconi, Marcella; Musella, Ilaria, E-mail: dallora@na.astro.it, E-mail: ripepi@na.astro.it, E-mail: marcella@na.astro.it, E-mail: ilaria@na.astro.it [INAF, Osservatorio Astronomico di Capodimonte, Napoli (Italy); and others

2012-06-10

We present a B, V color-magnitude diagram (CMD) of the Milky Way dwarf satellite Ursa Major II (UMa II), spanning the magnitude range from V {approx} 15 to V {approx} 23.5 mag and extending over an 18 Multiplication-Sign 18 arcmin{sup 2} area centered on the Galaxy. Our photometry goes down to about 2 mag below the Galaxy's main-sequence turnoff that we detected at V {approx} 21.5 mag. We have discovered a bona fide RR Lyrae variable star in UMa II, which we use to estimate a conservative dereddened distance modulus for the galaxy of (m - M){sub 0} = 17.70 {+-} 0.04 {+-} 0.12 mag, where the first error accounts for the uncertainties of the calibrated photometry, and the second reflects our lack of information on the metallicity of the star. The corresponding distance to UMa II is 34.7{sup +0.6}{sub -0.7}({sup +2.0}{sub -1.9}) kpc. Our photometry shows evidence of a spread in the Galaxy's subgiant branch, compatible with a spread in metal abundance in the range between Z = 0.0001 and Z = 0.001. Based on our estimate of the distance, a comparison of the fiducial lines of the Galactic globular clusters M68 and M5 ([Fe/H] = -2.27 {+-} 0.04 dex and -1.33 {+-} 0.02 dex, respectively), with the position on the CMD of spectroscopically confirmed Galaxy members, may suggest the existence of stellar populations of different metal abundance/age in the central region of UMa II.

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

Characterisation of four major histocompatibility complex class II genes of the koala (Phascolarctos cinereus).

Science.gov (United States)

Lau, Quintin; Jobbins, Sarah E; Belov, Katherine; Higgins, Damien P

2013-01-01

Major histocompatibility complex (MHC) class II molecules have an integral role in the adaptive immune response, as they bind and present antigenic peptides to T helper lymphocytes. In this study of koalas, species-specific primers were designed to amplify exon 2 of the MHC class II DA and DB genes, which contain much of the peptide-binding regions of the α and β chains. A total of two DA α1 domain variants and eight DA β1 (DAB), three DB α1 and five DB β1 variants were amplified from 20 koalas from two free-living populations from South East Queensland and the Port Macquarie region in northern New South Wales. We detected greater variation in the β1 than in the α1 domains as well as evidence of positive selection in DAB. The present study provides a springboard to future investigation of the role of MHC in disease susceptibility in koalas.

Acute effects of decaffeinated coffee and the major coffee components chlorogenic acid and trigonelline on glucose tolerance

NARCIS (Netherlands)

van Dijk, A.E.; Olthof, M.R.; Meeuse, J.C.; Seebus, E.; Heine, R.J.; van Dam, R.M.

2009-01-01

OBJECTIVE - Coffee consumption has been associated with lower risk of type 2 diabetes. We evaluated the acute effects of decaffeinated coffee and the major coffee components chlorogenic acid and trigonelline on glucose tolerance. RESEARCH DESIGN AND METHODS - We conducted a randomized crossover

Effects of major histocompatibility complex class II knockout on mouse bone mechanical properties during development

Science.gov (United States)

Simske, Steven J.; Bateman, Ted A.; Smith, Erin E.; Ferguson, Virginia L.; Chapes, Stephen K.

2002-01-01

We investigated the effect of major histocompatibility complex class II (MHC II) knockout on the development of the mouse peripheral skeleton. These C2D mice had less skeletal development at 8, 12 and 16 weeks of age compared to wild-type C57BL/6J (B6) male mice. The C2D mice had decreased femur mechanical, geometric and compositional measurements compared to wild type mice at each of these ages. C2D femur stiffness (S), peak force in 3-pt bending (Pm), and mineral mass (Min-M) were 74%, 64% and 66%, respectively, of corresponding B6 values at 8 weeks of age. Similar differences were measured at 12 weeks (for which C2D femoral S, Pm and Min-M were 71%, 72% and 73%, respectively, of corresponding B6 values) and at 16 weeks (for which C2D femoral S, Pm and Min-M were 80%, 66% and 61%, respectively, of corresponding B6 values). MHC II knockout delays the development of adult bone properties and is accompanied by lower body mass compared to wild-type controls.

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

Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant

Influence of collagen type II and nucleus pulposus cells on aggregation and differentiation of adipose tissue-derived stem cells

NARCIS (Netherlands)

Lu, Z.F.; Zandieh Doulabi, B.; Wuisman, P.I.; Bank, R.A.; Helder, M.N.

2008-01-01

Tissue microenvironment plays a critical role in guiding local stem cell differentiation. Within the intervertebral disc, collagen type II and nucleus pulposus (NP) cells are two major components. This study aimed to investigate how collagen type II and NP cells affect adipose tissue-derived stem

Autoimmunity and inflammation are independent of class II transactivator type PIV-dependent class II major histocompatibility complex expression in peripheral tissues during collagen-induced arthritis.

Science.gov (United States)

Waldburger, Jean-Marc; Palmer, Gaby; Seemayer, Christian; Lamacchia, Celine; Finckh, Axel; Christofilopoulos, Panayiotis; Baeten, Dominique; Reith, Walter; Gabay, Cem

2011-11-01

To determine the regulation of class II major histocompatibility complex (MHC) expression in fibroblast-like synoviocytes (FLS) in order to investigate their role as nonprofessional antigen-presenting cells in collagen-induced arthritis (CIA). Expression of class II MHC, class II MHC transactivator (CIITA), and Ciita isoforms PI, PIII, and PIV was examined by real-time quantitative polymerase chain reaction, immunohistochemistry, and flow cytometry in human synovial tissues, arthritic mouse joints, and human and murine FLS. CIA was induced in mice in which isoform PIV of Ciita was knocked out (PIV(-/-) ), in PIV(-/-) mice transgenic for CIITA in the thymus (K14 CIITA), and in their control littermates. HLA-DRA, total CIITA, and CIITA PIII messenger RNA levels were significantly increased in synovial tissue samples from patients with rheumatoid arthritis compared with the levels in tissue from patients with osteoarthritis. Human FLS expressed surface class II MHC via CIITA PIII and PIV, while class II MHC expression in murine FLS was entirely mediated by PIV. Mice with a targeted deletion of CIITA PIV lack CD4+ T cells and were protected against CIA. The expression of CIITA was restored in the thymus of PIV(-/-) K14 CIITA-transgenic mice, which had a normal CD4+ T cell repertoire and normal surface levels of class II MHC on professional antigen-presenting cells, but did not induce class II MHC on FLS. Synovial inflammation and immune responses against type II collagen were similar in PIV(-/-) K14 CIITA-transgenic mice and control mice with CIA, but bone erosion was significantly reduced in the absence of PIV. Overexpression of class II MHC is tightly correlated with CIITA expression in arthritic synovium and in FLS. Selective targeting of Ciita PIV in peripheral tissues abrogates class II MHC expression by murine FLS but does not protect against inflammation and autoimmune responses in CIA. Copyright © 2011 by the American College of Rheumatology.

Distinguishing bipolar II depression from major depressive disorder with comorbid borderline personality disorder: demographic, clinical, and family history differences.

Science.gov (United States)

Zimmerman, Mark; Martinez, Jennifer H; Morgan, Theresa A; Young, Diane; Chelminski, Iwona; Dalrymple, Kristy

2013-09-01

Because of the potential treatment implications, it is clinically important to distinguish between bipolar II depression and major depressive disorder with comorbid borderline personality disorder. The high frequency of diagnostic co-occurrence and resemblance of phenomenological features has led some authors to suggest that borderline personality disorder is part of the bipolar spectrum. Few studies have directly compared patients with bipolar disorder and borderline personality disorder. In the present study from the Rhode Island Methods to Improve Diagnostic Assessment and Services project, we compared these 2 groups of patients on demographic, clinical, and family history variables. From December 1995 to May 2012, 3,600 psychiatric patients presenting to the outpatient practice at Rhode Island Hospital (Providence, Rhode Island) were evaluated with semistructured diagnostic interviews for DSM-IV Axis I and Axis II disorders. The focus of the present study is the 206 patients with DSM-IV major depressive disorder and borderline personality disorder (MDD-BPD) and 62 patients with DSM-IV bipolar II depression without borderline personality disorder. The patients with MDD-BPD were significantly more often diagnosed with posttraumatic stress disorder (P depression had a significantly higher morbid risk for bipolar disorder in their first-degree relatives than the MDD-BPD patients (P depression and major depressive disorder with comorbid borderline personality disorder differed on a number of clinical and family history variables, thereby supporting the validity of this distinction. © Copyright 2013 Physicians Postgraduate Press, Inc.

Indirect dark matter searches in the dwarf satellite galaxy Ursa Major II with the MAGIC telescopes

Science.gov (United States)

Ahnen, M. L.; Ansoldi, S.; Antonelli, L. A.; Arcaro, C.; Baack, D.; Babić, A.; Banerjee, B.; Bangale, P.; Barres de Almeida, U.; Barrio, J. A.; Becerra González, J.; Bednarek, W.; Bernardini, E.; Berse, R. Ch.; Berti, A.; Bhattacharyya, W.; Biland, A.; Blanch, O.; Bonnoli, G.; Carosi, R.; Carosi, A.; Ceribella, G.; Chatterjee, A.; Colak, S. M.; Colin, P.; Colombo, E.; Contreras, J. L.; Cortina, J.; Covino, S.; Cumani, P.; Da Vela, P.; Dazzi, F.; De Angelis, A.; De Lotto, B.; Delfino, M.; Delgado, J.; Di Pierro, F.; Domínguez, A.; Dominis Prester, D.; Dorner, D.; Doro, M.; Einecke, S.; Elsaesser, D.; Fallah Ramazani, V.; Fernández-Barral, A.; Fidalgo, D.; Fonseca, M. V.; Font, L.; Fruck, C.; Galindo, D.; García López, R. J.; Garczarczyk, M.; Gaug, M.; Giammaria, P.; Godinović, N.; Gora, D.; Guberman, D.; Hadasch, D.; Hahn, A.; Hassan, T.; Hayashida, M.; Herrera, J.; Hose, J.; Hrupec, D.; Ishio, K.; Konno, Y.; Kubo, H.; Kushida, J.; Kuveždić, D.; Lelas, D.; Lindfors, E.; Lombardi, S.; Longo, F.; López, M.; Maggio, C.; Majumdar, P.; Makariev, M.; Maneva, G.; Manganaro, M.; Mannheim, K.; Maraschi, L.; Mariotti, M.; Martínez, M.; Masuda, S.; Mazin, D.; Mielke, K.; Minev, M.; Miranda, J. M.; Mirzoyan, R.; Moralejo, A.; Moreno, V.; Moretti, E.; Nagayoshi, T.; Neustroev, V.; Niedzwiecki, A.; Nievas Rosillo, M.; Nigro, C.; Nilsson, K.; Ninci, D.; Nishijima, K.; Noda, K.; Nogués, L.; Paiano, S.; Palacio, J.; Paneque, D.; Paoletti, R.; Paredes, J. M.; Pedaletti, G.; Peresano, M.; Persic, M.; Prada Moroni, P. G.; Prandini, E.; Puljak, I.; Garcia, J. R.; Reichardt, I.; Rhode, W.; Ribó, M.; Rico, J.; Righi, C.; Rugliancich, A.; Saito, T.; Satalecka, K.; Schweizer, T.; Sitarek, J.; Šnidarić, I.; Sobczynska, D.; Stamerra, A.; Strzys, M.; Surić, T.; Takahashi, M.; Takalo, L.; Tavecchio, F.; Temnikov, P.; Terzić, T.; Teshima, M.; Torres-Albà, N.; Treves, A.; Tsujimoto, S.; Vanzo, G.; Vazquez Acosta, M.; Vovk, I.; Ward, J. E.; Will, M.; Zarić, D.

2018-03-01

The dwarf spheroidal galaxy Ursa Major II (UMaII) is believed to be one of the most dark-matter dominated systems among the Milky Way satellites and represents a suitable target for indirect dark matter (DM) searches. The MAGIC telescopes carried out a deep observation campaign on UMaII between 2014 and 2016, collecting almost one hundred hours of good-quality data. This campaign enlarges the pool of DM targets observed at very high energy (E gtrsim 50 GeV) in search for signatures of DM annihilation in the wide mass range between ~100 GeV and ~100 TeV. To this end, the data are analyzed with the full likelihood analysis, a method based on the exploitation of the spectral information of the recorded events for an optimal sensitivity to the explored DM models. We obtain constraints on the annihilation cross-section for different channels that are among the most robust and stringent achieved so far at the TeV mass scale from observations of dwarf satellite galaxies.

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

International Nuclear Information System (INIS)

2001-02-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g., caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs; and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which involves the integration of

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

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.

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

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

Pharmacological activity of salvinorin A, the major component of Salvia divinorum.

Science.gov (United States)

Listos, Joanna; Merska, Alicja; Fidecka, Sylwia

2011-01-01

The hallucinogenic plant Salvia divinorum (i.e., "magic mint") is a member of the Sage family that has been historically used for divination and shamanism by the Mazatecs. Today, S. divinorum has become increasingly popular as a recreational drug for its hallucinogenic effects. The non-nitrogenous diterpene, salvinorin A, the major active component of S. divinorum, is responsible for the hallucinogenic effect of this plant. Here, we described the behavioral effects of salvinorin A in animals including the addictive, antinociception and antidepressant properties of the drug. The present paper also demonstrates the not well recognized (or unclear) mechanisms of action of salvinorin A. The last part of the paper presents information about the legal status of S. divinorum and its derivatives. Taking into account the increasing popularity and consumption of salvinorin A and S. divinorum today, it is important to collect all data on the pharmacological profile of this plant and its products.

Assessment and management of ageing of major nuclear power plant components important to safety: Steam generators

International Nuclear Information System (INIS)

1997-11-01

This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. They have been compiled using contributions from technical experts in typically 10 to 12 countries for each report, a feedback from a September 1994 Technical Committee Meeting attended by 53 technical experts from 21 Member States (who reviewed first drafts in specialized working groups), and review comments from invited specialists

Pulmonary surfactant and its components inhibit secretion of phosphatidylcholine from cultured rat alveolar type II cells

International Nuclear Information System (INIS)

Dobbs, L.G.; Wright, J.R.; Hawgood, S.; Gonzalez, R.; Venstrom, K.; Nellenbogen, J.

1987-01-01

Pulmonary surfactant is synthesized and secreted by alveolar type II cells. Radioactive phosphatidylcholine has been used as a marker for surfactant secretion. The authors report findings that suggest that surfactant inhibits secretion of 3 H-labeled phosphatidylcholine by cultured rat type II cells. The lipid components and the surfactant protein group of M/sub r/ 26,000-36,000 (SP 26-36) inhibit secretion to different extents. Surfactant lipids do not completely inhibit release; in concentrations of 100 μg/ml, lipids inhibit stimulated secretion by 40%. SP 26-36 inhibits release with an EC 50 of 0.1 μg/ml. At concentrations of 1.0 μg/ml, SP 26-36 inhibits basal secretion and reduces to basal levels secretion stimulated by terbutaline, phorbol 12-myristate 13-acetate, and the ionophore A23187. The inhibitory effect of SP 26-36 can be blocked by washing type II cells after adding SP 26-36, by heating the proteins to 100 0 C for 10 min, by adding antiserum specific to SP 26-36, or by incubating cells in the presence of 0.2 mM EGTA. SP 26-36 isolated from canine and human sources also inhibits phosphatidylcholine release from rat type II cells. Neither type I collagen nor serum apolipoprotein A-1 inhibits secretion. These findings are compatible with the hypothesis that surfactant secretion is under feedback regulatory control

THE DISTRIBUTION OF ORGANIC CARBON IN MAJOR COMPONENTS OF FORESTS LOCATED IN FIVE LIFE ZONES OF VENEZUELA

Science.gov (United States)

One of the major uncertainties concerning the role of tropical forests in the global carbon cycle is the lack of adequate data on the carbon content of all their components. The goal of this study was to contribute to filling this data gap by estimating the quantity of carbon in ...

RELAP5 based engineering simulator

International Nuclear Information System (INIS)

Charlton, T.R.; Laats, E.T.; Burtt, J.D.

1990-01-01

The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nuclear Plant Analyzer (NPA), have recently been improved significantly. This paper presents an overview of the INEL Engineering Simulation Center, and discusses the RELAP5/MOD3 and NPA/MOD1 codes, specifically how they are being used at the INEL Engineering Simulation Center. It provides an update on the modifications to these two codes and their application to the ATR engineering simulator project, as well as, a discussion on the reactor system representation, control system modeling, two phase flow and heat transfer modeling. It will also discuss how these two codes are providing desktop, stand-alone reactor simulation

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

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

Microwave Assisted Expeditious and Green Cu(II-Clay Catalyzed Domino One-Pot Three Component Synthesis of 2H-indazoles

Directory of Open Access Journals (Sweden)

Bashir Ahmad Dar

2018-01-01

How to Cite: Dar, B.A., Safvi, S.W., Rizvi, M.A. (2018. Microwave Assisted Expeditious and Green Cu(II-Clay Catalyzed Domino One-Pot Three Component Synthesis of 2H-indazoles. Bulletin of Chemical Reaction Engineering & Catalysis, 13 (1: 82-88 (doi:10.9767/bcrec.13.1.963.82-88

Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

International Nuclear Information System (INIS)

Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

2017-01-01

In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

A procedure for evaluating residual life of major components in light water reactors

International Nuclear Information System (INIS)

Uchida, S.; Fujimori, H.; Ibe, E.; Kuniya, J.; Hayashi, M.; Fuse, M.; Yamauchi, K.

1995-01-01

A computer program for evaluating residual life of major components in boiling water reactors is proposed. It divides the stress corrosion cracking process into two stages; a probabilistic crack generation stage and a deterministic crack propagation one. The minimum period of the crack generation stage is evaluated assuming an exponential distribution of the stage. The crack propagation rate is calculated by the slip-dissolution/film-rupture model. The neutron flux and fluence dependence of the neutron radiation effects on material properties was evaluated by using theoretical models of radiation damage. The computer program works on an engineering work station. Evaluated results are displayed as a map of the residual life, or as graphs of crack length evolution

Contribution from the United States of America

International Nuclear Information System (INIS)

Finck, Ph.

2000-01-01

It is expected that replacement of UO 2 fuel by MOX fuel in the core of BN type reactor would not be a major difficulty for standard Sn codes and ENDF data library. Replacement of the blanket by stainless steel would cause some difficulties although experience exists with ZPR-3, EBR-II and CIRANO reactors. Major difficulties that are expected are concerned with data and methods (high order 3-D transport calculations with anisotropic scattering). The effects on transients would not necessarily be large, as effects are local. A calculational scheme which would enable easy intercomparison includes US and European codes for both ENDFB-V, ENDFB-VI and JEF2 data processing; neutron calculations are to be compared to Monte Carlo calculations used as benchmark

Fuel-sodium reaction product formation in breached mixed-oxide fuel

International Nuclear Information System (INIS)

Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

1988-01-01

The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

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...

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.

Distribution of class ii major histocompatibility complex antigenexpressing cells in human dental pulp with carious lesions

Directory of Open Access Journals (Sweden)

Tetiana Haniastuti

2012-09-01

Full Text Available Background: Dental caries is a bacterial infection which causes destruction of the hard tissues of the tooth. Exposure of the dentin to the oral environment as a result of caries inevitably results in a cellular response in the pulp. The major histocompatibility complex (MHC is a group of genes that code for cell-surface histocompatibility antigens. Cells expressing class II MHC molecules participate in the initial recognition and the processing of antigenic substances to serve as antigen-presenting cells. Purpose: The aim of the study was to elucidate the alteration in the distribution of class II MHC antigen-expressing cells in human dental pulp as carious lesions progressed toward the pulp. Methods: Fifteen third molars with caries at the occlusal site at various stages of decay and 5 intact third molars were extracted and used in this study. Before decalcifying with 10% EDTA solution (pH 7.4, all the samples were observed by micro-computed tomography to confirm the lesion condition three-dimensionally. The specimens were then processed for cryosection and immunohistochemistry using an anti-MHC class II monoclonal antibody. Results: Class II MHC antigen-expressing cells were found both in normal and carious specimens. In normal tooth, the class II MHC-immunopositive cells were observed mainly at the periphery of the pulp tissue. In teeth with caries, class II MHC-immunopositive cells were located predominantly subjacent to the carious lesions. As the caries progressed, the number of class II MHC antigen-expressing cells was increased. Conclusion: The depth of carious lesions affects the distribution of class II MHC antigen-expressing cells in the dental pulp.Latar belakang: Karies merupakan penyakit infeksi bakteri yang mengakibatkan destruksi jaringan keras gigi. Dentin yang terbuka akibat karies akan menginduksi respon imun seluler pada pulpa. Kompleks histokompatibilitas utama (MHC merupakan sekumpulan gen yang mengkode histokompatibilitas

Angiotensin II Type 1 receptor (AT1) signaling in astrocytes regulates synaptic degeneration-induced leukocyte entry to the central nervous system

DEFF Research Database (Denmark)

Füchtbauer, L; Groth-Rasmussen, Maria; Holm, Thomas Hellesøe

2011-01-01

Astrocytes are the major cellular component of the blood-brain barrier glia limitans and act as regulators of leukocyte infiltration via chemokine expression. We have studied angiotensin-II receptor Type 1 (AT1) and related NF-κB signaling in astrocytes. Angiotensin II derives from cleavage of an...

Development of and verification test integral reactor major components - Development of manufacturing process and fabrication of prototype for SG and CEDM

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Hee; Park, Hwa Kyu; Kim, Yong Kyu; Choi, Yong Soon; Kang, Ki Su; Hyun, Young Min [Korea Heavy Industries and Construction Co., LTD., Changwon (Korea)

1999-03-01

Integral SMART(System integrated Modular Advanced Reactor) type reactor is under conceptual design. Because major components is integrated within in a single pressure vessel, compact design using advanced technology is essential. It means that manufacturing process for these components is more complex and difficult. The objective of this study is to confirm the possibility of manufacture of Steam Generator, Control Element Drive Mechanism(CEDM) and Reactor Assembly which includes Reactor Pressure Vessel, it is important to understand the design requirement and function of the major components. After understanding the design requirement and function, it is concluded that the helical bending and weld qualification of titanium tube for Steam Generator and the applicability of electron beam weld for CEDM step motor parts is the critical to fabricate the components. Therefore, bending mock-up and weld qualification of titanium tube was performed and the results are quite satisfactory. Also, it is concluded that electron beam welding technique can be applicable to the CEDM step motor part. (author). 22 refs., 14 figs., 46 tabs.

Biosphere II: engineering of manned, closed ecological systems.

Science.gov (United States)

Dempster, W F

1991-01-01

Space Biospheres and Ventures, a private, for-profit firm, has undertaken a major research and development project in the study of biospheres, with the objective of creating and producing biospheres. Biosphere II-scheduled for completion in March 1991-will be essentially isolated from the existing biosphere by a closed structure, composed of components derived from the existing biosphere. Like the biosphere of the Earth, Biosphere II will be essentially closed to exchanges of material or living organisms with the surrounding environment and open to energy and information exchanges. Also, like the biosphere of the Earth, Biosphere II will contain five kingdoms of life, a variety of ecosystems, plus humankind, culture, and technics. The system is designed to be complex, stable and evolving throughout its intended 100-year lifespan, rather than static. Biosphere II will cover approximately 1.3 hectare and contain 200,000 m3 in volume, with seven major biomes: tropical rainforest, tropical savannah, marsh, marine, desert, intensive agriculture, and human habitat. An interdisciplinary team of leading scientific, ecological, management, architectural, and engineering consultants have been contracted by Space Biospheres Ventures for the project. Potential applications for biospheric systems include scientific and ecological management research, refuges for endangered species, and life habitats for manned stations on spacecraft or other planets.

Major component analysis of dynamic networks of physiologic organ interactions

International Nuclear Information System (INIS)

Liu, Kang K L; Ma, Qianli D Y; Ivanov, Plamen Ch; Bartsch, Ronny P

2015-01-01

The human organism is a complex network of interconnected organ systems, where the behavior of one system affects the dynamics of other systems. Identifying and quantifying dynamical networks of diverse physiologic systems under varied conditions is a challenge due to the complexity in the output dynamics of the individual systems and the transient and nonlinear characteristics of their coupling. We introduce a novel computational method based on the concept of time delay stability and major component analysis to investigate how organ systems interact as a network to coordinate their functions. We analyze a large database of continuously recorded multi-channel physiologic signals from healthy young subjects during night-time sleep. We identify a network of dynamic interactions between key physiologic systems in the human organism. Further, we find that each physiologic state is characterized by a distinct network structure with different relative contribution from individual organ systems to the global network dynamics. Specifically, we observe a gradual decrease in the strength of coupling of heart and respiration to the rest of the network with transition from wake to deep sleep, and in contrast, an increased relative contribution to network dynamics from chin and leg muscle tone and eye movement, demonstrating a robust association between network topology and physiologic function. (paper)

Component fragility research program

International Nuclear Information System (INIS)

Tsai, N.C.; Mochizuki, G.L.; Holman, G.S.

1989-11-01

To demonstrate how ''high-level'' qualification test data can be used to estimate the ultimate seismic capacity of nuclear power plant equipment, we assessed in detail various electrical components tested by the Pacific Gas ampersand Electric Company for its Diablo Canyon plant. As part of our Phase I Component Fragility Research Program, we evaluated seismic fragility for five Diablo Canyon components: medium-voltage (4kV) switchgear; safeguard relay board; emergency light battery pack; potential transformer; and station battery and racks. This report discusses our Phase II fragility evaluation of a single Westinghouse Type W motor control center column, a fan cooler motor controller, and three local starters at the Diablo Canyon nuclear power plant. These components were seismically qualified by means of biaxial random motion tests on a shaker table, and the test response spectra formed the basis for the estimate of the seismic capacity of the components. The seismic capacity of each component is referenced to the zero period acceleration (ZPA) and, in our Phase II study only, to the average spectral acceleration (ASA) of the motion at its base. For the motor control center, the seismic capacity was compared to the capacity of a Westinghouse Five-Star MCC subjected to actual fragility tests by LLNL during the Phase I Component Fragility Research Program, and to generic capacities developed by the Brookhaven National Laboratory for motor control center. Except for the medium-voltage switchgear, all of the components considered in both our Phase I and Phase II evaluations were qualified in their standard commercial configurations or with only relatively minor modifications such as top bracing of cabinets. 8 refs., 67 figs., 7 tabs

Work disability following major organisational change: the Whitehall II study.

Science.gov (United States)

Virtanen, M; Kivimäki, M; Singh-Manoux, A; Gimeno, D; Shipley, M J; Vahtera, J; Akbaraly, T N; Marmot, M G; Ferrie, J E

2010-05-01

Privatisation and private sector practices have been increasingly applied to the public sector in many industrialised countries. Over the same period, long-term work disability has risen substantially. We examined whether a major organisational change--the transfer of public sector work to executive agencies run on private sector lines--was associated with an increased risk of work disability. The study uses self-reported data from the prospective Whitehall II cohort study. Associations between transfer to an executive agency assessed at baseline (1991-1994) and work disability ascertained over a period of approximately 8 years at three follow-up surveys (1995-1996, 1997-1999 and 2001) were examined using Cox proportional hazard models. In age- and sex-adjusted models, risk of work disability was higher among the 1263 employees who were transferred to an executive agency (HR 1.90, 95% CI 1.46 to 2.48) compared with the 3419 employees whose job was not transferred. These findings were robust to additional adjustment for physical and mental health and health behaviours at baseline. Increased work disability was observed among employees exposed to the transfer of public sector work to executive agencies run on private sector lines. This may highlight an unintentional cost for employees, employers and society.

Gallic Acid Is the Major Active Component of Cortex Moutan in Inhibiting Immune Maturation of Human Monocyte-Derived Dendritic Cells

Directory of Open Access Journals (Sweden)

Ben Chung Lap Chan

2015-09-01

Full Text Available Atopic dermatitis (AD is a widely prevalent and chronically relapsing inflammatory skin disease. Penta Herbs Formula (PHF is efficacious in improving the quality of life and reducing topical corticosteroid used in children with AD and one of the active herbs it contains is Cortex Moutan. Recent studies showed that altered functions of dendritic cells (DC were observed in atopic individuals, suggesting that DC might play a major role in the generation and maintenance of inflammation by their production of pro-inflammatory cytokines. Hence, the aims of the present study were to identify the major active component(s of Cortex Moutan, which might inhibit DC functions and to investigate their possible interactions with conventional corticosteroid on inhibiting the development of DC from monocytes. Monocyte-derived dendritic cells (moDC culture model coupled with the high-speed counter-current chromatography (HSCCC, high pressure liquid chromatography (HPLC and Liquid Chromatography-Mass Spectrometry (LCMS analyses were used. Gallic acid was the major active component from Cortex Moutan which could dose dependently inhibit interleukin (IL-12 p40 and the functional cluster of differentiation (CD surface markers CD40, CD80, CD83 and CD86 expression from cytokine cocktail-activated moDC. Gallic acid could also lower the concentration of hydrocortisone required to inhibit the activation of DC.

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and ware out of components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of heavy water moderated reactors (HWRs), boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

Giant panda genomic data provide insight into the birth-and-death process of mammalian major histocompatibility complex class II genes.

Directory of Open Access Journals (Sweden)

Qiu-Hong Wan

Full Text Available To gain an understanding of the genomic structure and evolutionary history of the giant panda major histocompatibility complex (MHC genes, we determined a 636,503-bp nucleotide sequence spanning the MHC class II region. Analysis revealed that the MHC class II region from this rare species contained 26 loci (17 predicted to be expressed, of which 10 are classical class II genes (1 DRA, 2 DRB, 2 DQA, 3 DQB, 1 DYB, 1 DPA, and 2 DPB and 4 are non-classical class II genes (1 DOA, 1 DOB, 1 DMA, and 1 DMB. The presence of DYB, a gene specific to ruminants, prompted a comparison of the giant panda class II sequence with those of humans, cats, dogs, cattle, pigs, and mice. The results indicated that birth and death events within the DQ and DRB-DY regions led to major lineage differences, with absence of these regions in the cat and in humans and mice respectively. The phylogenetic trees constructed using all expressed alpha and beta genes from marsupials and placental mammals showed that: (1 because marsupials carry loci corresponding to DR, DP, DO and DM genes, those subregions most likely developed before the divergence of marsupials and placental mammals, approximately 150 million years ago (MYA; (2 conversely, the DQ and DY regions must have evolved later, but before the radiation of placental mammals (100 MYA. As a result, the typical genomic structure of MHC class II genes for the giant panda is similar to that of the other placental mammals and corresponds to BTNL2 approximately DR1 approximately DQ approximately DR2 approximately DY approximately DO_box approximately DP approximately COL11A2. Over the past 100 million years, there has been birth and death of mammalian DR, DQ, DY, and DP genes, an evolutionary process that has brought about the current species-specific genomic structure of the MHC class II region. Furthermore, facing certain similar pathogens, mammals have adopted intra-subregion (DR and DQ and inter-subregion (between DQ and DP

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

One-pot four-component synthesis of 2-aryl-3,3-dihaloacrylonitriles using potassium hexacyanoferrate(II) as environmentally benign cyanide source

International Nuclear Information System (INIS)

Zhao, Zhouxing; Li, Zheng

2011-01-01

An efficient route to one-pot four-component reactions of aroyl chlorides, potassium hexacyanoferrate(II), triphenylphosphine and carbon tetrahalides to synthesize 2-aryl-3,3-dichloroacrylonitriles and 2-aryl-3,3-dibromoacrylonitriles was described. This protocol has advantages of use of non-toxic cyanide source, high yield and simple work-up procedure. (author)

Altered expression of γ-secretase components in animal model of major depressive disorder induced by reserpine administration.

Science.gov (United States)

Lee, Hye-Ryun; Hwang, In-Sik; Kim, Ji-Eun; Choi, Sun-Il; Lee, Young-Ju; Goo, Jun-Seo; Lee, Eon-Pil; Choi, Hae-Wook; Kim, Hong-Sung; Lee, Jae-Ho; Jung, Young-Jin; Hwang, Dae-Youn

2012-06-01

Altered expression of neurotrophic factors as well as neuroinflammation is commonly associated with Major depressive disorder (MDD) and Alzheimer's disease (AD). To investigate whether or not reserpine-induced MDD affects the expression of AD-related proteins, the expression of γ-secretase components and substrate were measured in brains of ICR mice following reserpine treatment for 15 days. In active avoidance test, total response time and peak slightly increased in the 2 mg/kg reserpine (RSP2)-treated group compared to vehicle-treated group (Pprecursor protein (APP) was enhanced in the RSP1 and RSP2-treated groups compared to the vehicle-treated group, whereas expression of γ-secretase components decreased (P<0.03). Among the three components of the γ-secretase complex, nicastrin protein underwent the largest decrease in expression, as detected by Western blotting (P<0.03). Therefore, the data presented here provide additional evidence about the pathological correlation between MDD and AD.

Relationships of neuroticism and extraversion with axis I and II comorbidity among patients with DSM-IV major depressive disorder.

Science.gov (United States)

Jylhä, Pekka; Melartin, Tarja; Isometsä, Erkki

2009-04-01

High comorbidity with axis I and II disorders among major depressive disorder (MDD) patients may in part be due to the predisposing personality dimensions of neuroticism and extraversion. However, a comprehensive view of this relationship is lacking. MDD patients (n=193) in the Vantaa Depression Study were interviewed at baseline and at 6 and 18 months with the SCAN and SCID-II, and a general population comparison group (n=388) surveyed by mail. Neuroticism and extraversion were measured with the Eysenck Personality Inventory. A dose-exposure relationship between standardized levels of neuroticism and extraversion and type and number of comorbid axis I and II disorders among patients with MDD was hypothesized. Prevalence and number of comorbid axis I and II disorders increased significantly with increasing level of neuroticism. In contrast, as the level of extraversion increased, the prevalences of social phobia and cluster C personality disorders decreased. Patients with pure MDD or with any comorbid axis I or II disorder had z-scores of neuroticism of +0.46, +0.90 and +1.30 and of extraversion of -0.34, -0.47 and -0.84, respectively. Patients' personality scores were not pre-morbid. Among MDD patients, a positive dose-exposure relationship appears to exist between neuroticism and prevalence and number of comorbid axis I and II disorders. A negative relationship exists between level of extraversion and prevalence of social phobia and cluster C personality disorders. These findings are consistent with the hypothesis that high neuroticism and low extraversion predispose to comorbid axis I and II disorders among patients with MDD.

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

Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

International Nuclear Information System (INIS)

Sudardjo; Histori; Triyadi, Ari

1998-01-01

The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

Development of technology on natural flaw fabrication and precise diagnosis for the major components in NPPs

International Nuclear Information System (INIS)

Han, Jung Ho; Choi, Myung Sik; Lee, Doek Hyun; Hur, Do Haeng

2002-01-01

The objective of this research is to develop a fabrication technology of natural flaw specimen of major components in NPPs and a technology of precise diagnosis for failure and degradation of components using natural flaw specimen. 1) Successful development of the natural flaw fabrication technology of SG tube 2) Evaluation of ECT signal and development of precise diagnosis using natural flaws. - Determination of length, depth, width, and multiplicity of fabricated natural flaws. - Informations about detectability and accuracy of ECT evaluation on various kinds of defects are collected when the combination of probe and frequency is changed. - An advanced technology for precise ECT evaluation is established. 3) Application of precise ECT diagnosis to failure analysis of SG tube in operation. - Fretting wear of KSNP SG. - ODSCC at tube expanded region of KSNP SG. - Determination of through/non-through wall of axial crack

Suicide attempts in major depressive episode: evidence from the BRIDGE-II-Mix study.

Science.gov (United States)

Popovic, Dina; Vieta, Eduard; Azorin, Jean-Michel; Angst, Jules; Bowden, Charles L; Mosolov, Sergey; Young, Allan H; Perugi, Giulio

2015-11-01

The Bipolar Disorders: Improving Diagnosis, Guidance, and Education (BRIDGE-II-Mix) study aimed to estimate the frequency of mixed states in patients with a major depressive episode (MDE) according to different definitions and to compare their clinical validity, looking into specific features such as suicidality. A total of 2,811 subjects were enrolled in this multicenter cross-sectional study. Psychiatric symptoms, and sociodemographic and clinical variables were collected. The analysis compared the characteristics of patients with MDE with (MDE-SA group) and without (MDE-NSA) a history of suicide attempts. The history of suicide attempts was registered in 628 patients (22.34%). In the MDE-SA group, women (72.5%, p = 0.028), (hypo)mania in first-degree relatives (20.5%, p suicide attempts. In the MDE-SA group, 75 patients (11.9%) fulfilled Diagnostic and Statistical Manual (DSM)-5 criteria for MDE with mixed features, and 250 patients (39.8%) fulfilled research-based diagnostic criteria for a mixed depressive episode. Important differences between MDE-SA and MDE-NSA patients have emerged. Early identification of symptoms such as risky behavior, psychomotor agitation, and impulsivity in patients with MDE, and treatment of mixed depressive states could represent a major step in suicide prevention. © 2015 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

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

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.

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

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

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

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.

An examination of reliability critical items in liquid metal reactors: An analysis by the Centralized Reliability Data Organization (CREDO)

International Nuclear Information System (INIS)

Humphrys, B.L.; Haire, M.J.; Koger, K.H.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.; Okubo, Y.

1987-01-01

The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 21,000 components and approximately 1300 event records. A conservative estimation is that the total component operating hours is approaching 3.5 billion hours. Because data gathering for CREDO concentrates on event (failure) information, the work reported here focuses on the reliability information contained in CREDO and the development of reliability critical items lists. That is, components are ranked in prioritized lists from worst to best performers from a reliability standpoint. For the data contained in the CREDO data base, FFTF and JOYO show reliability growth; EBR-II reveals a slight unreliability growth for those components tracked by CREDO. However, tabulations of events which cause reactor shutdowns decrease with time at each site

Characterization of secondary organic aerosol from photo-oxidation of gasoline exhaust and specific sources of major components.

Science.gov (United States)

Ma, Pengkun; Zhang, Peng; Shu, Jinian; Yang, Bo; Zhang, Haixu

2018-01-01

To further explore the composition and distribution of secondary organic aerosol (SOA) components from the photo-oxidation of light aromatic precursors (toluene, m-xylene, and 1,3,5-trimethylbenzene (1,3,5-TMB)) and idling gasoline exhaust, a vacuum ultraviolet photoionization mass spectrometer (VUV-PIMS) was employed. Peaks of the molecular ions of the SOA components with minimum molecular fragmentation were clearly observed from the mass spectra of SOA, through the application of soft ionization methods in VUV-PIMS. The experiments comparing the exhaust-SOA and light aromatic mixture-SOA showed that the observed distributions of almost all the predominant cluster ions in the exhaust-SOA were similar to that of the mixture-SOA. Based on the characterization experiments of SOA formed from individual light aromatic precursors, the SOA components with molecular weights of 98 and 110 amu observed in the exhaust-SOA resulted from the photo-oxidation of toluene and m-xylene; the components with a molecular weight of 124 amu were derived mainly from m-xylene; and the components with molecular weights of 100, 112, 128, 138, and 156 amu were mainly derived from 1,3,5-TMB. These results suggest that C 7 -C 9 light aromatic hydrocarbons are significant SOA precursors and that major SOA components originate from gasoline exhaust. Additionally, some new light aromatic hydrocarbon-SOA components were observed for the first time using VUV-PIMS. The corresponding reaction mechanisms were also proposed in this study to enrich the knowledge base of the formation mechanisms of light aromatic hydrocarbon-SOA compounds. Copyright © 2017 Elsevier Ltd. All rights reserved.

Metabolic syndrome definitions and components in predicting major adverse cardiovascular events after kidney transplantation.

Science.gov (United States)

Prasad, G V Ramesh; Huang, Michael; Silver, Samuel A; Al-Lawati, Ali I; Rapi, Lindita; Nash, Michelle M; Zaltzman, Jeffrey S

2015-01-01

Metabolic syndrome (MetS) associates with cardiovascular risk post-kidney transplantation, but its ambiguity impairs understanding of its diagnostic utility relative to components. We compared five MetS definitions and the predictive value of constituent components of significant definitions for major adverse cardiovascular events (MACE) in a cohort of 1182 kidney transplant recipients. MetS definitions were adjusted for noncomponent traditional Framingham risk factors and relevant transplant-related variables. Kaplan-Meier, logistic regression, and Cox proportional hazards analysis were utilized. There were 143 MACE over 7447 patient-years of follow-up. Only the World Health Organization (WHO) 1998 definition predicted MACE (25.3 vs 15.5 events/1000 patient-years, P = 0.019). Time-to-MACE was 5.5 ± 3.5 years with MetS and 6.8 ± 3.9 years without MetS (P < 0.0001). MetS was independent of pertinent MACE risk factors except age and previous cardiac disease. Among MetS components, dysglycemia provided greatest hazard ratio (HR) for MACE (1.814 [95% confidence interval 1.26-2.60]), increased successively by microalbuminuria (HR 1.946 [1.37-2.75]), dyslipidemia (3.284 [1.72-6.26]), hypertension (4.127 [2.16-7.86]), and central obesity (4.282 [2.09-8.76]). MetS did not affect graft survival. In summary, although the WHO 1998 definition provides greatest predictive value for post-transplant MACE, most of this is conferred by dysglycemia and is overshadowed by age and previous cardiac disease. © 2014 Steunstichting ESOT.

Statistical downscaling of historical monthly mean winds over a coastal region of complex terrain. II. Predicting wind components

Energy Technology Data Exchange (ETDEWEB)

Kamp, Derek van der [University of Victoria, Pacific Climate Impacts Consortium, Victoria, BC (Canada); University of Victoria, School of Earth and Ocean Sciences, Victoria, BC (Canada); Curry, Charles L. [Environment Canada University of Victoria, Canadian Centre for Climate Modelling and Analysis, Victoria, BC (Canada); University of Victoria, School of Earth and Ocean Sciences, Victoria, BC (Canada); Monahan, Adam H. [University of Victoria, School of Earth and Ocean Sciences, Victoria, BC (Canada)

2012-04-15

A regression-based downscaling technique was applied to monthly mean surface wind observations from stations throughout western Canada as well as from buoys in the Northeast Pacific Ocean over the period 1979-2006. A predictor set was developed from principal component analysis of the three wind components at 500 hPa and mean sea-level pressure taken from the NCEP Reanalysis II. Building on the results of a companion paper, Curry et al. (Clim Dyn 2011), the downscaling was applied to both wind speed and wind components, in an effort to evaluate the utility of each type of predictand. Cross-validated prediction skill varied strongly with season, with autumn and summer displaying the highest and lowest skill, respectively. In most cases wind components were predicted with better skill than wind speeds. The predictive ability of wind components was found to be strongly related to their orientation. Wind components with the best predictions were often oriented along topographically significant features such as constricted valleys, mountain ranges or ocean channels. This influence of directionality on predictive ability is most prominent during autumn and winter at inland sites with complex topography. Stations in regions with relatively flat terrain (where topographic steering is minimal) exhibit inter-station consistencies including region-wide seasonal shifts in the direction of the best predicted wind component. The conclusion that wind components can be skillfully predicted only over a limited range of directions at most stations limits the scope of statistically downscaled wind speed predictions. It seems likely that such limitations apply to other regions of complex terrain as well. (orig.)

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

Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

International Nuclear Information System (INIS)

2003-07-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. The

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

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.

Comparative study on major bioactive components in natural, artificial and in-vitro cultured Calculus Bovis.

Science.gov (United States)

Yan, Shi-Kai; Wu, Yan-Wen; Liu, Run-Hui; Zhang, Wei-Dong

2007-01-01

Major bioactive components in various Calculus Bovis, including natural, artificial and in-vitro cultured Calculus Bovis, were comparatively studied. An approach of high-performance liquid chromatography coupled with ultraviolet and evaporative light scattering detections (HPLC/UV/ELSD) was established to simultaneously determinate six bioactive components thereof, including five bile acids (cholic acid, deoxycholic acid, ursodeoxycholic, chenodeoxycholic acid, hyodeoxycholic acid) and bilirubin. ELSD and UV detector were applied to detect bile acids and bilirubin respectively. The assay was performed on a C(18) column with water-acetonitrile gradient elution and the investigated constituents were authenticated by comparing retention times and mass spectra with those of reference compounds. The proposed method was applied to analyze twenty-one Calculus Bovis extraction samples, and produced data with acceptable linearity, precision, repeatability and accuracy. The result indicated the variations among Calculus Bovis samples under different developmental conditions. Artificial and in-vitro cultured Calculus Bovis, especially in-vitro cultured ones, which contain total bioactive constituents no less than natural products and have the best batch-to-batch uniformity, suffice to be used as substitutes of natural Calculus Bovis.

Emotional Component in Teaching and Learning

Science.gov (United States)

Ponnambalam, Michael

2018-02-01

The laws of physics are often seen as objective truth, pure and simple. Hence, they tend to appear cerebral and cold. However, their presentation is necessarily subjective and may vary from being boring to being exciting. A detailed analysis of physics education reform efforts over the last three decades finds that interactive instruction results in greater learning gains than the traditional lecture format. In interactive engagement, the emotional component plays a far greater role than acknowledged by many. As an experienced physics teacher [(i) Four decades of teaching and research in four continents (teaching all courses to undergraduate physics majors and algebra-based physics to high school seniors as well as college freshmen), (ii) 11 years of volunteer work in Physics Popularization in six countries to many thousands of students in elementary, middle, and high schools as well as colleges and universities, and (iii) eight years as a Master Teacher and mentor], I feel that the emotional component in teaching and learning physics has been neglected. This paper presents the role of the emotional component in transforming ordinary teaching and learning of physics into an enjoyable and exciting experience for students as well as teachers.

Clinical, Immunological, and Molecular Findings in Five Patients with Major Histocompatibility Complex Class II Deficiency from India

Directory of Open Access Journals (Sweden)

Jahnavi Aluri

2018-02-01

Full Text Available Major histocompatibility complex (MHC class II deficiency is a rare autosomal recessive form of primary immunodeficiency disorder (PID characterized by the deficiency of MHC class II molecules. This deficiency affects the cellular and humoral immune response by impairing the development of CD4+ T helper (Th cells and Th cell-dependent antibody production by B cells. Affected children typically present with severe respiratory and gastrointestinal tract infections. Hematopoietic stem cell transplantation (HSCT is the only curative therapy available for treating these patients. This is the first report from India wherein we describe the clinical, immunological, and molecular findings in five patients with MHC class II deficiency. Our patients presented with recurrent lower respiratory tract infection as the most common clinical presentation within their first year of life and had a complete absence of human leukocyte antigen-antigen D-related (HLA-DR expression on B cells and monocytes. Molecular characterization revealed novel mutations in RFAXP, RFX5, and CIITA genes. Despite genetic heterogeneity, these patients were clinically indistinguishable. Two patients underwent HSCT but had a poor survival outcome. Detectable level of T cell receptor excision circles (TRECs were measured in our patients, highlighting that this form of PID may be missed by TREC-based newborn screening program for severe combined immunodeficiency.

Mechanical components: fabrication of major reactor structures

International Nuclear Information System (INIS)

Nicholson, S.

1985-01-01

The paper examines the validity of criticisms of quality assurance of mechanical plant and welded products within major reactor structures, taking into account experience gained on the AGR's. Various constructive recommendations are made aimed at furthering the objectives of quality assurance in the nuclear industry and making it more cost-effective. Current levels of quality related costs in the fabrication industry are provided as a basis for discussion. (U.K.)

Simple replacement of violaxanthin by zeaxanthin in LHC-II does not cause chlorophyll fluorescence quenching.

Science.gov (United States)

Dreuw, Andreas; Wormit, Michael

2008-03-01

Recently, a mechanism for the energy-dependent component (qE) of non-photochemical quenching (NPQ), the fundamental photo-protection mechanism in green plants, has been suggested. Replacement of violaxanthin by zeaxanthin in the binding pocket of the major light harvesting complex LHC-II may be sufficient to invoke efficient chlorophyll fluorescence quenching. Our quantum chemical calculations, however, show that the excited state energies of violaxanthin and zeaxanthin are practically identical when their geometry is constrained to the naturally observed structure of violaxanthin in LHC-II. Therefore, since violaxanthin does not quench LHC-II, zeaxanthin should not either. This theoretical finding is nicely in agreement with experimental results obtained by femtosecond spectroscopy on LHC-II complexes containing violaxanthin or zeaxanthin.

Antibodies against major histocompatibility complex class II antigens directly inhibit the growth of T cells infected with Theileria parva without affecting their state of activation

OpenAIRE

Eichhorn, M; Prospero, T D; Heussler, Volker; Dobbelaere, D A

1993-01-01

We have analyzed the effect of antibodies (Abs) directed against major histocompatibility complex (MHC) class II Abs on the proliferation of Theileria parva-infected (Tpi) T cells. Anti-MHC class II Abs exert a direct effect on Tpi T cells causing an acute block in their proliferation. The inhibition does not involve apoptosis and is also entirely reversible. The rapid arrest of DNA synthesis caused by anti- MHC class II Abs is not due to interference with the state of activation of the T cel...

Metal components analysis of metallothionein-III in the brain sections of metallothionein-I and metallothionein-II null mice exposed to mercury vapor with HPLC/ICP-MS

Energy Technology Data Exchange (ETDEWEB)

Kameo, Satomi; Nakai, Kunihiko; Kurokawa, Naoyuki; Satoh, Hiroshi [Tohoku University, Graduate School of Medicine, Aoba-ku, Sendai (Japan); Kanehisa, Tomokazu; Naganuma, Akira [Tohoku University, Graduate School of Pharmaceutical Sciences, Sendai (Japan)

2005-04-01

Mercury vapor is effectively absorbed via inhalation and easily passes through the blood-brain barrier; therefore, mercury poisoning with primarily central nervous system symptoms occurs. Metallothionein (MT) is a cysteine-rich metal-binding protein and plays a protective role in heavy-metal poisoning and it is associated with the metabolism of trace elements. Two MT isoforms, MT-I and MT-II, are expressed coordinately in all mammalian tissues, whereas MT-III is a brain-specific member of the MT family. MT-III binds zinc and copper physiologically and is seemed to have important neurophysiological and neuromodulatory functions. The MT functions and metal components of MTs in the brain after mercury vapor exposure are of much interest; however, until now they have not been fully examined. In this study, the influences of the lack of MT-I and MT-II on mercury accumulation in the brain and the changes of zinc and copper concentrations and metal components of MTs were examined after mercury vapor exposure by using MT-I, II null mice and 129/Sv (wild-type) mice as experimental animals. MT-I, II null mice and wild-type mice were exposed to mercury vapor or an air stream for 2 h and were killed 24 h later. The brain was dissected into the cerebral cortex, the cerebellum, and the hippocampus. The concentrations of mercury in each brain section were determined by cold vapor atomic absorption spectrometry. The concentrations of mercury, copper, and zinc in each brain section were determined by inductively coupled plasma mass spectrometry (ICP-MS). The mercury accumulated in brains after mercury vapor exposure for MT-I, II null mice and wild-type mice. The mercury levels of MT-I, II null mice in each brain section were significantly higher than those of wild-type mice after mercury vapor exposure. A significant change of zinc concentrations with the following mercury vapor exposure for MT-I, II null mice was observed only in the cerebellum analyzed by two-way analysis of

Status of Gerda Phase II

Energy Technology Data Exchange (ETDEWEB)

Wagner, Victoria [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Collaboration: GERDA-Collaboration

2016-07-01

The GERDA experiment is designed to search for neutrinoless double beta (0νββ) decay of {sup 76}Ge. In Phase I of the experiment a background index (BI) of 10{sup -2} cts/(keV.kg.yr) was reached. No signal has been found and a lower limit on the half-life of 2.1.10{sup 25} yr (at 90% C.L.) is extracted. The aim of Phase II is to double the Ge mass and further reduce the BI by an order of magnitude to explore half-lives of about 10{sup 26} yr. Thirty new Broad Energy Germanium (BEGe) detectors have been produced. These detectors are distinct for their improved energy resolution and enhanced pulse shape discrimination of signal from background events. Further background reduction will be reached by an active veto to read out argon scintillation light. The Phase II commissioning showed that two of the major background components, external γ-rays from {sup 214}Bi and {sup 208}Tl decays, can be suppressed up to two orders of magnitude. This talk presents the current status of the GERDA Phase II upgrade.

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.

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

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

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

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

Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

International Nuclear Information System (INIS)

Choi, Jae-Boong; Yeum, Seung-Won; Ko, Han-Ok; Kim, Young-Jin; Kim, Hong-Key; Choi, Young-Hwan; Park, Youn-Won

2010-01-01

Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.

Development of a web-based aging monitoring system for an integrity evaluation of the major components in a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae-Boong, E-mail: boong33@skku.ed [SAFE Research Centre, School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon, Kyonggi-do 440-746 (Korea, Republic of); Yeum, Seung-Won; Ko, Han-Ok; Kim, Young-Jin [SAFE Research Centre, School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon, Kyonggi-do 440-746 (Korea, Republic of); Kim, Hong-Key; Choi, Young-Hwan; Park, Youn-Won [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yusong-ku, Teajon 305-338 (Korea, Republic of)

2010-01-15

Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.

Melittin, a major peptide component of bee venom, and its conjugates in cancer therapy.

Science.gov (United States)

Rady, Islam; Siddiqui, Imtiaz A; Rady, Mohamad; Mukhtar, Hasan

2017-08-28

Melittin (MEL), a major peptide component of bee venom, is an attractive candidate for cancer therapy. This agent has shown a variety of anti-cancer effects in preclinical cell culture and animal model systems. Despite a convincing efficacy data against variety of cancers, its applicability to humans has met with challenges due to several issues including its non-specific cytotoxicity, degradation and hemolytic activity. Several optimization approaches including utilization of nanoparticle based delivery of MEL have been utilized to circumvent the issues. Here, we summarize the current understanding of the anticancer effects of bee venom and MEL on different kinds of cancers. Further, we also present the available information for the possible mechanism of action of bee venom and/or MEL. Copyright © 2017 Elsevier B.V. All rights reserved.

Stress-based fatigue assessment of major component in NPP using modified Green's function approach

International Nuclear Information System (INIS)

Ko, Han Ok; Jhung, Myung Jo; Choi, Jae Boong

2013-01-01

In this paper, the modified GFA which can consider temperature-dependent material properties is proposed by using a neural network (NN) and weight factor. To verify the modified GFA, thermal stresses by the proposed method are compared with those by FEM. Finally, pros and cons of the new method as well as technical findings from the assessment are discussed to show applicability of them. In this paper, the modified GFA considering temperature-dependent material properties is proposed by using NN and weight factor. To verify the proposed method, thermal stresses by the modified Green's function are compared with those by FEM and the results between two methods show a good agreement. Finally, it is anticipated that more precise fatigue evaluation is performed by using the proposed method. Recently, 434 nuclear reactors are being operated in the world. Among them, about 40% reactors are being operated beyond their design life or will be approaching their life. During the long term operation, various degradation mechanisms are occurred. Fatigue damage caused by alternating operational stresses in terms of temperature or pressure change is the one of important damage mechanisms in the nuclear power plants (NPPs). Although components important to safety were designed to withstand the fatigue damage, cumulative usage factor (CUF) at some locations can exceed the design limit beyond the design life. So, it is necessary to monitor the fatigue damage of major components during the long term operation. Researches on fatigue monitoring system (FMS) have been widely performed. In USA, the FatiguePro was developed by EPRI and was applied to the CE, WEC, B and W and GE type reactors. In Korea, the Kori unit 1 which started commercial operation in 1978 is being operated beyond its design life. At the stage of the license renewal, various plans for degradation mechanisms were established and reviewed. And, in case of fatigue damage, to monitor the fatigue damage of major components

Fire protection program evaluation of Argonne National Laboratory, West for the Department of Energy

International Nuclear Information System (INIS)

1984-01-01

A fire protection engineering survey was conducted of the Argonne National Laboratory, West Facility, near Idaho Falls, Idaho. This facility includes EBR-II, TREAT, ZPPR, and HFEF. The facility meets the improved risk criteria as set forth in DOE Order 5480.1, Chapter VII. Some recommendations are given

Breached-pin testing in the US

International Nuclear Information System (INIS)

Mahagin, D.E.; Lambert, J.D.B.

1981-04-01

Experience gained at EBR-II by the late 1970's from a significant number of failures in experimental fuel-pin irradiations forms the basis of a program directed towards the characterization of breached pins. The questions to be answered and the issues raised by further testing are discussed

Microstructure characterizaton of advanced oxide fuel

International Nuclear Information System (INIS)

Rasmussen, D.E.; Gerber, E.W.; McCord, R.B.

1977-01-01

Preirradiation porosity, grain size, and microcomposition characteristics are presented for selected advanced oxide (PuO 2 -UO 2 ) LMFBR developmental fuels fabricated for irradiation testing in EBR-II. Quantitative microscopy, electron microprobe analysis, and a recently developed quantitative autoradiographic technique are utilized to relate microstructure characteristics to fabrication parameters

Beam energy loss to parasitic modes in SPEAR II

International Nuclear Information System (INIS)

Allen, M.A.; Paterson, J.M.; Rees, J.R.; Wilson, P.B.

1975-01-01

The energy loss due to the excitation of parasitic modes in the SPEAR II rf cavities and vacuum chamber components was measured by observing the shift in synchronous phase angle as a function of circulating beam current and accelerating cavity voltage. The resulting parasitic mode loss resistance is 5 M OMEGA at a bunch length of 6.5 cm. The loss resistance varies with bunch length sigma/sub z/ approximately as exp(-0.3 sigma/sub z/). If the measured result is compared with reasonable theoretical predictions, it may be inferred that the major portion of the parasitic loss takes place in ring vacuum components rather than in the rf cavities. (auth)

A CK2 site is reversibly phosphorylated in the photosystem II subunit CP29.

Science.gov (United States)

Testi, M G; Croce, R; Polverino-De Laureto, P; Bassi, R

1996-12-16

Protein phosphorylation is a major mechanism in the regulation of protein function. In chloroplast thylakoids several photosystem II subunits, including the major antenna light-harvesting complex II and several core complex components, are reversibly phosphorylated depending on the redox state of the electron carriers. A previously unknown reversible phosphorylation event has recently been described on the CP29 subunit which leads to conformational changes and protection from cold stress (Bergantino, E., Dainese, P., Cerovic, Z. Sechi, S. and Bassi, R. (1995) J. Biol Chem. 270, 8474-8481). In this study, we have identified the phosphorylation site on the N-terminal, stroma-exposed domain, showing that it is located in a sequence not homologous to the other members of the Lhc family. The phosphorylated sequence is unique in chloroplast membranes since it meets the requirements for CK2 (casein kinase II) kinases. The possibility that this phosphorylation is involved in a signal transduction pathway is discussed.

Fusion Power Demonstrations I and II

International Nuclear Information System (INIS)

Doggett, J.N.

1985-01-01

In this report we present a summary of the first phase of the Fusion Power Demonstration (FPD) design study. During this first phase, we investigated two configurations, performed detailed studies of major components, and identified and examined critical issues. In addition to these design specific studies, we also assembled a mirror-systems computer code to help optimize future device designs. The two configurations that we have studied are based on the MARS magnet configuration and are labeled FPD-I and FPD-II. The FPD-I configuration employs the same magnet set used in the FY83 FPD study, whereas the FPD-II magnets are a new, much smaller set chosen to help reduce the capital cost of the system. As part of the FPD study, we also identified and explored issues critical to the construction of an Engineering Test Reactor (ETR). These issues involve subsystems or components, which because of their cost or state of technology can have a significant impact on our ability to meet FPD's mission requirements on the assumed schedule. General Dynamics and Grumman Aerospace studied two of these systems, the high-field choke coil and the halo pump/direct converter, in great detail and their findings are presented in this report

Cardiac resynchronization induces major structural and functional reverse remodeling in patients with New York Heart Association class I/II heart failure

DEFF Research Database (Denmark)

St John Sutton, Martin; Ghio, Stefano; Plappert, Ted

2009-01-01

BACKGROUND: Cardiac resynchronization therapy (CRT) improves LV structure, function, and clinical outcomes in New York Heart Association class III/IV heart failure with prolonged QRS. It is not known whether patients with New York Heart Association class I/II systolic heart failure exhibit left...... ventricular (LV) reverse remodeling with CRT or whether reverse remodeling is modified by the cause of heart failure. METHODS AND RESULTS: Six hundred ten patients with New York Heart Association class I/II heart failure, QRS duration > or =120 ms, LV end-diastolic dimension > or =55 mm, and LV ejection...... reduction in LV end-diastolic and end-systolic volume indexes and a 3-fold greater increase in LV ejection fraction in patients with nonischemic causes of heart failure. CONCLUSIONS: CRT in patients with New York Heart Association I/II resulted in major structural and functional reverse remodeling at 1 year...

Structural analysis of group II chitinase (ChtII) catalysis completes the puzzle of chitin hydrolysis in insects.

Science.gov (United States)

Chen, Wei; Qu, Mingbo; Zhou, Yong; Yang, Qing

2018-02-23

Chitin is a linear homopolymer of N -acetyl-β-d-glucosamines and a major structural component of insect cuticles. Chitin hydrolysis involves glycoside hydrolase family 18 (GH18) chitinases. In insects, chitin hydrolysis is essential for periodic shedding of the old cuticle ecdysis and proceeds via a pathway different from that in the well studied bacterial chitinolytic system. Group II chitinase (ChtII) is a widespread chitinolytic enzyme in insects and contains the greatest number of catalytic domains and chitin-binding domains among chitinases. In Lepidopterans, ChtII and two other chitinases, ChtI and Chi-h, are essential for chitin hydrolysis. Although ChtI and Chi-h have been well studied, the role of ChtII remains elusive. Here, we investigated the structure and enzymology of Of ChtII, a ChtII derived from the insect pest Ostrinia furnacalis We present the crystal structures of two catalytically active domains of Of ChtII, Of ChtII-C1 and Of ChtII-C2, both in unliganded form and complexed with chitooligosaccharide substrates. We found that Of ChtII-C1 and Of ChtII-C2 both possess long, deep substrate-binding clefts with endochitinase activities. Of ChtII exhibited structural characteristics within the substrate-binding cleft similar to those in Of Chi-h and Of ChtI. However, Of ChtII lacked structural elements favoring substrate binding beyond the active sites, including an extra wall structure present in Of Chi-h. Nevertheless, the numerous domains in Of ChtII may compensate for this difference; a truncation containing one catalytic domain and three chitin-binding modules ( Of ChtII-B4C1) displayed activity toward insoluble polymeric substrates that was higher than those of Of Chi-h and Of ChtI. Our observations provide the last piece of the puzzle of chitin hydrolysis in insects. © 2018 by The American Society for Biochemistry and Molecular Biology, Inc.

SAF-A forms a complex with BRG1 and both components are required for RNA polymerase II mediated transcription.

Directory of Open Access Journals (Sweden)

Dzeneta Vizlin-Hodzic

Full Text Available BACKGROUND: Scaffold attachment factor A (SAF-A participates in the regulation of gene expression by organizing chromatin into transcriptionally active domains and by interacting directly with RNA polymerase II. METHODOLOGY: Here we use co-localization, co-immunoprecipitation (co-IP and in situ proximity ligation assay (PLA to identify Brahma Related Gene 1 (BRG1, the ATP-driven motor of the human SWI-SNF chromatin remodeling complex, as another SAF-A interaction partner in mouse embryonic stem (mES cells. We also employ RNA interference to investigate functional aspects of the SAF-A/BRG1 interaction. PRINCIPAL FINDINGS: We find that endogenous SAF-A protein interacts with endogenous BRG1 protein in mES cells, and that the interaction does not solely depend on the presence of mRNA. Moreover the interaction remains intact when cells are induced to differentiate. Functional analyses reveal that dual depletion of SAF-A and BRG1 abolishes global transcription by RNA polymerase II, while the nucleolar RNA polymerase I transcription machinery remains unaffected. CONCLUSIONS: We demonstrate that SAF-A interacts with BRG1 and that both components are required for RNA Polymerase II Mediated Transcription.

MHC class II restricted innate-like double negative T cells contribute to optimal primary and secondary immunity to Leishmania major.

Science.gov (United States)

Mou, Zhirong; Liu, Dong; Okwor, Ifeoma; Jia, Ping; Orihara, Kanami; Uzonna, Jude Ezeh

2014-09-01

Although it is generally believed that CD4(+) T cells play important roles in anti-Leishmania immunity, some studies suggest that they may be dispensable, and that MHC II-restricted CD3(+)CD4(-)CD8(-) (double negative, DN) T cells may be more important in regulating primary anti-Leishmania immunity. In addition, while there are reports of increased numbers of DN T cells in Leishmania-infected patients, dogs and mice, concrete evidence implicating these cells in secondary anti-Leishmania immunity has not yet been documented. Here, we report that DN T cells extensively proliferate and produce effector cytokines (IFN-γ, TNF and IL-17) and granzyme B (GrzB) in the draining lymph nodes and spleens of mice following primary and secondary L. major infections. DN T cells from healed mice display functional characteristics of protective anti-Leishmania memory-like cells: rapid and extensive proliferation and effector cytokines production following L. major challenge in vitro and in vivo. DN T cells express predominantly (> 95%) alpha-beta T cell receptor (αβ TCR), are Leishmania-specific, restricted mostly by MHC class II molecules and display transcriptional profile of innate-like genes. Using in vivo depletion and adoptive transfer studies, we show that DN T cells contribute to optimal primary and secondary anti-Leishmania immunity in mice. These results directly identify DN T cells as important players in effective and protective primary and secondary anti-L. major immunity in experimental cutaneous leishmaniasis.

Receptors for insulin-like growth factor II (IGF-II) in the rat kidney glomerulus

International Nuclear Information System (INIS)

Haskell, J.F.; Pillion, D.J.; Meezan, E.

1986-01-01

Renal glomeruli were isolated by a technique involving renal perfusion with a solution containing magnetic iron oxide particles, followed by homogenization, sieving and isolation over a strong magnet. Isolated glomeruli were treated with 1% Triton X-100 to solubilize plasma membrane components while insoluble basement membrane components were removed by centrifugation. [ 125 I]Insulin-like growth factor-II (IGF-II) binding to this preparation was competitively inhibited by increasing amounts of unlabelled IGF-II, with 50% inhibition of binding observed at an IGF-II concentration of 1 ng/ml. [ 125 I]IGF-II was covalently cross-linked to its receptor with disuccinimidyl suberate in two tissues known to contain IGF-II receptors, the rat chondrosarcoma chondrocyte and the rat kidney tubule, as well as in rat renal glomeruli. In all three cases, a specific high-molecular weight (Mr = 255,000) band could be identified on autoradiograms of dodecyl sulfate polyacrylamide gels. These results indicate that the rat glomerulus contains a high-affinity receptor for IGF-II. This finding is consistent with the hypothesis that IGF-II plays a role in glomerular growth and differentiation

Temperament and character profiles in bipolar I, bipolar II and major depressive disorder: Impact over illness course, comorbidity pattern and psychopathological features of depression.

Science.gov (United States)

Zaninotto, Leonardo; Souery, Daniel; Calati, Raffaella; Di Nicola, Marco; Montgomery, Stuart; Kasper, Siegfried; Zohar, Joseph; Mendlewicz, Julien; Robert Cloninger, C; Serretti, Alessandro; Janiri, Luigi

2015-09-15

Studies comparing temperament and character traits between patients with mood disorders and healthy individuals have yielded variable results. The Temperament and Character Inventory (TCI) was administered to 101 bipolar I (BP-I), 96 bipolar II (BP-II), 123 major depressive disorder (MDD) patients, and 125 HS. A series of generalized linear models were performed in order to: (a) compare the TCI dimensions across groups; (b) test any effect of the TCI dimensions on clinical features of mood disorders; and (c) detect any association between TCI dimensions and the psychopathological features of a major depressive episode. Demographic and clinical variables were also included in the models as independent variables. Higher Harm Avoidance was found in BP-II and MDD, but not in BP-I. Higher Self-Transcendence was found in BP-I. Our models also showed higher Self-Directedness in HS, either vs MDD or BP-II. No association was found between any TCI dimension and the severity of symptoms. Conversely, a positive association was found between Harm Avoidance and the overall burden of depressive episodes during lifetime. The cross-sectional design and the heterogeneity of the sample may be the main limitations of our study. In general, our sample seems to support the view of a similar profile of temperament and character between MDD and BP-II, characterized by high Harm Avoidance and low Self-Directedness. In contrast, patients with BP-I only exhibit high Self-Transcendence, having a near-normal profile in terms of Harm Avoidance or Self-Directedness. Copyright © 2015 Elsevier B.V. All rights reserved.

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

The relationship between Class I and Class II methanol masers at high angular resolution

Science.gov (United States)

McCarthy, T. P.; Ellingsen, S. P.; Voronkov, M. A.; Cimò, G.

2018-06-01

We have used the Australia Telescope Compact Array (ATCA) to make the first high-resolution observations of a large sample of class I methanol masers in the 95-GHz (80-71A+) transition. The target sources consist of a statistically complete sample of 6.7-GHz class II methanol masers with an associated 95-GHz class I methanol maser, enabling a detailed study of the relationship between the two methanol maser classes at arcsecond angular resolution. These sources have been previously observed at high resolution in the 36- and 44-GHz transitions, allowing comparison between all three class I maser transitions. In total, 172 95-GHz maser components were detected across the 32 target sources. We find that at high resolution, when considering matched maser components, a 3:1 flux density ratio is observed between the 95- and 44-GHz components, consistent with a number of previous lower angular resolution studies. The 95-GHz maser components appear to be preferentially located closer to the driving sources and this may indicate that this transition is more strongly inverted nearby to background continuum sources. We do not observe an elevated association rate between 95-GHz maser emission and more evolved sources, as indicated by the presence of 12.2-GHz class II masers. We find that in the majority of cases where both class I and class II methanol emission is observed, some component of the class I emission is associated with a likely outflow candidate.

Identification and differentiation of major components in three different “Sheng-ma” crude drug species by UPLC/Q-TOF-MS

Directory of Open Access Journals (Sweden)

Mengxue Fan

2017-03-01

Full Text Available Cimicifugae Rhizoma (Sheng ma is a Ranunculaceae herb belonging to a composite family and well known in China. has been widely used in traditional Chinese medicine. The Pharmacopoeia of the People׳s Republic of China contains three varieties (Cimicifuga dahurica (Turcz., Cimicifuga foetida L. and Cimicifuga heracleifolia Kom. which have been used clinically as “Sheng-ma”. However, the chemical constituents of three components of “Sheng-ma” have never been documented. In this study, a rapid method for the analysis of the main components of “Sheng-ma” was developed using ultra-high performance liquid chromatography with quadrupole-time-of-flight mass spectrometry (UPLC/Q-TOF-MS. The present study reveals the major common and distinct chemical constituents of C. dahurica, C. foetida and C. heracleifolia and also reports principal component and statistical analyses of these results. The components were identified by comparing the retention time, accurate mass, mass spectrometric fragmentation characteristic ions and matching empirical molecular formula with that of the published compounds. A total of 32 common components and 8 markers for different “Sheng-ma” components were identified. These findings provide an important basis for the further study and clinical utilities of the three “Sheng-ma” varieties.

Correlation of major components of ocular astigmatism in myopic patients.

Science.gov (United States)

Mohammadpour, Mehrdad; Heidari, Zahra; Khabazkhoob, Mehdi; Amouzegar, Afsaneh; Hashemi, Hassan

2016-02-01

To investigate the correlation of major components of ocular astigmatism in myopic patients in an academic hospital. This cross-sectional study was conducted on 376 eyes of 188 patients who were referred to Farabi Eye Hospital for refractive surgery. Preoperative examinations including refraction and corneal topography were performed for all candidates to measure refractive and corneal astigmatism. Ocular residual astigmatism was calculated using vector analysis. Pearson's correlation and ANOVA analysis were used to evaluate the strength of the association between different types of astigmatism. Both eyes were defined as cluster and the Generalized Estimating Equations (GEE) analysis were performed. Mean age of 119 women (63.3%) and 69 men (36.7%) was 27.8 ± 5.7 years. Mean refractive error based on spherical equivalent was -3.59 ± 1.95D (range, -0.54 to -10.22D). Mean refractive and corneal astigmatism was 1.97 ± 1.3D and 1.85 ± 1.01D, respectively. Mean amount of ORA was 0.65 ± 0.36D.There was a significant correlation between ORA and refractive astigmatism(r=0.23, pcorrelation between ORA and corneal astigmatism (r=0.13, p=0.014). There was a significant correlation between J0 and J45 values of ORA and corneal astigmatism (pcorrelation between ORA and refractive astigmatism, refractive and corneal astigmatism and a weak correlation between ORA and corneal astigmatism in refractive surgery candidates. Identifying the type of astigmatism and preoperative measurement of ocular residual astigmatism is highly recommended prior to any refractive surgery, especially in cases with significant astigmatism. Copyright © 2015 Elsevier Ltd. All rights reserved.

Manganese Loading and Photosystem II Stability are Key Components of Manganese Efficiency in Plants

DEFF Research Database (Denmark)

Schmidt, Sidsel Birkelund

Manganese (Mn) deficiency constitutes a major plant nutritional problem in commercial crop production of winter cereals. In plants, Mn has an indispensable role in the oxygen evolving complex (OEC) of photosystem II (PSII). Hence, the consequences of Mn deficiency are reduced plant growth......, and eventually substantial yield losses. It is well known, that genotypes within plant species differ considerably in tolerance to growth under Mn limiting conditions, a phenomenon designated as Mn efficiency. However, the physiological responses reflecting the underlying mechanisms of Mn efficiency are still...... not fully understood. In this PhD study, a new method for determination and characterization of metal binding in size-fractionated photosynthetic protein complexes from barley thylakoids was established. The applicability of the method was shown by quantification of Mn binding in PSII from thylakoids of two...

Widespread IgE-mediated hypersensitivity in the Sudan to the 'green nimitti' midge, Cladotanytarsus lewisi (Diptera: Chironomidae) II. Identification of a major allergen.

Science.gov (United States)

Gad El Rab, M O; Thatcher, D R; Kay, A B

1980-01-01

A major allergen has been identified in an aqueous extract of the 'green nimitti' midge, Cladotanytarsus lewisi (Diptera: Chironomidae). Following chromatography on Sephadex G-100 allergenic activity, as assessed by skin ('prick') testing, eluted as two closely related peaks (pools I and II) at about 50% bed volume. When these pools were applied separately to columns of CM-cellulose, activity in each eluted with 0 . 05 M NaCl. Isoelectric focusing of the unfractionated allergen gave a single peak of activity at pI 4 . 3. By SDS--PAGE, biological activity in the whole 'green nimitti' extract and the material eluting from both pools I and II of the Sephadex G-100 column migrated to the same positions and were associated with a molecular size of 15,000--20,000 daltons. Skin test reactivity of the unfractionated material and the Sephadex G-100 pool I and II eluates were all destroyed following incubation with trypsin, chymotrypsin, thermolysin and neuraminidase. These experiments indicate that a major allergen derived from the 'green nimitti' midge, a cause of widespread and severe immediate-type allergy in the Sudan, is an acidic glycoprotein of 15,000--20,000 molecular weight. Images Fig. 3 PMID:7438559

Ceftriaxone-associated gallbladder sludge. Identification of calcium-ceftriaxone salt as a major component of gallbladder precipitate

International Nuclear Information System (INIS)

Park, H.Z.; Lee, S.P.; Schy, A.L.

1991-01-01

Ceftriaxone, a third-generation cephalosporin, is partially excreted into bile. With its clinical use, the formation of gallbladder sludge detected by ultrasonography has been reported. Four surgical specimens were examined and no gallstones were found. Instead, fine precipitates of 20-250 microns were present. Microscopically, there was a small number of cholesterol monohydrate crystals and bilirubin granules among an abundant amount of granular-crystalline material that was not morphologically cholesterol monohydrate crystals. The chemical composition of the precipitates (n = 4) was determined. There was a small amount of cholesterol (1.7% +/- 0.8%) and bilirubin (13.9% +/- 0.74%). The major component of the precipitate was a residue. On further analysis using thin-layer chromatography, high-performance liquid chromatography, and electron microprobe analysis, the residue was identified as a calcium salt of ceftriaxone. The residue also had identical crystal morphology and chromatographic elution profile as authentic calcium-ceftriaxone standards. It is concluded that ceftriaxone, after excretion and being concentrated in the gallbladder bile, can form a precipitate. The major constituent has been identified as a ceftriaxone-calcium salt

Intelligent distributed control for nuclear power plants

International Nuclear Information System (INIS)

Klevans, E.H.

1991-01-01

In September of 1989 work began on the DOE University Program grant DE-FG07-89ER12889. The grant provides support for a three year project to develop and demonstrate Intelligent Distributed Control (IDC) for Nuclear Power Plants. The body of this First Annual Technical Progress report summarizes the first year tasks while the appendices provide detailed information presented at conference meetings. One major addendum report, authored by M.A. Schultz, describes the ultimate goals and projected structure of an automatic distributed control system for EBR-2. The remaining tasks of the project develop specific implementations of various components required to demonstrate the intelligent distributed control concept

Efficacy and safety of elbasvir/grazoprevir and sofosbuvir/pegylated interferon/ribavirin

DEFF Research Database (Denmark)

Sperl, Jan; Horvath, Gabor; Halota, Waldemar

2016-01-01

daily (n=129) or sofosbuvir (400mg once daily) plus PR (n=128). Primary efficacy objective was sustained virologic response 12weeks after the end of therapy (SVR12, HCV RNA tier 1 safety event. RESULTS: The majority...... the lower bound of the 1-sided 1-sample exact test was greater than -10% and greater than zero, both non-inferiority and superiority of EBR/GZR vs. SOF/PR were established. The frequency of tier 1 safety events was lower among patients receiving EBR/GZR than SOF/PR (0.8% vs. 27.8%, between group difference...... pegylated interferon/ribavirin (SOF/PR) in patients with HCV infection. METHODS: This was a randomized, open-label, phase III trial. Two hundred fifty-seven patients with HCV genotype (GT)1 or 4 infection and baseline viral load >10,000IU/ml were randomized to receive 12weeks of EBR/GZR 50mg/100mg once...

User's manual for the BNW-II optimization code for dry/wet-cooled power plants

International Nuclear Information System (INIS)

Braun, D.J.; Bamberger, J.A.; Braun, D.J.; Faletti, D.W.; Wiles, L.E.

1978-05-01

The User's Manual describes how to operate BNW-II, a computer code developed by the Pacific Northwest Laboratory (PNL) as a part of its activities under the Department of Energy (DOE) Dry Cooling Enhancement Program. The computer program offers a comprehensive method of evaluating the cost savings potential of dry/wet-cooled heat rejection systems. Going beyond simple ''figure-of-merit'' cooling tower optimization, this method includes such items as the cost of annual replacement capacity, and the optimum split between plant scale-up and replacement capacity, as well as the purchase and operating costs of all major heat rejection components. Hence the BNW-II code is a useful tool for determining potential cost savings of new dry/wet surfaces, new piping, or other components as part of an optimized system for a dry/wet-cooled plant

Biogenic non-crystalline U(IV) revealed as major component in uranium ore deposits

Science.gov (United States)

Bhattacharyya, Amrita; Campbell, Kate M.; Kelly, Shelly D.; Roebbert, Yvonne; Weyer, Stefan; Bernier-Latmani, Rizlan; Borch, Thomas

2017-06-01

Historically, it is believed that crystalline uraninite, produced via the abiotic reduction of hexavalent uranium (U(VI)) is the dominant reduced U species formed in low-temperature uranium roll-front ore deposits. Here we show that non-crystalline U(IV) generated through biologically mediated U(VI) reduction is the predominant U(IV) species in an undisturbed U roll-front ore deposit in Wyoming, USA. Characterization of U species revealed that the majority (~58-89%) of U is bound as U(IV) to C-containing organic functional groups or inorganic carbonate, while uraninite and U(VI) represent only minor components. The uranium deposit exhibited mostly 238U-enriched isotope signatures, consistent with largely biotic reduction of U(VI) to U(IV). This finding implies that biogenic processes are more important to uranium ore genesis than previously understood. The predominance of a relatively labile form of U(IV) also provides an opportunity for a more economical and environmentally benign mining process, as well as the design of more effective post-mining restoration strategies and human health-risk assessment.

Upgrade of the SLAC SLED II Pulse Compression System Based on Recent High Power Tests

International Nuclear Information System (INIS)

Vlieks, A.E.; Fowkes, W.R.; Loewen, R.J.; Tantawi, S.G.

2011-01-01

In the Next Linear Collider (NLC) it is expected that the high power rf components be able to handle peak power levels in excess of 400 MW. We present recent results of high power tests designed to investigate the RF breakdown limits of the X-band pulse compression system used at SLAC. (SLED-II). Results of these tests show that both the TE 01 -TE 10 mode converter and the 4-port hybrid have a maximum useful power limit of 220-250 MW. Based on these tests, modifications of these components have been undertaken to improve their peak field handling capability. Results of these modifications will be presented. As part of an international effort to develop a new 0.5-1.5 TeV electron-positron linear collider for the 21st century, SLAC has been working towards a design, referred to as 'The Next Linear Collider' (NLC), which will operate at 11.424 GHz and utilize 50-75 MW klystrons as rf power sources. One of the major challenges in this design, or any other design, is how to generate and efficiently transport extremely high rf power from a source to an accelerator structure. SLAC has been investigating various methods of 'pulse compressing' a relatively wide rf pulse ((ge) 1 μs) from a klystron into a narrower, but more intense, pulse. Currently a SLED-II pulse compression scheme is being used at SLAC in the NLC Test Accelerator (NLCTA) and in the Accelerator Structures Test Area (ASTA) to provide high rf power for accelerator and component testing. In ASTA, a 1.05 μs pulse from a 50 MW klystron was successfully pulse compressed to 205 MW with a pulse width of 150 ns. Since operation in NLC will require generating and transporting rf power in excess of 400 MW it was decided to test the breakdown limits of the SLED-II rf components in ASTA with rf power up to the maximum available of 400 MW. This required the combining of power from two 50 MW klystrons and feeding the summed power into the SLED-II pulse compressor. Results from this experiment demonstrated that two of

Fabrication of ThO2, UO2, and PuO2-UO2 pellets

International Nuclear Information System (INIS)

Rasmussen, D.E.; Jentzen, W.R.; McCord, R.B.

1978-01-01

Fabrication of ThO pellets for EBR-II irradiation testing and fabrication of UO 2 and PuO 2 -UO 2 pellets for United Kingdom Prototype Fast Reactor (PFR) irradiation testing is discussed. Effect of process parameters on density and microstructure of pellets fabricated by the cold press and sinter technique is reviewed

Real time operating system for a nuclear power plant computer

International Nuclear Information System (INIS)

Alger, L.S.; Lala, J.H.

1986-01-01

A quadruply redundant synchronous fault tolerant processor (FTP) is now under fabrication at the C.S. Draper Laboratory to be used initially as a trip monitor for the Experimental Breeder Reactor EBR-II operated by the Argonne National Laboratory in Idaho Falls, Idaho. The real time operating system for this processor is described

Application of microprocessor based controller in the Breeder Reactor Program

International Nuclear Information System (INIS)

Messick, N.C.; Lukas, M.P.

1985-01-01

This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment

Catalog of physical protection equipment. Book 1: Volume II. Intrusion detection components

International Nuclear Information System (INIS)

Haberman, W.

1977-06-01

This volume covers acoustic components, microwave/radar components, electro-optic barriers, electric field components, orientation components, ferrous metal detection components, proximity detection components, vibration detection components, seismic components, pressure-sensitive components, pressure mats, continuity components, electrical/magnetic switches, fire detection components, and mechanical contact switches

Plants lacking the main light-harvesting complex retain photosystem II macro-organization

OpenAIRE

Ruban, AV; Wentworth, M; Yakushevska, AE; Andersson, J; Lee, PJ; Keegstra, W; Dekker, JP; Boekema, EJ; Jansson, S; Horton, P

2003-01-01

Photosystem II (PSII) is a key component of photosynthesis, the process of converting sunlight into the chemical energy of life. In plant cells, it forms a unique oligomeric macrostructure in membranes of the chloroplasts(1). Several light-harvesting antenna complexes are organized precisely in the PSII macrostructure-the major trimeric complexes (LHCII)(2) that bind 70% of PSII chlorophyll and three minor monomeric complexes(3)-which together form PSII supercomplexes(4-6). The antenna comple...

Glycosylinositol phosphorylceramides from Rosa cell cultures are boron-bridged in the plasma membrane and form complexes with rhamnogalacturonan II.

Science.gov (United States)

Voxeur, Aline; Fry, Stephen C

2014-07-01

Boron (B) is essential for plant cell-wall structure and membrane functions. Compared with its role in cross-linking the pectic domain rhamnogalacturonan II (RG-II), little information is known about the biological role of B in membranes. Here, we investigated the involvement of glycosylinositol phosphorylceramides (GIPCs), major components of lipid rafts, in the membrane requirement for B. Using thin-layer chromatography and mass spectrometry, we first characterized GIPCs from Rosa cell culture. The major GIPC has one hexose residue, one hexuronic acid residue, inositol phosphate, and a ceramide moiety with a C18 trihydroxylated mono-unsaturated long-chain base and a C24 monohydroxylated saturated fatty acid. Disrupting B bridging (by B starvation in vivo or by treatment with cold dilute HCl or with excess borate in vitro) enhanced the GIPCs' extractability. As RG-II is the main B-binding site in plants, we investigated whether it could form a B-centred complex with GIPCs. Using high-voltage paper electrophoresis, we showed that addition of GIPCs decreased the electrophoretic mobility of radiolabelled RG-II, suggesting formation of a GIPC-B-RG-II complex. Last, using polyacrylamide gel electrophoresis, we showed that added GIPCs facilitate RG-II dimerization in vitro. We conclude that B plays a structural role in the plasma membrane. The disruption of membrane components by high borate may account for the phytotoxicity of excess B. Moreover, the in-vitro formation of a GIPC-B-RG-II complex gives the first molecular explanation of the wall-membrane attachment sites observed in vivo. Finally, our results suggest a role for GIPCs in the RG-II dimerization process. © 2014 The Authors. The Plant Journal published by Society for Experimental Biology and John Wiley & Sons Ltd.

Policy issues on the control of major accident hazards and the new Seveso II directive

International Nuclear Information System (INIS)

Porter, S.; Wettig, J.

1999-01-01

No one wants a major accident to occur! This paper describes the development of EC policies to convert this simple and obvious fact into a coherent prevention strategy which can deliver a high level of protection throughout the European Community. The Seveso II Directive is described in detail, being the relevant Community Instrument which Member States must implement in their national laws. The need to achieve the correct balance between setting general goals and being over prescriptive is discussed, commensurate with the intent to be flexible but yet consistent and effective at the same time. The main changes from Seveso I are discussed, including requirements related to the operator's management systems, the competent authority's systems for inspection, and information and consultation arrangements with the public. (Copyright (c) 1999 Elsevier Science B.V., Amsterdam. All rights reserved.)

Open-access MIMIC-II database for intensive care research.

Science.gov (United States)

Lee, Joon; Scott, Daniel J; Villarroel, Mauricio; Clifford, Gari D; Saeed, Mohammed; Mark, Roger G

2011-01-01

The critical state of intensive care unit (ICU) patients demands close monitoring, and as a result a large volume of multi-parameter data is collected continuously. This represents a unique opportunity for researchers interested in clinical data mining. We sought to foster a more transparent and efficient intensive care research community by building a publicly available ICU database, namely Multiparameter Intelligent Monitoring in Intensive Care II (MIMIC-II). The data harnessed in MIMIC-II were collected from the ICUs of Beth Israel Deaconess Medical Center from 2001 to 2008 and represent 26,870 adult hospital admissions (version 2.6). MIMIC-II consists of two major components: clinical data and physiological waveforms. The clinical data, which include patient demographics, intravenous medication drip rates, and laboratory test results, were organized into a relational database. The physiological waveforms, including 125 Hz signals recorded at bedside and corresponding vital signs, were stored in an open-source format. MIMIC-II data were also deidentified in order to remove protected health information. Any interested researcher can gain access to MIMIC-II free of charge after signing a data use agreement and completing human subjects training. MIMIC-II can support a wide variety of research studies, ranging from the development of clinical decision support algorithms to retrospective clinical studies. We anticipate that MIMIC-II will be an invaluable resource for intensive care research by stimulating fair comparisons among different studies.

A Mixed-Mode (I-II) Fracture Criterion for AS4/8552 Carbon/Epoxy Composite Laminate

Science.gov (United States)

Karnati, Sidharth Reddy

A majority of aerospace structures are subjected to bending and stretching loads that introduce peel and shear stresses between the plies of a composite laminate. These two stress components cause a combination of mode I and II fracture modes in the matrix layer of the composite laminate. The most common failure mode in laminated composites is delamination that affects the structural integrity of composite structures. Damage tolerant designs of structures require two types of materials data: mixed-mode (I-II) delamination fracture toughness that predicts failure and delamination growth rate that predicts the life of the structural component. This research focuses determining mixed-mode (I-II) fracture toughness under a combination of mode I and mode II stress states and then a fracture criterion for AS4/8552 composite laminate, which is widely used in general aviation. The AS4/8552 prepreg was supplied by Hexcel Corporation and autoclave fabricated into a 20-ply unidirectional laminate with an artificial delamination by a Fluorinated Ethylene Propylene (FEP) film at the mid-plane. Standard split beam specimens were prepared and tested in double cantilever beam (DCB) and end notched flexure modes to determine mode I (GIC) and II (GIIC) fracture toughnesses, respectively. The DCB specimens were also tested in a modified mixed-mode bending apparatus at GIIm /GT ratios of 0.18, 0.37, 0.57 and 0.78, where GT is total and GIIm is the mode II component of energy release rates. The measured fracture toughness, GC, was found to follow the locus a power law equation. The equation was validated for the present and literature experimental data.

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

International experience with the bundle behavior of fuel elements of sodium cooled reactors; derivation of a figure of merit for the judgement of fuel pin bundle parameters with respect to abrasion due to thermoelastic pin-pin interaction

International Nuclear Information System (INIS)

Toebbe, H.

1987-10-01

The report describes the status of experience with respect to the abrasion behavior of bundles in standard fuel elements and test elements with wire or grid spacing in the reactors Rapsodie fortissimo, Phenix, DFR, PFR, EBR-II, FFTF, JOYO and KNK II. With the help of simple considerations concerning thermoelastic pin-pin interactions a figure of merit is deduced from the different bundle parameters, which allows a comparative judgement of the parameters of different bundle concepts [de

Applying neural networks to optimize instrumentation performance

Energy Technology Data Exchange (ETDEWEB)

Start, S.E.; Peters, G.G.

1995-06-01

Well calibrated instrumentation is essential in providing meaningful information about the status of a plant. Signals from plant instrumentation frequently have inherent non-linearities, may be affected by environmental conditions and can therefore cause calibration difficulties for the people who maintain them. Two neural network approaches are described in this paper for improving the accuracy of a non-linear, temperature sensitive level probe ised in Expermental Breeder Reactor II (EBR-II) that was difficult to calibrate.

Applying neural networks to optimize instrumentation performance

International Nuclear Information System (INIS)

Start, S.E.; Peters, G.G.

1995-01-01

Well calibrated instrumentation is essential in providing meaningful information about the status of a plant. Signals from plant instrumentation frequently have inherent non-linearities, may be affected by environmental conditions and can therefore cause calibration difficulties for the people who maintain them. Two neural network approaches are described in this paper for improving the accuracy of a non-linear, temperature sensitive level probe ised in Expermental Breeder Reactor II (EBR-II) that was difficult to calibrate

Initial results for electrochemical dissolution of spent EBR-II fuel

International Nuclear Information System (INIS)

Li, S. X.

1998-01-01

Initial results are reported for the anode behavior of spent metallic nuclear fuel in an electrorefining process. The anode behavior has been characterized in terms of the initial spent fuel composition and the final composition of the residual cladding hulls. A variety of results have been obtained depending on the experimental conditions. Some of the process variables considered are average and maximum cell voltage, average and maximum anode voltage, amount of electrical charge passed (coulombs or amp-hours) during the experiment, and cell resistance. The main goal of the experiments has been the nearly complete dissolution of uranium with the retention of zirconium and noble metal fission products in the cladding hulls. Analysis has shown that the most indicative parameters for determining an endpoint to the process, recognizing the stated goal, are the maximum anode voltage and the amount of electrical charge passed. For the initial experiments reported here, the best result obtained is greater than 98% uranium dissolution with approximately 50% zirconium retention. Noble metal fission product retention appears to be correlated with zirconium retention

Recent advances during the treatment of spent EBR-II fuel

International Nuclear Information System (INIS)

Westphal, B.R.; Mariani, R.D.; Vaden, D.E.; Sherman, S.R.; Li, S.X.; Keiser, D.D. Jr.

2000-01-01

Several recent advances have been achieved for the electrometallurgical treatment of spent nuclear fuel. In anticipation of production operations at Argonne National Laboratory-West, development of both electrorefining and metal processing has been ongoing in the post-demonstration phase in order to further optimize the process. These development activities show considerable promise. This paper discusses the results of recent experiments as well as plans for future investigations

Optimization of simultaneous electrochemical determination of Cd(II), Pb(II), Cu(II) and Hg(II) at carbon nanotube-modified graphite electrodes.

Science.gov (United States)

Pikna, L'ubomír; Heželová, Mária; Kováčová, Zuzana

2015-01-01

The health of the environment is worsening every day. Monitoring of potentially toxic elements and remediation of environmental pollution are necessary. Therefore, the research and development of simple, inexpensive, portable and effective sensors is important. Electrochemistry is a useful component of the field of environment monitoring. The present study focuses on evaluating and comparing three types of electrodes (PIGE, PIGE/MWCNT/HNO3 and PIGE/MWCNT/EDTA/HNO3) employed for the simultaneous electrochemical determination of four potentially toxic elements: Cd(II), Pb(II), Cu(II) and Hg(II). Cyclic voltammograms were measured in an acetate buffer. The LOD, LOQ, the standard and relative precisions of the method and a prediction intervals were calculated (according to the technical procedure DIN 32 645) for the three electrodes and for each measured element. The LOD for PIGE/CNT/HNO3 (the electrode with narrowest calculated prediction intervals) was 2.98 × 10(-7) mol L(-1) for Cd(II), 4.83 × 10(-7) mol L(-1) for Pb(II), 3.81 × 10(-7) mol L(-1) for Cu(II), 6.79 × 10(-7) mol L(-1) for Hg(II). One of the benefits of this study was the determination of the amount of Hg(II) in the mixture of other elements.

Design fix for vibration-induced wear in fuel pin bundles

International Nuclear Information System (INIS)

Naas, D.F.; Heck, E.N.

1976-01-01

In summary, results at 45,000 MWd/MTM burnup from the FFTF mixed oxide fuel pin irradiation tests in EBR-II show that reduction of the initial fuel pin bundle clearance and use of 20 percent cold-worked stainless steel ducts virtually eliminate vibration and wear observed in an initial series of 61-pin tests

Validation of an intermediate heat exchanger model for real time analysis

International Nuclear Information System (INIS)

Tzanos, C.P.

1986-11-01

A new method was presented for LMFBR intermediate heat exchanger (IHX) analysis in real time for purposes of continuous on-line data validation, plant state verification and fault identification. For the validation of this methodology the EBR-II IHX transient during Test 8A was analyzed. This paper presents the results of this analysis

Evaluation of cross sections of Th-232 and U-233

International Nuclear Information System (INIS)

Dias, A.M.

1978-01-01

The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

Post Process Characterization of Friction Stir Welded Components, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — Luna Innovations Incorporated proposes in this STTR Phase II project to continue development and validation of Luna's amplitude-dependent, nonlinear ultrasonic...

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

High reliability fuel in the US

International Nuclear Information System (INIS)

Neuhold, R.J.; Leggett, R.D.; Walters, L.C.; Matthews, R.B.

1986-05-01

The fuels development program of the United States is described for liquid metal reactors (LMR's). The experience base, status and future potential are discussed for the three systems - oxide, metal and carbide - that have proved to have high reliability. Information is presented showing burnup capability of the oxide fuel system in a large core, e.g., FFTF, to be 150 MWd/kgM with today's technology with the potential for a capability as high as 300 MWd/kgM. Data provided for the metal fuel system show 8 at. % being routinely achieved as the EBR-II driver fuel with good potential for extending this to 15 at. % since special test pins have already exceeded this burnup level. The data included for the carbide fuel system are from pin and assembly irradiations in EBR-II and FFTF, respectively. Burnup to 12 at. % appears readily achievable with burnups to 20 at. % being demonstrated in a few pins. Efforts continue on all three systems with the bulk of the activity on metal and oxide

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)

454 sequencing reveals extreme complexity of the class II Major Histocompatibility Complex in the collared flycatcher

Directory of Open Access Journals (Sweden)

Gustafsson Lars

2010-12-01

Full Text Available Abstract Background Because of their functional significance, the Major Histocompatibility Complex (MHC class I and II genes have been the subject of continuous interest in the fields of ecology, evolution and conservation. In some vertebrate groups MHC consists of multiple loci with similar alleles; therefore, the multiple loci must be genotyped simultaneously. In such complex systems, understanding of the evolutionary patterns and their causes has been limited due to challenges posed by genotyping. Results Here we used 454 amplicon sequencing to characterize MHC class IIB exon 2 variation in the collared flycatcher, an important organism in evolutionary and immuno-ecological studies. On the basis of over 152,000 sequencing reads we identified 194 putative alleles in 237 individuals. We found an extreme complexity of the MHC class IIB in the collared flycatchers, with our estimates pointing to the presence of at least nine expressed loci and a large, though difficult to estimate precisely, number of pseudogene loci. Many similar alleles occurred in the pseudogenes indicating either a series of recent duplications or extensive concerted evolution. The expressed alleles showed unambiguous signals of historical selection and the occurrence of apparent interlocus exchange of alleles. Placing the collared flycatcher's MHC sequences in the context of passerine diversity revealed transspecific MHC class II evolution within the Muscicapidae family. Conclusions 454 amplicon sequencing is an effective tool for advancing our understanding of the MHC class II structure and evolutionary patterns in Passeriformes. We found a highly dynamic pattern of evolution of MHC class IIB genes with strong signals of selection and pronounced sequence divergence in expressed genes, in contrast to the apparent sequence homogenization in pseudogenes. We show that next generation sequencing offers a universal, affordable method for the characterization and, in perspective

Component failure data base of TRIGA reactors

International Nuclear Information System (INIS)

Djuricic, M.

2004-10-01

This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

Detection of Copper (II) and Cadmium (II) binding to dissolved organic matter from macrophyte decomposition by fluorescence excitation-emission matrix spectra combined with parallel factor analysis

International Nuclear Information System (INIS)

Yuan, Dong-hai; Guo, Xu-jing; Wen, Li; He, Lian-sheng; Wang, Jing-gang; Li, Jun-qi

2015-01-01

Fluorescence excitation-emission matrix (EEM) spectra coupled with parallel factor analysis (PARAFAC) was used to characterize dissolved organic matter (DOM) derived from macrophyte decomposition, and to study its complexation with Cu (II) and Cd (II). Both the protein-like and the humic-like components showed a marked quenching effect by Cu (II). Negligible quenching effects were found for Cd (II) by components 1, 5 and 6. The stability constants and the fraction of the binding fluorophores for humic-like components and Cu (II) can be influenced by macrophyte decomposition of various weight gradients in aquatic plants. Macrophyte decomposition within the scope of the appropriate aquatic phytomass can maximize the stability constant of DOM-metal complexes. A large amount of organic matter was introduced into the aquatic environment by macrophyte decomposition, suggesting that the potential risk of DOM as a carrier of heavy metal contamination in macrophytic lakes should not be ignored. - Highlights: • Macrophyte decomposition increases fluorescent DOM components in the upper sediment. • Protein-like components are quenched or enhanced by adding Cu (II) and Cd (II). • Macrophyte decomposition DOM can impact the affinity of Cu (II) and Cd (II). • The log K M and f values showed a marked change due to macrophyte decomposition. • Macrophyte decomposition can maximize the stability constant of DOM-Cu (II) complexes. - Macrophyte decomposition DOM can influence on the binding affinity of metal ions in macrophytic lakes

MENO-II: An AI-Based Programming Tutor.

Science.gov (United States)

Soloway, Elliot; And Others

This report examines the features and performance of the BUG-FINDing component of MENO-II, a computer-based tutor for beginning PASCAL programming students. A discussion of the use of artificial intelligence techniques is followed by a summary of the system status and objectives. The two main components of MENO-II are described, beginning with the…