EBR-II: search for the lost subassembly

International Nuclear Information System (INIS)

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

1983-01-01

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

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

The physics design of EBR-II

International Nuclear Information System (INIS)

Loewenstein, W.B.

1962-01-01

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

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

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.

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

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

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

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

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

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

Operating and test experience of EBR-II

International Nuclear Information System (INIS)

Sackett, J.I.

1991-01-01

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

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.

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

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

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.)

Planning for closure and deactivation of the EBR-II complex

International Nuclear Information System (INIS)

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

1997-01-01

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

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

System modeling and simulation at EBR-II

International Nuclear Information System (INIS)

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

1986-01-01

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

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.)

Demonstration of passive safety features in EBR-II

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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

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.)

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

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

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

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

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 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

Component configuration control system development at EBR-II

International Nuclear Information System (INIS)

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

1984-01-01

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

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

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

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

Recent operating experiences and programs at EBR-II

International Nuclear Information System (INIS)

Lentz, G.L.

1984-01-01

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

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

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

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

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)

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

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

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

EBR-II experience with sodium cleaning and radioactivity decontamination

International Nuclear Information System (INIS)

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

1978-01-01

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

EBR-II experience with sodium cleaning and radioactivity decontamination

Energy Technology Data Exchange (ETDEWEB)

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

1978-08-01

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

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

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

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

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)

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

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

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.

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

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

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.

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

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

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

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

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

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

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

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

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

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

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

EBR-II Primary Tank Wash-Water Alternatives Evaluation

International Nuclear Information System (INIS)

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

2009-01-01

The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. Approximately 1100 kg of residual sodium remained in the primary system after draining the bulk sodium. To stabilize the remaining sodium, both the primary and secondary systems were treated with a purge of moist carbon dioxide. The passivation treatment was stopped in 2005 and the primary system is maintained under a blanket of dry carbon dioxide. Approximately 670 kg of sodium metal remains in the primary system in locations that were inaccessible to passivation treatment or in pools of sodium that were too deep for complete penetration of the passivation treatment. The EBR-II reactor was permitted by the Idaho Department of Environmental Quality (DEQ) in 2002 under a RCRA permit that requires removal of all remaining sodium. The proposed baseline closure method would remove the large components from the primary tank, fill the primary system with water, react the remaining sodium with the water and dissolve the reaction products in about 100,000 gallons of wash water. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to evaluate alternatives that could meet the RCRA permit clean closure requirements and minimize the quantity of hazardous waste generated by the cleanup process. The workshop convened a panel of national and international sodium cleanup specialists, subject matter experts from the INL, and the EBR-II Wash Water Project team that organized the workshop. The workshop was conducted by a trained facilitator using Value Engineering techniques to elicit the most technically sound solutions from the workshop participants. A brainstorming session was held to identify possible alternative treatment methods that would meet the primary functions and criteria of neutralizing the hazards, maximizing byproduct removal and minimizing waste generation. An initial list of some 20 probable alternatives was evaluated and refined down

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

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

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

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

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

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.

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

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

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

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

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.)

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

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

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

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

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

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

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

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

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1988-01-01

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

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.)

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

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

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

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

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

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

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

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.

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)

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

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

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

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

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

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

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

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

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.

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

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

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

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1986-01-01

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

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

EBR-II facility for cleaning and maintenance of LMR components

International Nuclear Information System (INIS)

Washburn, R.A.

1986-01-01

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

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)

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

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

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)

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.

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

Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

Science.gov (United States)

Geithoff, D.; Mühling, G.; Richter, K.

1992-06-01

The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

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

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

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

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

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.

Irradiation performance of full-length metallic IFR fuels

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1992-07-01

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

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.)

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

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

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

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

2010-01-31

An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

Estimates of time-dependent fatigue behavior of Type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

International Nuclear Information System (INIS)

Brinkman, C.R.; Liu, K.C.; Grossbeck, M.L.

1978-01-01

Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irradiated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63*10 26 neutrons (n)/m 2 (E>0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20 percent cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradiations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadings ranging from 2 to 5 MW/m 2 were used. 27 refs

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.

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.

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

Calculation of displacement and helium production at the LAMPF irradiation facility

International Nuclear Information System (INIS)

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

1984-01-01

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 SPECTER 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 EBR-II and RTNS-II. The results indicate substantial contributions to the displacement and helium production rates due to neutrons in the high-energy tail (above 40 MeV) of the LAMPF spallation neutron spectrum. Still higher production rates are calculated for irradiations in the direct proton beam. These results will provide useful background information for research to be conducted at a new irradiation facility at LAMPF

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

Direct electrical heating of irradiated metal fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

1985-01-01

The Integral Fast Reactor (IFR) concept proposed by Argonne National Laboratory utilizes a metal fuel core. Reactor safety analysis requires information on the potential for fuel axial expansion during severe thermal transients. In addition to a comparatively large thermal expansion coefficient, metallic fuel has a unique potential for enhanced pre-failure expansion driven by retained fission gas and ingested bond sodium. In this paper, the authors present preliminary results from three direct electrical heating (DEH) experiments performed on irradiated metal fuel to investigate axial expansion behavior. The test samples were from Experimental Breeder Reactor II (EBR-II) driver fuel ML-11 irradiated to 8 at.% burnup. Preliminary analysis of the results suggest that enhanced expansion driven by trapped fission gas can occur

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

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.)

Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1985-11-01

This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs

Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Maziasz, P.J.

1985-11-01

This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

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

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

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

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

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

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

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.

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

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

Irradiation performance of metallic fuels

International Nuclear Information System (INIS)

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

1989-01-01

Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

Effect of material variables on the irradiation performance of boron carbide

International Nuclear Information System (INIS)

Basmajian, J.A.; Hollenberg, G.W.

1980-01-01

Boron carbide pellets were fabricated with variations in material parameters. These pellets were irradiated in the Experimental Breeder Reactor-II (EBR-II) to determine the effect of these variations on the performance. Helium release from the material and swelling of the pellets are the primary measures of performance. It was determined that material with a smaller grain size released more helium and swelled less. The pellets with boron-to-carbon ratios greater than 4 to 1 did not perform well. Iron additions improved the performance of the material while density variations had little effect

Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

1994-10-01

This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

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

Interaction of irradiation creep and swelling in the creep disappearance regime

International Nuclear Information System (INIS)

Garner, F.A.; Toloczko, M.B.

1992-01-01

The objective of this effort is to determine the relationship between applied stresses and irradiation-induced dimensional changes in structural metals for fusion applications. Reanalysis of an earlier data set derived from irradiation of long creep tubes in EBR-II at 550 C has shown that the creep-swelling coupling coefficient is relatively independent of temperature at ∼0.6 x 10 -2 MPa -1 , but falls with increases in the swelling rate, especially at high stress levels. The action of stress-affected swelling and carbide precipitation exert different influences on the derivation of this coefficient

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.

Swelling in neutron-irradiated titanium alloys

International Nuclear Information System (INIS)

Peterson, D.T.

1982-04-01

Immersion density measurements have been performed on a series of titanium alloys irradiated in EBR-II to a fluence of 5 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 and 550 0 C. The materials irradiated were the near-alpha alloys Ti-6242S and Ti-5621S, the alpha-beta alloy Ti-64, and the beta alloy Ti-38644. Swelling was observed in all alloys with the greater swelling being observed at 550 0 C. Microstructural examination revealed the presence of voids in all alloys. Ti-38644 was found to be the most radiation resistant. Ti-6242S and Ti-5621S also displayed good radiation resistance, whereas considerable swelling and precipitation were observed in Ti-64 at 550 0 C

Precipitation response of annealed type 316 stainless steel in HFIR irradiations at 550 to 6800C

International Nuclear Information System (INIS)

Maziasz, P.J.

1978-01-01

Precipitation in annealed type 316 stainless steel after HFIR irradiation at 550--680 0 C to fluences producing 2000--3300 at. ppM He and 30--47 dpa is changed relative to fast reactor or thermal aging exposure to similar temperatures and times. The phases observed after HFIR irradiation are the same as those observed after aging to temperatures 70--200 0 C higher or for much longer times. There is a similar temperature shift in addition to different phases observed for HFIR irradiation compared with EBR-II. The changes observed are coincident with including simultaneous helium production to high levels in the irradiation damage products of the material

An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys

International Nuclear Information System (INIS)

Fidleris, V.; Tucker, R.P.; Adamson, R.B.

1987-01-01

This paper presents an overview of factors affecting irradiation growth of zirconium alloys. Recent data obtained from irradiation programs in EBR-II, ATR, and NRU reactors are used to illustrate the effects of various microstructural and experimental factors on the growth of Zircaloy, zirconium, and zirconium-biobium alloys irradiated to fluences up to 2 X 10 26 nm -2 (E > 1 MeV) over the temperature range 330 to 720 K. Open literature results are also used to confirm or illustrate various effects. Important factors are texture, grain boundary parameters, residual stresses, original dislocation density, microstructure evolution, temperature during irradiation, solute effects, and fluence

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

Reliability of fast reactor mixed-oxide fuel during operational transients

International Nuclear Information System (INIS)

Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

1991-07-01

Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

Calculation of displacement and helium production at the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF) irradiation facility

International Nuclear Information System (INIS)

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

1984-01-01

CT: Differential and total displacement and helium production rates are calculated for copper irradiated by spallation neutrons and 760 MeV protons at the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF). The calculations are performed using the SPECTER 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 EBR-II and RTNS-II. The results indicate substantial contributions to the displacement and helium production rates due to neutrons in the high-energy tail (above 20 MeV) of the LAMPF spallation neutron spectrum. Still higher production rates are calculated for irradiations in the direct proton beam. These results will provide useful background information for research to be conducted at a new irradiation facility at LAMPF

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.

Effect of irradiation on the tensile properties of niobium-base alloys

International Nuclear Information System (INIS)

Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

1986-11-01

The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions

Swelling in several commercial alloys irradiated to very high neutron fluence

International Nuclear Information System (INIS)

Gelles, D.S.; Pintler, J.S.

1984-01-01

Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10 23 n/cm 2 (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650 0 C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys

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

Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

International Nuclear Information System (INIS)

Lawrence, L.A.; Woodley, R.E.; Weber, E.T.

1976-06-01

Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

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

Fission gas retention in irradiated metallic fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Gruber, E.; Kramer, J.M.

1987-01-01

Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

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

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

International Nuclear Information System (INIS)

Hollenberg, G.W.

1983-04-01

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

Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

Energy Technology Data Exchange (ETDEWEB)

Rabenberg, Ellen M.; Jaques, Brian J. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Sencer, Bulent H. [Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garner, Frank A. [Radiation Effects Consulting, 2003 Howell Ave., Richland, WA 99354 (United States); Freyer, Paula D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Okita, Taira [Research Into Artifacts Dept., Center for Engineering, University of Tokyo, Tokyo (Japan); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

2014-05-01

The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. An effective tensile strain hardening exponent was also obtained from the data which shows a relative decrease in ductility of steel with increased irradiation damage. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

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

Effect of fast-neutron irradiation on plastic deformation of Type 304 stainless steel

International Nuclear Information System (INIS)

Yamada, H.

1978-01-01

Plastic deformation of EBR-II-irradiated Type 304 stainless steel was investigated by a stress-relaxation method. The stress-strain-rate relationships for the irradiated specimens at room temperature are concave upward, which are similar to those for the unirradiated specimens. However, concave downward behavior in the stress-strain-rate relationships were observed at much lower temperatures for the irradiated specimens in contrast to the unirradiated specimens. These results were analyzed succccessfully using Hart's mechanical equation-of-state concept. It was found that the hardness sigma*, which is the minimum stress necessary for the dislocation to overcome obstacles without thermal activation, increases linearly with fast-neutron fluence. This increase in sigma* is consistent with so-called ''irradiation hardening.'' In addition, resistance to dislocation glide, which is quantitatively measured in terms of sigma 0 , was observed to decrease linearly with fast-neutron fluence. The decrease in sigma 0 can be attributed to a decrease of solute drag due to irradiation-induced solute segregation

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

Helium diffusion in irradiated boron carbide

International Nuclear Information System (INIS)

Hollenberg, G.W.

1981-03-01

Boron carbide has been internationally adopted as the neutron absorber material in the control and safety rods of large fast breeder reactors. Its relatively large neutron capture cross section at high neutron energies provides sufficient reactivity worth with a minimum of core space. In addition, the commercial availability of boron carbide makes it attractive from a fabrication standpoint. Instrumented irradiation experiments in EBR-II have provided continuous helium release data on boron carbide at a variety of operating temperatures. Although some microstructural and compositional variations were examined in these experiments most of the boron carbide was prototypic of that used in the Fast Flux Test Facility. The density of the boron carbide pellets was approximately 92% of theoretical. The boron carbide pellets were approximately 1.0 cm in diameter and possessed average grain sizes that varied from 8 to 30 μm. Pellet centerline temperatures were continually measured during the irradiation experiments

Development of the flow control irradiation facility for JOYO

International Nuclear Information System (INIS)

Soroi, Masatoshi; Miyakawa, Shun-ichi

1998-05-01

This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor 'JOYO'. FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. We are going to freeze and review the FLORA project, taking into consideration the fuel development situation and the research project of JOYO MK-III core. (J.P.N.)

Tensile properties of vanadium alloys irradiated at <430{degrees}C

Energy Technology Data Exchange (ETDEWEB)

Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

1997-08-01

Recent attention to vanadium alloys has focused on significant susceptibility to loss of work-hardening capability in irradiation experiments at <430{degrees}C. An evaluation of this phenomenon was conducted on V-Ti, V-Cr-Ti, and V-Ti-Si alloys irradiated in several conventional and helium-charging irradiation experiments in the FFTF-MOTA, HFIR, and EBR-II. Work hardening capability and uniform tensile elongation appear to vary strongly from alloy and heat to heat. A strong heat-to-heat variation has been observed in V-4Cr-4Ti alloys tested, i.e., a 500-kg heat (No. 832665), a 100-kg heat (VX-8), and a 30-kg heat (BL-47). The significant differences in susceptibility to loss of work-hardening capability from one heat to another are estimated to correspond to a difference of {approx}100{degrees}C or more in minimum allowable operating temperature (e.g., 450 versus 350{degrees}C).

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.

Microbiological Quality of Mixed Vegetables Salad and Insuring its Safety by Irradiation

International Nuclear Information System (INIS)

Hammad, A.A.; Abu El-Nour, S.A.; Swailam, H.M.; Serag, M.S.; Mansour, F.A.

2008-01-01

Fifteen prepackaged mixed vegetables salad samples were collected from different local supermarkets. They were tested for their microbiological quality. TAPC of mix salad samples ranged from 2.5 x 10 5 to 7.0 x 10 7 cfu/g; LAB between 2.0 x 10 3 and 7.1 x 10 6 cfu/g; M and Y ranged from 1. 10 2 to 4.5 x 10 3 cfu/g. All tested mix salad samples contained coliform bacteria, E. coli and Ent. Faecalis. Staph. aureus was found in 12 (80%) samples, while A. hydrophila was found in all mix salad samples. L. monocytogenes was present in only two samples and Salmonella spp. was detected in only one sample. Mix salad samples were irradiated at 1, 2 and 3 kGy, and then stored at refrigeration temperature (4 degree C±1). Generally, all irradiation doses used reduced the initial TAPC, LAB and TM and Y and the reduction was proportional with irradiation dose. The optimum irradiation dose for irradiating mix salad was identified to be 3 kGy as no pathogens were detected in mix salad samples exposed to this irradiation dose. This irradiation dose had no adverse effect on physical and sensorial quality attributes of mix salad and extended its shelf-life to 15 days against only 7 days for unirradiated samples

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.)

RTNS-II irradiations and operations

International Nuclear Information System (INIS)

Logan, C.M.; Heikkinen, D.W.

1982-01-01

The objectives of this work are operation of RTNS-II (a 14-MeV neutron source facility), machine development, and support of the experimental program that utilizes this facility. Experimenter services include dosimetry handling, scheduling, coordination, and reporting. RTNS-II is dedicated to materials research for the fusion power program. Its primary use is to aid in the development of models of high-energy neutron effects. Such models are needed in interpreting and projecting to the fusion environment engineering data obtained in other neutron spectra. Irradiations were performed for a total of twenty-nine different experimenters during this quarter. A JOEL 200 CX TEM and other post-irradiation test equipment have been installed

Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

Energy Technology Data Exchange (ETDEWEB)

Pokor, C

2003-07-01

The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

Copper(II) Thiosemicarbazone Complexes and Their Proligands upon UVA Irradiation: An EPR and Spectrophotometric Steady-State Study.

Science.gov (United States)

Hricovíni, Michal; Mazúr, Milan; Sîrbu, Angela; Palamarciuc, Oleg; Arion, Vladimir B; Brezová, Vlasta

2018-03-21

X- and Q-band electron paramagnetic resonance (EPR) spectroscopy was used to characterize polycrystalline Cu(II) complexes that contained sodium 5-sulfonate salicylaldehyde thiosemicarbazones possessing a hydrogen, methyl, ethyl, or phenyl substituent at the terminal nitrogen. The ability of thiosemicarbazone proligands to generate superoxide radical anions and hydroxyl radicals upon their exposure to UVA irradiation in aerated aqueous solutions was evidenced by the EPR spin trapping technique. The UVA irradiation of proligands in neutral or alkaline solutions and dimethylsulfoxide (DMSO) caused a significant decrease in the absorption bands of aldimine and phenolic chromophores. Mixing of proligand solutions with the equimolar amount of copper(II) ions resulted in the formation of 1:1 Cu(II)-to-ligand complex, with the EPR and UV-Vis spectra fully compatible with those obtained for the dissolved Cu(II) thiosemicarbazone complexes. The formation of the complexes fully inhibited the photoinduced generation of reactive oxygen species, and only subtle changes were found in the electronic absorption spectra of the complexes in aqueous and DMSO solutions upon UVA steady-state irradiation. The dark redox activity of copper(II) complexes and proligand/Cu(II) aqueous solutions towards hydrogen peroxide which resulted in the generation of hydroxyl radicals, was confirmed by spin trapping experiments.

Microstructural examination of several commercial ferritic alloys irradiated to high fluence

International Nuclear Information System (INIS)

Gelles, D.S.

1981-01-01

Microstructural observations are reported for a series of five commercial ferritic alloys, 2 1/4 Cr-1 Mo, H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650 0 C and to a maximum fluence of 17.6 x 10 22 n/cm 2 (E > 0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400 0 C to 14.0 x 10 22 n/cm 2 . A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6 C, Mo 2 C, Chi, Laves, M 23 C 6 , α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation

Microstructural examination of several commercial ferritic alloys irradiated to high fluence

Science.gov (United States)

Gelles, D. S.

Microstructural observations are reported for a series of five commercial ferritic alloys, 2 {1}/{4}Cr-1Mo , H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650°C and to a maximum fluence of 1.76 × 10 23 n/cm 2 (E >0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400°C to 1.40 × 10 23 n/cm 2. A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6C, Mo 2C, Chi, Laves, M 23C 6, α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation.

Irradiations at RTNS-II

International Nuclear Information System (INIS)

Heikkinen, D.W.; Logan, C.M.

1982-01-01

The RTNS-II 14-MeV neutron source facility at Lawrence Livermore National Laboratory is described. Average neutron source parameters are outlined. A brief general description of the irradiation program to the present time is given. A short discussion of guidelines for prospective users is also given

Dielectric changes in neutron-irradiated rf window materials

International Nuclear Information System (INIS)

Frost, H.M.; Clinard, F.W. Jr.

1987-01-01

Ceramics used for windows in ECRH heating systems for magnetically-confined fusion reactors must retain adequate properties during and after intense neutron irradiation. Of particular concern is a decrease in transmissivity, a parameter inversely related to the product of dielectric constant K and loss tangent tanδ. Samples of polycrystalline Al 2 O 3 and BeO were irradiated to 1 x 10 26 n/m 2 at 660K in the EBR-II fission reactor, and the above properties subsequently measured at 95 GHz. It was found that ktanδ for both materials doubled, implying a doubling of thermal stresses and a consequent reduction of time-to-failure from an assumed one year to 20 min for beryllia and 2 s for alumina. In the case of BeO, a large increase in reflectance of the incident millimeter-wave power results from dielectrically uncompensated swelling. This phenomenon could significantly degrade source performance

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

Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

1993-12-01

A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

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

Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors

International Nuclear Information System (INIS)

Pokor, C.

2003-01-01

The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

Microstructural interpretation of the fluence and temperature dependence of the mechanical properties of irradiated AISI 316

International Nuclear Information System (INIS)

Johnson, G.D.; Garner, F.A.; Brager, H.R.; Fish, R.L.

1980-01-01

The effects of neutron irradiation on the mechanical properties of annealed and 20% cold-worked AISI 316 irradiated in EBR-II were determined for the temperature regime of 370 to 760 0 C for fluences up to 8.4 x 10 22 n/cm 2 (E > 0.1 MeV). At irradiation temperatures below about 500 0 C, both annealed and cold-worked material exhibit a substantial increase in the flow stress with increasing fluence. Furthermore, both materials eventually exhibit the same flow stress, which is independent of fluence. At temperatures in the range of 538 to 650 0 C, the cold-worked material exhibits a softening with increasing fluence. Annealed AISI 316 in this temperature regime exhibits hardening and at a fluence of 2 to 3 x 10 22 n/cm 2 (E > 0.1 MeV) reaches the same value of flow stress as the cold-worked material

Neutron irradiation damage in Al2O3 and Y2O3

International Nuclear Information System (INIS)

Clinard, F.W. Jr.; Bunch, J.M.; Ranken, W.A.

1975-01-01

Two ceramics under consideration for use in fusion reactors, Al 2 O 3 and Y 2 O 3 , were irradiated in the EBR-II fission reactor at 650, 875, and 1025 0 K to fluences between 2 and 6 x 10 21 n/cm 2 (E greater than 0.1 MeV). Samples evaluated include sapphire, Lucalox, alumina, Y 2 O 3 , and Y 2 O 3 -10 percent ZrO 2 (Yttralox). All Al 2 O 3 specimens swelled significantly (1 to 3 percent), with most of the growth observed in sapphire along the c-axis at the higher temperatures. Al 2 O 3 samples irradiated at 875 to 1025 0 K contained a high density of small aligned ''pores''. Irradiated Y 2 O 3 -based ceramics exhibited dimensional stability and a defect content consisting primarily of unresolved damage and/or dislocation loops. The behavior of these ceramics under irradiation is discussed, and the relevance of fission neutron damage studies to fusion reactor applications is considered. (auth)

Additive action model for mixed irradiation

International Nuclear Information System (INIS)

Lam, G.K.Y.

1984-01-01

Recent experimental results indicate that a mixture of high and low LET radiation may have some beneficial features (such as lower OER but with skin sparing) for clinical use, and interest has been renewed in the study of mixtures of high and low LET radiation. Several standard radiation inactivation models can readily accommodate interaction between two mixed radiations, however, this is usually handled by postulating extra free parameters, which can only be determined by fitting to experimental data. A model without any free parameter is proposed to explain the biological effect of mixed radiations, based on the following two assumptions: (a) The combined biological action due to two radiations is additive, assuming no repair has taken place during the interval between the two irradiations; and (b) The initial physical damage induced by radiation develops into final biological effect (e.g. cell killing) over a relatively long period (hours) after irradiation. This model has been shown to provide satisfactory fit to the experiment results of previous studies

Comparison of compression properties and swelling of beryllium irradiated at various temperatures

Energy Technology Data Exchange (ETDEWEB)

Beeston, J.M.; Miller, L.G.; Wood, E.L. Jr.; Moir, R.W.

1983-01-01

A beryllium cylinder irradiated in Experimental Breeder Reactor (EBR-II) for four years at 700 to 760 K to a neutron fluence of 8.13 x 10/sup 22/ n/cm/sup 2/ (total) or 1 x 10/sup 22/ n/cm/sup 2/ (E > 1 MeV) was cut into samples and tested. Yield strength and plastic strain was determined in compression tests at 300, 723, 823 K and after annealing at 1173 K for one hour. The immersion density and helium content were measured on samples. An equation for swelling was derived from the data by regression analysis. The microstructure showed agglomeration of helium in voids or bubbles at the grain boundaries.

Commercial Applications at FRM II Based on Neutron Irradiations

Energy Technology Data Exchange (ETDEWEB)

Gerstenberg, H.; Draack, A.; Kastenmuller, A. [Technische Universitaet Muenchen, Munchen (Germany)

2013-07-01

Due to its design as a heavy water moderated reactor with a very compact core FRM II, Germany's most modern and most powerful research reactor, offers excellent conditions for basic research using beam tubes. On the other hand it is equipped with various irradiation facilities to be used mainly for industrial purposes. From the very beginning of reactor operation a dedicated department had been implemented in order to provide a neutron irradiation service to interested parties on a commercial basis. As of today the most widely used application is Si doping. The semiautomatic doping facility accepts ingots with diameters between 125 mm and 200 mm and a maximum height of 500 mm. The irradiation channel is located deep in the heavy water tank and exhibits a ratio of thermal/fast neutron flux density of > 1000. This value allows the doping of Si to a target resistivity as high as 1100 Ωcm within the tight limits regarding accuracy and homogeneity specified by the customer. Typically the throughput of Si doped in FRM II sums up to about 15 t/year. Another topic of growing importance is the use of FRM II aiming the production of radioisotopes mainly for the radiopharmaceutical industry. The maybe most challenging example is the production of Lu-177 n. c. a. based on the irradiation of Yb{sub 2}O{sub 3} to a high fluence of thermal neutrons of typically 1.5E20 cm{sup -2}. The Lu-177 activity delivered to the customer is in the range of 750 GBq. With respect to further processing it turned out to be a highly advantageous to have the laboratories of ITG, the company extracting the Lu-177 from the freshly irradiated Yb{sub 2}O{sub 3} on site FRM II. Further irradiation facilities are available at FRM II in order to allow the activation of samples for analytical purposes or to irradiate samples for geochronological investigations using the fission track technique. Finally a project on the future installation of a facility dedicated to the irradiation of U-targets for

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

Correlation of fracture toughness with tensile properties for irradiated 20% cold-worked 316 stainless steel

International Nuclear Information System (INIS)

Hamilton, M.L.; Garner, F.A.; Wolfer, W.G.

1983-08-01

A correlation has been developed which allows an estimate to be made of the toughness of austenitic alloys using more easily obtained tensile data. Tensile properties measured on 20% cold-worked AISI 316 specimens made from ducts and cladding irradiated in EBR-II were used to predict values for the plane strain fracture toughness according to a model originally developed by Krafft. Some microstructural examination is required to determine a parameter designated as the process zone size. In contrast to the frequently employed Hahn-Rosenfeld model, this model gives results which agree with recent experimental determinations of toughness performed in the transgranular failure regime

Irradiation induced precipitation in tungsten based, W-Re alloys

Science.gov (United States)

Williams, R. K.; Wiffen, F. W.; Bentley, J.; Stiegler, J. O.

1983-03-01

Tungsten-base alloys containing 5, 11, and 25 pct Re were irradiated in the EBR-II reactor. Irradiation temperatures ranged from 600 to 1500 °C. All compositions were irradiated to fluences in the range 4.3 to 6.1 X 1025 n/m2 (E > 0.1 MeV), and three 25 pct Re samples were also irradiated to 3.7 X 1026 n/m2 at temperatures 700 to 900 °C. Postirradiation examination included measurement of electrical resistivity at room temperature and lower temperatures, X-ray diffraction, optical metallography, microprobe analysis, and transmission electron microscopy. Irradiation induced resistivity decreases observed in most of the samples suggested second-phase precipitation. Complete results confirmed the precipitate formation in all samples, in disagreement with existing phase diagrams for the W-Re system. Electron diffraction showed the precipitates to be consistent with the cubic, Re-rich X-phase and inconsistent with the σ-phase. Large variations in precipitate morphology and distribution were observed between the different compositions and irradiation conditions. For the 5 and 11 pct Re-alloys, spherically symmetric strain fields surrounded the equiaxed precipitate particles, and were observed even where no particles were visible. These strain fields are believed to arise from local Re enrichment. Thermoelectric data show that the precipitation can lead to decalibration of W/Re thermocouples.

Irradiation mixing of Al into U3Si

International Nuclear Information System (INIS)

Birtcher, R.C.; Ding, F.R.; Kestel, B.J.; Baldo, P.M.; Zaluzec, N.J.

1995-11-01

Thermal and irradiation induced intermixing of uranium silicide reactor fuels with the aluminum cladding is an important consideration in understanding their fission gas and fuel swelling behavior. The authors have used Rutherford backscattering to follow the behavior of an Al thin film on U 3 Si and U 3 Si 2 during 1.5 MeV Kr ion irradiation at temperatures of 30 and 350 C. After an initial dose during which no intermixing occurs, the Al mixes quickly into U 3 Si. The threshold dose is believed to be associated with an oxide layer between the Al and the uranium silicide. At 300 C and doses greater than threshold, rates of mixing and aluminide phase growth are extracted

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

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

The influence of nickel content on microstructures of Fe-Cr-Ni austenitic alloys irradiated with nickel ions

International Nuclear Information System (INIS)

Muroga, T.; Yoshida, N.; Garner, F.A.

1990-11-01

The objectives of this effort is to identify the mechanisms involved in the radiation-induced evolution of microstructure in materials intended for fusion applications. The results of this study are useful in interpreting the results of several other ongoing experiments involving either spectral or isotopic tailoring to study the effects of helium on microstructure evolution. Ion-irradiated Fe-15Cr-XNi (X = 20, 35, 45, 60, 75) ternary alloys and a 15Cr-85Ni binary alloy were examined after bombardment at 675 degree C and compared to earlier observations made on these same alloys after irradiation in EBR-II at 510 or 538 degree C. The response of the ion-irradiated microstructures to nickel content appears to be very consistent with that of neutron irradiation even though there are four orders of magnitude difference in displacement rate and over 200 degree C difference in temperature. It appears that the transition to higher rates of swelling during both types of irradiation is related to the operation of some mechanisms that is not directly associated with void nucleation. 6 refs., 8 figs

Estimates of time-dependent fatigue behavior of type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

International Nuclear Information System (INIS)

Brinkman, C.R.; Liu, K.C.; Grossbeck, M.L.

1979-01-01

Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irraidated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63 x 10 26 neutrons (n)/m 2 E > 0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20% cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadins ranging from 2 to 5 MW/m 2 were used. Results, although conjectural because of the many assumptions, tended to show that 20% cold-worked Type 316 stainless steel could be used as a first-wall material meeting a 7.5 go 8.5 MW-year/m 2 lifetime goal provided the neutron wall loading does not exceed more than about 2 MW/m 2 . These results were obtained for an air environment, ant it is expected that the actual vacuum environment will extend lifetime beyond 10 MW-year/m 2

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

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.

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

Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J.; Alexander, D.J.; Robertson, J.P. [Oak Ridge National Lab., TN (United States)] [and others

1997-04-01

Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimens were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.

Neutron irradiation creep in stainless steel alloys

Energy Technology Data Exchange (ETDEWEB)

Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

1994-09-01

Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300 C and 500 C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of primary'' creep stage is observed for doses up to 3-5 dpa after which dose the secondary'' creep stage begins. The primary'' creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These primary'' creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of [alpha]-ferrite below about 400 C and of carbides below about 700 C, and not to irradiation creep. The secondary'' creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature (Q[sub irr]=0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels. ((orig.))

Safety analysis of a pool Genesis II irradiator

International Nuclear Information System (INIS)

Rodrigues Junior, Ary de A.

2011-01-01

The Genesis II Irradiator manufactured by GRAT * STAR Inc. (USA) is a category III gamma irradiator in which the sealed source is contained in a water filled storage pool and is shielded permanently, i.e. the material has to move down to the source. Even though the pool is 5.6 m deep, what would happen if the water level lowered? There are a series of safety devices that will avoid this situation and calculations show that the water level has to be very low in order to deliver a significant dose; moreover, only in case a person remains at the border of the pool for a long time this would be risky. In conclusion, it is practically impossible for someone to be exposed to radiation from a Genesis II Irradiator source. (author)

Effect of helium irradiation on fracture modes

International Nuclear Information System (INIS)

Hanamura, T.; Jesser, W.A.

1982-01-01

The objective of this work is to determine the crack opening mode during in-situ HVEM tensile testing and how it is influenced by test temperature and helium irradiation. Most cracks were mixed mode I and II. However, between 250 0 C and room temperature the effect of helium irradiation is to increase the amount of mode I crack propagation. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in irradiated specimens tested between 250 0 C and room temperature, but could be restored by a post irradiation anneal

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

Energy Technology Data Exchange (ETDEWEB)

Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-15

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

Rheological evaluation of the irradiated pectin/gelatin mixed systems

International Nuclear Information System (INIS)

Inamura, Patricia Y.; Mastro, Nelida L. del

2011-01-01

The main biopolymers used in the edible films production are polysaccharides and proteins. Pectin is a heterosaccharidic polymer derived from the vegetal cell wall. Gelatin is a heterogeneous mixture of water-soluble proteins of high average molecular mass derived by hydrolytic action from animal collagen. The aim of this research was to evaluate the effect of ionizing radiation on either the biopolymers alone or on the mixed systems prepared with high-and low-methoxyl pectin and gelatin in solution and mixed gel. The results showed that gelatin viscosity remained almost unaffected by the irradiation with doses from 1 to 15 kGy, with a slight increase at 3 kGy. On the other hand, there was a sharp decrease of viscosity values of all pectin solutions upon irradiation, being this behavior predominant when both polysaccharides and proteins were present in a mixed system. The gel hardness and gel brittleness of the gelatin were affected by the increase of radiation dose. (author)

Experience of the irradiation method under mixed gas (95% O2 plus 5% CO2) inhalation

International Nuclear Information System (INIS)

Ikeda, Michio; Tazaki, Eio

1978-01-01

The irradiation method under mixed gas of 95% O 2 plus CO 2 inhalation at one atomosphere was discussed to improve therapeutic results, in malignant tumors which are not greatly sensitive to irradiation. Randomized study was done in each attending institute, with common protocols. As a result, no positive effect was recognized in irradiation method under mixed gas inhalation with daily dose of 200 rad and 5 fractions per week, which is widely used clinically. But when irradiation dose was increased up to 500 to 600 rad per fraction, effect of the mixed gas was remarkable. But against this, observing for years, results in irradiation under mixed gas inhalation were not always related to the improvement of the long term survival. (author)

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

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)

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

International Nuclear Information System (INIS)

Strain, R.V.; Johnson, C.E.

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

Energy Technology Data Exchange (ETDEWEB)

Strain, R V; Johnson, C E

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

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.

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.

Consequences of metallic fuel-cladding liquid phase attack during over-temperature transient on fuel element lifetime

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Seidel, B.R.

1990-01-01

Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs

Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

International Nuclear Information System (INIS)

Gilbert, E.R.

1976-04-01

Irradiation creep studies with pressurized tubes of 20 percent cold worked Type 316 stainless steel were conducted in EBR-2. Results showed that as atom displacements are extended above 5 dpa and temperatures are increased above 375 0 C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation-induced effective creep strains up to 1.8 percent were observed without specimen failure. 13 figures

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

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

Thermal Decompositon Studies Of Pre-Irradiated Nickel (II) Azides ...

African Journals Online (AJOL)

The effect of pre-irradiation on the thermal decomposition of three samples of nickel (II) azide was studied. It was found that the rates of thermal decomposition of Ni(OH)N3 increased substantially with increase in pre-irradiation dosage. The initial reaction rates change from time-dependant nucleation law for the unirradiated ...

Fractographic examination of HT-9 and 9Cr-1Mo Charpy specimens irradiated in the AD-2 test

International Nuclear Information System (INIS)

Gelles, D.S.; Hu, W.L.

1983-01-01

Fracture surface topologies have been examined using scanning electron microscopy for 20 selected half sized Charpy impact specimens of HT-9 and Modified 9Cr-1Mo in order to provide improved understanding of fracture toughness degradation as a result of irradiation for Path E alloys. The specimen matrix included unirradiated specimens and specimens irradiated in EBR-II in the AD-2 experiment. Also, hardness measurements have been made on selected irradiated Charpy specimens. The results of examinations indicate that irradiation hardening due to G-phase formation at 390 0 C is responsible for the large shift in ductile-to-brittle transition temperature (DBTT) found in HT-9. Toughness degradation in HT-9 observed following higher temperature irradiations is attributed to precipitation at delta ferrite stringers. Reductions in toughness as a consequence of irradiation in Modified 9Cr-1Mo are attributed to in-reactor precipitation of (V,Nb)C and M 23 C 6 . It is shown that crack propagation rates for ductile and brittle failure modes can be measured, that they differ by over an order of magnitude and that unexpected multiple shifts in fracture mode from ductile to brittle failure can be attributed to the effect of delta ferrite stringers on crack propagation rates

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.

Irradiation effects on reactor structural materials. Semi-annual progress report, August 1974--February 1975

International Nuclear Information System (INIS)

Claudson, T.T.

1975-03-01

Data are reported on: effects of cold work on creep-fatigue of irradiated 304 and 316 stainless steel (ss); swelling of 304 and 316 ss irradiated with protons and fast neutrons; effects of hold time on fatigue crack propagation in neutron-irradiated 20 percent cold-worked 316 ss; radiation resistance of 0.03 percent Cu A533-B steel; microstructure of irradiated Inconel 718, Incoloy 800, PH13-8Mo, Mo, and Nb; dose dependence of 2.8-MeV Ni + ion damage (swelling) in Ni; notch ductility and strength of 316 ss submerged arc weld deposits; effects of microstructure of 316 ss on its irradiation response; in-reactor deformation of 20 percent cold-worked 316 ss; microstructure of HFIR-irradiated 316 ss; void microstructures of V bombarded by 46-MeV Ni 6+ ions (with and without preinjected helium) or 7.5-MeV Ta 3+ ions; swelling of Mo, Mo--0.5 Ti, Nb, Nb--1 Zr, W, and W--25 Re after fast neutron irradiation; swelling of V ion-irradiated Mo; creep of 20 percent cold-worked 316 ss at 850, 1000, and 1100 0 F; effects of fast neutrons on mechanical properties of 20 percent cold-worked 316 ss; notch effects in tensile behavior of irradiated, annealed 304 ss (EBR-II duct thimbles); equations for thermal creep in pressurized tubes of 20 percent cold-worked 316 ss; irradiation creep in cold-worked 316 ss; helium production cross sections in neutron-irradiated elements; and radiation effects on various alloys. (U.S.)

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

International Nuclear Information System (INIS)

Pahl, R.G.; Wisner, R.S.; Billone, M.C.; Hofman, G.L.

1990-01-01

Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs

Investigation of the interaction of copper(II) oxide and electron beam irradiation crosslinkable polyethylene

International Nuclear Information System (INIS)

Bee, Soo-Tueen; Sin, Lee Tin; Ratnam, C.T.; Haraveen, K.J.S.; Tee, Tiam-Ting; Rahmat, A.R.

2015-01-01

In this study, the effects of electron beam irradiation on the properties of copper(II) oxide when added to low-density polyethylene (LDPE) blends were investigated. It was found that the addition of low loading level of copper(II) oxide (⩽2 phr) to LDPE results in significantly poorer gel content and hot set results. However, the incorporation of higher loading level of copper(II) oxide (⩾3 phr) could slightly increase the degree of crosslinking in all irradiated LDPE composites. This is due to the fact that higher amounts of copper(II) oxide could slightly induce the formation of free radicals in LDPE matrix. Besides, increasing irradiation doses was also found to gradually increase the gel content of LDPE composites by generating higher amounts of free radicals. As a consequence, these higher amounts of free radicals released in the LDPE matrix could significantly increase the degree of crosslinking. The addition of copper(II) oxide could reduce the tensile strength and fracture strain (elongation at break) of LDPE composites because of poorer interfacial adhesion effect between copper(II) oxide particles and LDPE matrix. Meanwhile, increasing irradiation doses on all copper(II) oxide added LDPE composites could marginally increase the tensile strength. In addition, increasing irradiation dose could enhance the thermal stability of LDPE composites by increasing the decomposition temperature. The oxidation induction time (OIT) analysis showed that, because of the crosslinking network in the copper(II) oxide added LDPE composites, oxidation reaction is much delayed.

Investigation of the interaction of copper(II) oxide and electron beam irradiation crosslinkable polyethylene

Energy Technology Data Exchange (ETDEWEB)

Bee, Soo-Tueen, E-mail: direct.beest@gmail.com [Department of Chemical Engineering, Lee Kong Chian Faculty of Engineering and Science, Universiti Tunku Abdul Rahman, Jalan Sungai Long, Bandar Sungai Long, Cheras, 43000 Kajang, Selangor (Malaysia); Sin, Lee Tin, E-mail: direct.tinsin@gmail.com [Department of Chemical Engineering, Lee Kong Chian Faculty of Engineering and Science, Universiti Tunku Abdul Rahman, Jalan Sungai Long, Bandar Sungai Long, Cheras, 43000 Kajang, Selangor (Malaysia); Ratnam, C.T. [Radiation Processing Technology Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia); Haraveen, K.J.S.; Tee, Tiam-Ting [Department of Chemical Engineering, Lee Kong Chian Faculty of Engineering and Science, Universiti Tunku Abdul Rahman, Jalan Sungai Long, Bandar Sungai Long, Cheras, 43000 Kajang, Selangor (Malaysia); Rahmat, A.R. [Department of Polymer Engineering, Faculty of Chemical Engineering, Universiti Teknologi Malaysia, 81310 UTM Skudai, Johor (Malaysia)

2015-10-01

In this study, the effects of electron beam irradiation on the properties of copper(II) oxide when added to low-density polyethylene (LDPE) blends were investigated. It was found that the addition of low loading level of copper(II) oxide (⩽2 phr) to LDPE results in significantly poorer gel content and hot set results. However, the incorporation of higher loading level of copper(II) oxide (⩾3 phr) could slightly increase the degree of crosslinking in all irradiated LDPE composites. This is due to the fact that higher amounts of copper(II) oxide could slightly induce the formation of free radicals in LDPE matrix. Besides, increasing irradiation doses was also found to gradually increase the gel content of LDPE composites by generating higher amounts of free radicals. As a consequence, these higher amounts of free radicals released in the LDPE matrix could significantly increase the degree of crosslinking. The addition of copper(II) oxide could reduce the tensile strength and fracture strain (elongation at break) of LDPE composites because of poorer interfacial adhesion effect between copper(II) oxide particles and LDPE matrix. Meanwhile, increasing irradiation doses on all copper(II) oxide added LDPE composites could marginally increase the tensile strength. In addition, increasing irradiation dose could enhance the thermal stability of LDPE composites by increasing the decomposition temperature. The oxidation induction time (OIT) analysis showed that, because of the crosslinking network in the copper(II) oxide added LDPE composites, oxidation reaction is much delayed.

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

Manganese(II), iron(II), and mixed-metal metal-organic frameworks based on chains with mixed carboxylate and azide bridges: magnetic coupling and slow relaxation.

Science.gov (United States)

Wang, Yan-Qin; Yue, Qi; Qi, Yan; Wang, Kun; Sun, Qian; Gao, En-Qing

2013-04-15

Mn(II) and Fe(II) compounds derived from azide and the zwitterionic 1-carboxylatomethylpyridinium-4-carboxylate ligand are isomorphous three-dimensional metal-organic frameworks (MOFs) with the sra net, in which the metal ions are connected into anionic chains by mixed (μ-1,1-azide)bis(μ-carboxylate) triple bridges and the chains are cross-linked by the cationic backbones of the zwitterionic ligands. The Mn(II) MOFs display typical one-dimensional antiferromagnetic behavior. In contrast, with one more d electron per metal center, the Fe(II) counterpart shows intrachain ferromagnetic interactions and slow relaxation of magnetization attributable to the single-chain components. The activation energies for magnetization reversal in the infinite- and finite-chain regimes are Δτ1 = 154 K and Δτ2 = 124 K, respectively. Taking advantage of the isomorphism between the Mn(II) and Fe(II) MOFs, we have prepared a series of mixed-metal Mn(II)(1-x)Fe(II)(x) MOFs with x = 0.41, 0.63, and 0.76, which intrinsically feature random isotropic/anisotropic sites and competing antiferromagnetic-ferromagnetic interactions. The materials show a gradual antiferromagnetic-to-ferromagnetic evolution in overall behaviors as the Fe(II) content increases, and the Fe-rich materials show complex relaxation processes that may arise for mixed SCM and spin-glass mechanisms. A general trend is that the activation energy and the blocking temperature increase with the Fe(II) content, emphasizing the importance of anisotropy for slow relaxation of magnetization.

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics

International Nuclear Information System (INIS)

Senor, D.J.; Youngblood, G.E.; Moore, C.E.; Trimble, D.J.; Woods, J.J.

1996-06-01

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation of irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD β-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics

International Nuclear Information System (INIS)

Senor, D.J.; Youngblood, G.E.; Moore, C.E.; Trimble, D.J.; Woods, J.J.

1997-05-01

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation after irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD β-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination

On the ductile-to-brittle transition behavior of martensitic alloys neutron irradiated to 26 dpa

International Nuclear Information System (INIS)

Hu, W.L.; Gelles, D.S.

1987-01-01

Charpy impact tests were conducted on specimens made of HT-9 and 9Cr-1Mo in various heat treatment conditions which were irradiated in EBR-II to 26 dpa at 390 to 500 0 C. The results are compared with previous results on specimens irradiated to 13 dpa. HT-9 base metal irradiated at low temperatures showed a small additional increase in ductile brittle transition temperature and a decrease in upper shelf energy from 13 to 26 dpa. No fluence effect was observed in 9Cr-1Mo base metal. The 9Cr-1Mo weldment showed degraded DBTT but improved USE response compared to base metal, contrary to previous findings on HT-9. Significant differences were observed in HT-9 base metal between mill annealed material and normalized and tempered material. The highest DBTT for HT-9 alloys was 50 0 C higher than for the worst case in 9Cr-1Mo alloys. Fractography and hardness measurements were also obtained. Significant differences in fracture appearance were observed in different product forms, although no dependence on fluence was observed. Failure was controlled by the preirradiation microstructure

Response to annealing and reirradiation of AISI 304L stainless steel following initial high-dose neutron irradiation in EBR-II

International Nuclear Information System (INIS)

Porter, D.L.; McVay, G.L.; Walters, L.C.

1980-01-01

The object of this study was to measure the stability of irradiation-induced microstructure upon annealing and, by selectively annealing out some of these features and reirradiating the material, it was expected that information could be gained concerning the role of microstructural changes in the void swelling process. Transmission electron microscopic examinations of isochronally annealed (200 to 1050 0 C) AISI 304L stainless steel, which had been irradiated at approximately 415 0 C to a fast (E > 0.1 MeV) neutron fluence of approximately 5.1 x 10 26 n/m 2 , verified that the two-stage hardness recovery with temperatures was related to a low temperature annealing of dislocation structures and a higher temperature annealing of voids and solute redistribution

Helium production in mixed spectrum reactor-irradiated pure elements

International Nuclear Information System (INIS)

Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

1986-01-01

The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

The adsorption of Sr(II) and Cs(I) ions by irradiated Saccharomyces cerevisiae

International Nuclear Information System (INIS)

Yiming Tan; Jundong Feng; Liang Qiu; Zhentian Zhao; Xiaohong Zhang; Haiqian Zhang

2017-01-01

Adsorption behavior and mechanism of Sr(II) and Cs(I) in single and binary solutions using irradiated Saccharomyces cerevisiae was investigated. The effects of several environmental factors on Sr(II) and Cs(I) adsorption to irradiated Saccharomyces cerevisiae was determined. The equilibrium experimental data were simulated by different kinetic models and isotherm models. The combined effect of Sr(II) and Cs(I) on Saccharomyces cerevisiae is generally antagonistic. SEM and EDS analyses indicate that crystals formed on the cell surface are precipitate of Sr(II) and Cs(I), respectively. (author)

Irradiation effects on low-friction coatings for LMFBR applications

International Nuclear Information System (INIS)

Ward, A.L.; Johnson, R.N.; Guthrie, G.L.; Aungst, R.C.

1975-11-01

A variety of wear-resistant low-friction materials has been irradiated in the EBR-II in order to assess their reponse to LMFBR environments. Pre- and postirradiation testing and examination efforts have concentrated on candidate materials for application to the wear pads on FTR ducts (fuel, control, and reflector assemblies), and a significant result has been qualification of a proprietary detonation-gun-applied chromium carbide coating which employs a Ni Cr binder. Additional materials such as Inconel-718, Haynes-273, aluminides, and various chromium carbide/binder combinations, and other application processes such as plasma-spray, weld-overlays, diffusion bonding and explosive bonding, have also been studied. The most detailed examinations were conducted on selected chromium carbide coatings and included visual inspection, weight and dimensional measurements, metallography, electron microprobe, epoxy-lift-off, and x-ray diffraction analysis. Chromium carbide coatings applied by the detonation-gun process have demonstrated a marked superiority to those applied by plasma-spray techniques

Fuel-cladding chemical interaction in mixed-oxide fuels

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, J.W.; Devary, J.L.

1978-10-01

The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

Irradiation creep of the mixed oxide UPuO2

International Nuclear Information System (INIS)

Combette, Patrick; Milet, Claude

1976-01-01

The irradiation creep under compression of the mixed oxide UO 2 -PuO 2 was studied up to fission yields of 6x10 13 fcm -3 s -1 , under stresses -2 , in the temperature range 700-900 deg C. The creep rate is proportional to the applied stress and fission yield, athermal in the studied temperature range and non-dependent of burnup (up to 30000MWjt -1 ). In a sample under compression, swelling is observed due to the formation of fission products during the irradiation and the swelling rate is of the same order that in a cladded fuel element [fr

Fission-fusion correlations for swelling and microstructure in stainless steels: effect of the helium-to-displacement-per-atom ratio

International Nuclear Information System (INIS)

Odette, G.R.; Maziaz, P.J.; Spitznagel, J.A.

1981-01-01

The initial irradiated structural materials data base for fusion applications will be developed in fission reactors. Hence, this data may need to be adjusted using physically-based procedures to represent behavior in fusion environments, viz. - fission-fusion correlations. Such correlation should reflect a sound mechanistic understanding, and be verified in facilities which most closely simulate fusion conditions. In this paper we review the effects of only one of a number of potentially significant damage variables, the helium to displacement per atom ratio, on microstructural evolution in austenitic stainless steels. Dual-ion and helium preinjection data are analyzed to provide mechanistic guidance; these results appear to be qualitatively consistent with a more detailed comparison made between fast (EBR-II) and mixed (HFIR) spectrum neutron data for a single heat of 20% cold-worked 316 stainless steel. These two fission environments bound fusion (He/dpa ratios. A model calibrated to the fission reactor data is used to extrapolate to fusion conditions. Both the theory and broad empirical observation suggest that helium to dpa ratios have both a qualitative and quantitative influence on microstructural evolution; and that the very high and low ratios found in HFIR and EBR-II may not result in behavior which brackets intermediate fusion conditions

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

Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

Energy Technology Data Exchange (ETDEWEB)

Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

2002-12-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

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

SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

International Nuclear Information System (INIS)

Dutt, D.S.; Baker, R.B.

1975-06-01

The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

The effect of low dose rate irradiation on the swelling of 12% cold-worked 316 stainless steel

International Nuclear Information System (INIS)

Allen, T. R.

1999-01-01

In pressurized water reactors (PWRs), stainless steel components are irradiated at temperatures that may reach 400 C due to gamma heating. If large amounts of swelling (>10%) occur in these reactor internals, significant swelling related embrittlement may occur. Although fast reactor studies indicate that swelling should be insignificant at PWR temperatures, the low dose rate conditions experienced by PWR components may possibly lead to significant swelling. To address these issues, JNC and ANL have collaborated to analyze swelling in 316 stainless steel, irradiated in the EBR-II reactor at temperatures from 376-444 C, at dose rates between 4.9 x 10 -8 and 5.8 x 10 -7 dpa/s, and to doses of 56 dpa. For these irradiation conditions, the swelling decreases markedly at temperatures less than approximately 386 C, with the extrapolated swelling at 100 dpa being around 3%. For temperatures greater than 386 C, the swelling extrapolated to 100 dpa is around 9%. For a factor of two difference in dose rate, no statistically significant effect of dose rate on swelling was seen. For the range of dose rates analyzed, the swelling measurements do not support significant (>10%) swelling of 316 stainless steel in PWRs

Investigation of irradiated rats DNA in the presence of Cu(II) chelates of amino acids Schiff bases.

Science.gov (United States)

Karapetyan, N H; Torosyan, A L; Malakyan, M; Bajinyan, S A; Haroutiunian, S G

2016-01-01

The new synthesized Cu(II) chelates of amino acids Schiff bases were studied as a potential radioprotectors. Male albino rats of Wistar strain were exposed to X-ray whole-body irradiation at 4.8 Gy. This dose caused 30% mortality of the animals (LD30). The survival of animals exposed to radiation after preliminary administration of 10 mg/kg Cu(II)(Nicotinyl-L-Tyrosinate)2 or Cu(II)(Nicotinyl-L-Tryptophanate)2 prior to irradiation was registered about 80 and 100% correspondingly. Using spectrophotometric melting and agarose gel electrophoresis methods, the differences between the DNA isolated from irradiated rats and rats pretreated with Cu(II) chelates were studied. The fragments of DNA with different breaks were revealed in DNA samples isolated from irradiated animals. While, the repair of the DNA structure was observed for animals pretreated with the Cu(II) chelates. The results suggested that pretreatment of the irradiated rats with Cu(II)(Nicotinyl-L-Tyrosinate)2 and Cu(II)(Nicotinyl-L-Tryptophanate)2 compounds improves the liver DNA characteristics.

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

Swelling of austenitic iron-nickelchromium ternary alloys during fast neutron irradiation

International Nuclear Information System (INIS)

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

1984-01-01

Swelling data are now available for 15 iron-nickel-chromium ternary alloys irradiated to exposures as high as 110 displacements per atom (dpa) in Experimental Breeder Reactor-II (EBR-II) between 400 and 650 0 C. These data confirm trends observed at lower exposure levels and extend the generality of earlier conclusions to cover a broader range of composition and temperature. It appears that all austenitic iron-nickel-chromium ternary alloys eventually approach an intrinsic swelling rate of about1%/dpa over a range of temperature even wider than studied in this experiment. The duration of the transient regime that precedes the attainment of this rate is quite sensitive to nickel and chromium content, however. At nickel and chromium levels typical of 300 series steels, swelling does not saturate at engineering-relevant levels. However, there appears to be a tendency toward saturation that increases with declining temperature, increasing nickel and decreasing chromium levels. Comparisons of these results are made with those of similar studies conducted with charged particles. Conclusions are then drawn concerning the validity of charged particle simulation studies to determine the compositional and temperature dependence of swelling

Reassessment of the swelling behavior of AISI 304 stainless steel

International Nuclear Information System (INIS)

Garner, F.A.; Porter, D.L.

1982-03-01

Published swelling data derived from EBR-II irradiations of AISI 304 and 304L have been reanalyzed in light of insights gained from irradiation of AISI 316 and Fe-15Cr-25Ni. The primary influence of temperature, displacement rate and compositional variations in the 300 series stainless steels lies in the duration of the transient regime of swelling and not in the steady-state or constant swelling rate regime

Irradiation devices at the upgraded research reactor BER II

International Nuclear Information System (INIS)

Gawlik, D.; Robertson, T.

1992-06-01

An overview is given of those properties of the BER II research reactor which are important for carrying out irradiation experiments. The subsequent section describes the irradiation devices currently installed in the reactor, or which are under construction, and some of the experiments which can be conducted using them. The field of application of these experiments extends from the study of the metabolism of trace elements in man, employing a highly sensitive element analysis, via radiation damage of high-tech materials, to the identification of paintings of the old masters. The report concludes with a review of the technical details of the irradiation devices, giving information of interest for potential users. (orig.)

Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

International Nuclear Information System (INIS)

Brant Pinheiro, R.

1993-01-01

Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

Synthesis of Novel Polymeric Resins by Gamma Irradiation for Separation of In(III) ions from Cd(II) in Aqueous Media

International Nuclear Information System (INIS)

Massoud, A.; Abou El-Nour, F.; Killa, H.

2012-01-01

In this work, Zn(II)polymethacrylates and poly(acrylamide-acrylic acid) were prepared by gamma irradiation polymerization technique of the corresponding monomer at 30 kGy. The polymeric resins were mixed with Indium ions to determine its capacity in aqueous solutions using batch experiment. The adsorption efficiency of obtained polymeric resins toward In(III) and Cd(II) in different experimental conditions was established. Batch and column methods were applied for separation of indium and cadmium. The effects of various eluants such as H 2 SO 4 , NH 4 NO 3 , HNO 3 and HCl on the recovery of both metal ions were studied. The polymeric resins may be regenerated using 3M HCl solutions.

Irradiation-induced creep in 316 and 304L stainless steels

International Nuclear Information System (INIS)

Walters, L.C.; McVay, G.L.; Hudman, G.D.

1977-01-01

Recent results are presented from the in-reactor creep experiments that are being conducted by Argonne National Laboratory. The experiments consist of four subassemblies that contain helium-pressurized as well as unstressed capsules of 316 and 304L stainless steels in several metallurgical conditions. Experiments are being irradiated in row 7 of the EBR-II sodium-cooled fast breeder reactor. Three of the subassemblies are being irradiated at temperatures near 400 0 C, and the fourth subassembly is being irradiated at a temperature of 550 0 C. Creep and swelling strains were determined by profilometer measurements on the full length of the capsules after each irradiation cycle. The accumulated neutron dose on the 304L capsules at 385 0 C was 45 dpa; on the 316 capsules at 400 0 C, 40 dpa; and on the 316 capsules at 550 0 C, 25 dpa. It was found that the in-reactor creep rates were linearly dependent on hoop stress, with the exception being capsules of 316 stainless steel that had been given long-term carbide aging treatment and then irradiated at 550 0 C. Those capsules exhibited much higher creep and swelling rates than their unaged counterparts. For the metallurgical conditions where significant swelling was observed (solution-annealed 304L and aged 316 stainless steels), it was found that the in-reactor creep rates were readily fit to a model that related the creep rates to accumulated swelling. Additionally, it was found that the stress-normalized creep rate for 20%-cold-worked 316 stainless steel at a temperature of 550 0 C was 1.6 times that observed at 400 0 C

Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

2000-03-01

Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Irradiation and examination results of the AC-3 mixed-carbide test

International Nuclear Information System (INIS)

Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

1992-01-01

The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

International Nuclear Information System (INIS)

Kalinowski, J.E.

1979-01-01

Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

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

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki, E-mail: koyanagit@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kumar, N.A.P. Kiran [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hwang, Taehyun [Tohoku University, Sendai, 980-8579 (Japan); Garrison, Lauren M.; Hu, Xunxiang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2017-07-15

Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

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

Floquet Weyl semimetals in light-irradiated type-II and hybrid line-node semimetals

Science.gov (United States)

Chen, Rui; Zhou, Bin; Xu, Dong-Hui

2018-04-01

Type-II Weyl semimetals have recently attracted intensive research interest because they host Lorentz-violating Weyl fermions as quasiparticles. The discovery of type-II Weyl semimetals evokes the study of type-II line-node semimetals (LNSMs) whose linear dispersion is strongly tilted near the nodal ring. We present here a study on the circularly polarized light-induced Floquet states in type-II LNSMs, as well as those in hybrid LNSMs that have a partially overtilted linear dispersion in the vicinity of the nodal ring. We illustrate that two distinct types of Floquet Weyl semimetal (WSM) states can be induced in periodically driven type-II and hybrid LNSMs, and the type of Floquet WSMs can be tuned by the direction and intensity of the incident light. We construct phase diagrams of light-irradiated type-II and hybrid LNSMs which are quite distinct from those of light-irradiated type-I LNSMs. Moreover, we show that photoinduced Floquet type-I and type-II WSMs can be characterized by the emergence of different anomalous Hall conductivities.

Preparation and properties of functional mixed-lipid liposomes by γ-ray irradiation

International Nuclear Information System (INIS)

Hosoi, Fumio; Omichi, Hideki; Akama, Kazuhiro; Awai, Kouji; Yano, Yoshihiro; Nakano, Yoshio

1998-01-01

The feature of mixed-lipid liposomes such as polymerization and polymerized liposomes stability were investigated to find means for producing red cells containing hemoglobin inside the liposomes. The surface pressure-area isotherm values of the mixed-lipid monolayer indicated 1-stearoyl-2-(2,4-octadecadienoyl)-glycero-3-phosphocholine (SOPC) to be immiscible in cholesterol (Chol) and stearic acid (SA), and each component to contain separate domains in the bilayer membrane of liposomes. Radiation induced polymerization of mixed-SOPC liposomes was carried out using γ-rays from 60 Co at 4degC to stabilize lipid bilayers. The polymer yield increased significantly by adding Chol and SA to SOPC. The rate of polymerization of SOPC liposomes increased linearly with increasing of dose rate. The molecular weight of the polymer decreased with an increase in irradiation time. Irradiated SOPC/Chol/SA liposome vesicle size was affected by freeze-thawing. The vesicle size did not change when SOPC/Chol/SA was present in the system due to the addition of immiscible saturated 1,2-dipalmitoyl-glycero-3-phosphocholine (DPPC). (author)

Performance degradation of ferrofluidic feedthroughs in a mixed irradiation field

Science.gov (United States)

Simos, Nikolaos; Fernandes, S.; Mittig, Wolfgang; Pellemoine, Frederique; Avilov, M.; Kostin, M.; Mausner, L.; Ronningen, R.; Schein, M.; Bollen, G.

2017-01-01

Ferrofluidic feedthrough (FF) rotary seals containing either NdFeB or SmCo-type permanent magnets have been considered for use in the target and beam dump systems of the Facility for Rare Isotope Beams (FRIB). To evaluate their performance under irradiation three FF seals were irradiated in a mixed field consisting of fast neutrons, protons and γ-rays to an average absorbed dose of 0.2, 2.0, and 20.0 MGy at the Brookhaven Linac Isotope Producer facility (BLIP). The radiation types and energy profiles mimic those expected at the FRIB facility. Degradation of the operational performance of these devices due to irradiation is expected to be the result of the de-magnetization of the permanent magnets contained within the seal and the changes in the ferrofluid properties. Post-irradiation performance was evaluated by determining the ferrofluidic seal vacuum tightness and torque under static and dynamic conditions. The study revealed that the ferrofluidic feedthrough seal irradiated to a dose of 0.2 MGy maintained its vacuum tightness under both static and rotational condition while the one irradiated to a dose of 2.0 MGy exhibited signs of ferrofluid damage but no overall performance loss. At 20 MGy dose the effects of irradiation on the ferrofluid properties (viscosity and particle agglomeration) were shown to be severe. Furthermore, limited de-magnetization of the annular shaped Nd2Fe14B and Sm2Co17 magnets located within the irradiated FFs was observed for doses of 0.2 MGy and 20 MGy respectively.

Effect of dietary poly unsaturated fatty acids on total brain lipid concentration and anxiety levels of electron beam irradiated mice

International Nuclear Information System (INIS)

Suchetha Kumari; Bekal, Mahesh

2013-01-01

The whole brain irradiation causes injury to the nervous system at various levels. Omega-3 poly unsaturated fatty acids are very much essential for the growth and development of nervous system. Dietary supplementation of these nutrients will promote the development of injured neuronal cells. Therefore this study was undertaken to establish the role of Omega-3 poly unsaturated fatty acids on total brain lipid concentration, lipid peroxidation and anxiety levels in the irradiated mice. The effect of Electron Beam Radiation (EBR) on total brain lipid concentration, lipid peroxidation and anxiety level were investigated in male Swiss albino mice. The study groups were subjected to a sub-lethal dose of EBR and also the flax seed extract and fish oil were given orally to the irradiated mice. Irradiated groups show significant elevation in anxiety levels when compared to control group, indicating the acute radiation effects on the central nervous system. But the oral supplementation of dietary PUFA source decrees the anxiety level in the irradiated group. The analysis of lipid peroxidation showed a significant level of changes when compared between control and radiation groups. Dietary PUFA supplementation showed a significant level of decrease in the lipid peroxidation in the irradiated groups. The observation of total lipids in brain shows decrease in concentration in the irradiated groups, the differences in the variables follow the similar patterns as of that the MDA levels. This study suggests that the dietary intake of PUFAs may help in prevention and recovery of the oxidative stress caused by radiation. (author)

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)

Grain boundary sinks in neutron-irradiated Zr and Zr-alloys

International Nuclear Information System (INIS)

Griffiths, M.; Gilbert, R.W.; Coleman, C.E.

1988-01-01

Samples of annealed sponge and crystal-bar Zr and Zircaloy-2 have been examined following irradiation in EBR-II at temperatures ≅ 700 K. Loop analysis shows that there is selective denuding of interstitial loops near to some grain boundaries indicating that such boundaries are net sinks for interstitial point defects. Furthermore, in sponge Zr and Zircaloy-2, vacancy c-component loops are observed running into the grain boundaries showing that the grain boundaries are not preferred sinks for vacancies. Cavities are observed in all samples. In crystal-bar Zr and sponge Zr they are mostly observed adjacent to grain boundaries. They are also sometimes found within grains associated with precipitates. The cavities are more common in the crystal-bar Zr and this is probably because both the sponge Zr and Zircaloy-2 contain vacancy c-component loops which compete for vacancies (assuming that the cavities are vacancy sinks). Only some of the grain boundaries have cavities adjacent to them and this may be related to the orientation of the boundary. (orig.)

Growth of micropropagated lowbush blueberry with defined fungi in irradiated peat mix

International Nuclear Information System (INIS)

Litten, Walter; Smagula, J.M.; Dalpe, Yolande

1992-01-01

There is an interest in vegetative multiplication of high-yielding clones of Vaccinium angustifolium Ait. to establish or enhance blueberry production. This study evaluates mycorrhizal inoculation as an aid in such propagation from microcuttings. Shoots of Vaccinium angustifolium (clone 7062) generated in vitro were rooted in a peat-vermiculite-perlite substrate with or without ericoid mycorrhizal fungi fortification by Hymenoscyphus ericae or Scytalidium vaccinii and with or without peat sterilization by γ irradiation. Both in irradiated peat mix inoculated with S. vaccinii and in unirradiated peat mix with H. ericae, microcuttings grew taller and branched more than with the four other treatments. The profusely rooted plantlets available from all treatments of the cuttings put on significantly more total length of stems and branches after 167 days in the greenhouse when growing with either inoculant in unirradiated peat than in the unirradiated peat without inoculation. However, the magnitude of difference might be of borderline importance in commercial nursery operations. A higher level of copper and zinc in stem tissue was observed in stem tissue of plants grown with H. ericae with or without irradiation but not with S. vaccinii

Hot fuel examination facility element spacer wire-wrap machine

International Nuclear Information System (INIS)

Tobias, D.A.; Sherman, E.K.

1989-01-01

Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

Current Status of the LIFE Fast Reactors Fuel Performance Codes

International Nuclear Information System (INIS)

Yacout, A.M.; Billone, M.C.

2013-01-01

The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

Therapy of infections in mice irradiated in mixed neutron/photon fields and inflicted with wound trauma: a review of current work.

Science.gov (United States)

Ledney, G D; Madonna, G S; Elliott, T B; Moore, M M; Jackson, W E

1991-10-01

When host antimicrobial defenses are severely compromised by radiation or trauma in conjunction with radiation, death from sepsis results. To evaluate therapies for sepsis in radiation casualties, we developed models of acquired and induced bacterial infections in irradiated and irradiated-wounded mice. Animals were exposed to either a mixed radiation field of equal proportions of neutrons and gamma rays (n/gamma = 1) from a TRIGA reactor or pure gamma rays from 60[Co sources. Skin wounds (15% of total body surface area) were inflicted under methoxyflurane anesthesia 1 h after irradiation. In all mice, wounding after irradiation decreased resistance to infection. Treatments with the immunomodulator synthetic trehalose dicorynomycolate (S-TDCM) before or after mixed neutron-gamma irradiation or gamma irradiation increased survival. Therapy with S-TDCM for mice irradiated with either a mixed field or gamma rays increased resistance to Klebsiella pneumoniae-induced infections. Combined therapy with S-TDCM and ceftriaxone for K. pneumoniae infections in mice exposed to a mixed radiation field or to gamma rays was more effective than single-agent therapy. In all irradiated-wounded mice, single therapy of acquired infections with an antibiotic or S-TDCM did not increase survival. Survival of irradiated-wounded mice after topical application of gentamicin sulfate cream suggested that bacteria colonizing the wound disseminated systemically in untreated irradiated mice, resulting in death from sepsis. In lethal models of acquired infections in irradiated-wounded mice, significant increases in survival were achieved when systemic treatments with S-TDCM or gentamicin were combined with topical treatments of gentamicin cream. Therapies for sepsis in all mice exposed to a mixed field were less effective than in mice exposed to gamma rays. Nonetheless, the data show a principle by which successful therapy may be provided to individuals receiving tissue trauma in conjunction with

The Columbia University microbeam II endstation for cell imaging and irradiation

International Nuclear Information System (INIS)

Bigelow, A.W.; Ross, G.J.; Randers-Pehrson, G.; Brenner, D.J.

2005-01-01

The Columbia University Microbeam II has been built to provide a focused ion beam for irradiating designated mammalian cells with single particles. With the interest in irradiating non-stained cells and cells in three-dimensional tissue samples, the endstation was designed to accommodate a variety of imaging techniques, in addition to fluorescent microscopy. Non-stained cells are imaged either by quantitative phase microscopy (QPm) [IATIA, Box Hill North, Victoria, 3129, Australia [1

Large and almost maximal neutrino mixing within the type II see-saw mechanism

International Nuclear Information System (INIS)

Lindner, Manfred; Rodejohann, Werner

2007-01-01

Within the type II see-saw mechanism the light neutrino mass matrix is given by a sum of a direct (or triplet) mass term and the conventional (type I) see-saw term. Both versions of the see-saw mechanism explain naturally small neutrino masses, but the type II scenario offers interesting additional possibilities to explain large or almost maximal or vanishing mixings which are discussed in this paper. We first introduce 'type II enhancement' of neutrino mixing, where moderate cancellations between the two terms can lead to large neutrino mixing even if all individual mass matrices and terms generate small mixing. However, nearly maximal or vanishing mixings are not naturally explained in this way, unless there is a certain initial structure (symmetry) which enforces certain elements of the matrices to be identical or related in a special way. We therefore assume that the leading structure of the neutrino mass matrix is the triplet term and corresponds to zero U e3 and maximal θ 23 . Small but necessary corrections are generated by the conventional see-saw term. Then we assume that one of the two terms corresponds to an extreme mixing scenario, such as bimaximal or tri-bimaximal mixing. Deviations from this scheme are introduced by the second term. One can mimic Quark-Lepton Complementarity in this way. Finally, we note that the neutrino mass matrix for tri-bimaximal mixing can be-depending on the mass hierarchy-written as a sum of two terms with simple structure. Their origin could be the two terms of type II see-saw

Experimental studies on the effect of perfluorochemicals in tumor irradiation

International Nuclear Information System (INIS)

Shinoda, Jun; Iwai, Tomohiko; Hattori, Tatsuaki; Kondo, Hiroaki; Sakai, Noboru; Yamada, Hiroshi

1984-01-01

The effects of radiation therapy with Fluosol-DA on rat mammary tumors were studied. The tissue oxygen tension values of tumors in breathing mixed gas (5% carbon dioxide and 95% oxygen) with Fluosol-DA (25 ml/kg, i.v.) were significantly higher than those in room air without Fluosol-DA. The rats were divided into three groups: Group I received Fluosol-DA but no irradiation, Group II was treated with 1000 rads of irradiation using 60 Co without Fluosol-DA in room air and Group III received the same irradiation and Fluosol-DA in breathig mixed gas. In the latter group we observed a prolongation of the survival time and suppression of the tumor growth. (author)

Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

International Nuclear Information System (INIS)

Chung, H.M.; Strain, R.V.; Shack, W.J.

2001-01-01

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20-100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ∼50 dpa at ∼370 deg. C. Slow-strain-rate tensile tests were conducted at 289 degree sign C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at a low ECP, and this susceptibility led to a poor work-hardening capability and low ductility

Research of CITP-II tritium production irradiation device design

International Nuclear Information System (INIS)

Zhang Zhihua; Deng Yongjun; Mi Xiangmiao; Li Rundong; Liu Zhiyong

2012-01-01

As the core component of CITP-II, the online tritium production irradiation device is the pivotal equipment in the research on tritium production and release of tritium breeders. The design of CITP-II online tritium production irradiation device creatively makes replacing the breeders online come true; as tritium production capacity, the self-shielding factor of device, and neutron flux were studied. The influence of different load models and load thicknesses of breeders to tritium production capacity was calculated. The hydrodynamics parameters of device in solid-gas phase were computed. Thermal parameters, such as the heat power of breeders, hotspot, temperature grads distributions, utmost temperature, uneven factors, were analyzed. Creatively designed nonlinear electric heater equalized breeders' even heat power. The influence laws of the components, pressure of gap gas and carrier gas to the balance temperature were got. And the key thermal parameters were ascertained. The key thermal parameters and the changing laws were got and provide the basis for structural optimization and safety analysis. They can also be referenced for the study of breeders' tritium production and release. (authors)

Hematologic status of mice submitted to sublethal total body irradiation with mixed neutron-gamma radiation

International Nuclear Information System (INIS)

Herodin, F.; Court, L.

1989-01-01

The hematologic status of mice exposed to sublethal whole body irradiation with mixed neutron-gamma radiation (mainly neutrons) is studied. A slight decrease of the blood cell count is still observed below 1 Gy. The recovery of bone marrow granulocyte-macrophage progenitors seems to require more time than after pure gamma irradiation [fr

A retrospective study of Class II mixed-dentition treatment.

Science.gov (United States)

Oh, Heesoo; Baumrind, Sheldon; Korn, Edward L; Dugoni, Steven; Boero, Roger; Aubert, Maryse; Boyd, Robert

2017-01-01

To consider the effectiveness of early treatment using one mixed-dentition approach to the correction of moderate and severe Class II malocclusions. Three groups of Class II subjects were included in this retrospective study: an early treatment (EarlyTx) group that first presented at age 7 to 9.5 years (n = 54), a late treatment (LateTx) group whose first orthodontic visit occurred between ages 12 and 15 (n = 58), and an untreated Class II (UnTx) group to assess the pretreatment comparability of the two treated groups (n = 51). Thirteen conventional cephalometric measurements were reported for each group and Class II molar severity was measured on the study casts of the EarlyTx and LateTx groups. Successful Class II correction was observed in approximately three quarters of both the EarlyTx group and the LateTx group at the end of treatment. EarlyTx patients had fewer permanent teeth extracted than did the LateTx patients (5.6% vs 37.9%, P < .001) and spent less time in full-bonded appliance therapy in the permanent dentition than did LateTx patients (1.7 ± 0.8 vs 2.6 ± 0.7years, P < .001). When supervision time is included, the EarlyTx group had longer total treatment time and averaged more visits than did the LateTx group (53.1 ± 18. 8 vs 33.7 ± 8.3, P < .0001). Fifty-five percent of the LateTx extraction cases involved removal of the maxillary first premolars only and were finished in a Class II molar relationship. EarlyTx comprehensive mixed-dentition treatment was an effective modality for early correction of Class II malocclusions.

Detection of irradiated spice in blend of irradiated and un-irradiated spices using thermoluminescence method

International Nuclear Information System (INIS)

Goto, Michiko; Yamazaki, Masao; Sekiguchi, Masayuki; Todoriki, Setsuko; Miyahara, Makoto

2007-01-01

Five blended spice sample were prepared by mixing irradiated and un-irradiated black pepper and paprika at different ratios. Blended black pepper containing 2%(w/w) of 5.4 kGy-irradiated black pepper showed no maximum at glow1. Irradiated black pepper samples, mixed to 5 or 10%(w/w), were identified as 'irradiated' or 'partially irradiated' or 'un-irradiated'. All samples with un-irradiated pepper up to 20%(w/w) were identified as irradiated'. In the case 5.0 kGy-irradiated paprika were mixed with un-irradiated paprika up to 5%(w/w), all samples were identified as irradiated'. The glow1 curves of samples, including irradiated paprika at 0.2%(w/w) or higher, exhibited a maximum between 150 and 250degC. The results suggest the existence of different critical mixing ratio for the detection of irradiation among each spices. Temperature range for integration of the TL glow intensity were compared between 70-400degC and approximate 150-250degC, and revealed that the latter temperature range was determined based on the measurement of TLD100. Although TL glow ratio in 150-250degC was lower than that of 70-400degC range, identification of irradiation was not affected. Treatment of un-irradiated black pepper and paprika with ultraviolet rays had no effect on the detection of irradiation. (author)

Tritium and helium retention and release from irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; Longhurst, G.R.; Oates, M.A.; Pawelko, R.J. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

1998-01-01

This paper reports the results of an experimental effort to anneal irradiated beryllium specimens and characterize them for steam-chemical reactivity experiments. Fully-dense, consolidated powder metallurgy Be cylinders, irradiated in the EBR-II to a fast neutron (>0.1 MeV) fluence of {approx}6 x 10{sup 22} n/cm{sup 2}, were annealed at temperatures from 450degC to 1200degC. The releases of tritium and helium were measured during the heat-up phase and during the high-temperature anneals. These experiments revealed that, at 600degC and below, there was insignificant gas release. Tritium release at 700degC exhibited a delayed increase in the release rate, while the specimen was at 700degC. For anneal temperatures of 800degC and higher, tritium and helium release was concurrent and the release behavior was characterized by gas-burst peaks. Essentially all of the tritium and helium was released at temperatures of 1000degC and higher, whereas about 1/10 of the tritium was released during the anneals at 700degC and 800degC. Measurements were made to determine the bulk density, porosity and specific surface area for each specimen before and after annealing. These measurements indicated that annealing caused the irradiated Be to swell, by as much as 14% at 700degC and 56% at 1200degC. Kr gas adsorption measurements for samples annealed at 1000degC and 1200degC determined specific surface areas between 0.04 m{sup 2}/g and 0.1 m{sup 2}/g for these annealed specimens. The tritium and helium gas release measurements and the specific surface area measurements indicated that annealing of irradiated Be caused a porosity network to evolve and become surface-connected to relieve internal gas pressure. (author)

Dissociation behavior of methane--ethane mixed gas hydrate coexisting structures I and II.

Science.gov (United States)

Kida, Masato; Jin, Yusuke; Takahashi, Nobuo; Nagao, Jiro; Narita, Hideo

2010-09-09

Dissociation behavior of methane-ethane mixed gas hydrate coexisting structures I and II at constant temperatures less than 223 K was studied with use of powder X-ray diffraction and solid-state (13)C NMR techniques. The diffraction patterns at temperatures less than 203 K showed both structures I and II simultaneously convert to Ih during the dissociation, but the diffraction pattern at temperatures greater than 208 K showed different dissociation behavior between structures I and II. Although the diffraction peaks from structure II decreased during measurement at constant temperatures greater than 208 K, those from structure I increased at the initial step of dissociation and then disappeared. This anomalous behavior of the methane-ethane mixed gas hydrate coexisting structures I and II was examined by using the (13)C NMR technique. The (13)C NMR spectra revealed that the anomalous behavior results from the formation of ethane-rich structure I. The structure I hydrate formation was associated with the dissociation rate of the initial methane-ethane mixed gas hydrate.

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

Ionophoretic method in the study of mixed ligand ternary chelates of UO2(II), Ni(II) and Zn(II) involving nitrilotriacetate and cytosine as ligands

International Nuclear Information System (INIS)

Mishra, A.P.; Mishra, S.K.; Yadava, K.L.

1987-01-01

A novel electrophoretic technique is described for the assessment of the equilibria in mixed-ligand complex system in solution. It is based on the movement of spot of the metal ion under an electric field with the complexants added in the background electrolyte at fixed pH. The concentration of primary ligand nitrilotriacetate was constant while that of secondary ligand (cytosine) was varied. The plot of log (cytosine) against mobility was used to obtain information on the formation of the mixed complexes and to calculate its stability constants. Experimentally obtained logK values are as 5.62, 4.55 and 4.42 for mixed complexes of UO 2 (II), Ni(II) and Zn(II) respectively at μ=0.1 and temp.=35 +- 01.degC. (author). 10 refs

Fragrance mix II in the baseline series contributes significantly to detection of fragrance allergy

DEFF Research Database (Denmark)

Heisterberg, Maria V; Andersen, Klaus E; Avnstorp, Christian

2010-01-01

Fragrance mix II (FM II) is a relatively new screening marker for fragrance contact allergy. It was introduced in the patch test baseline series in Denmark in 2005 and contains six different fragrance chemicals commonly present in cosmetic products and which are known allergens.......Fragrance mix II (FM II) is a relatively new screening marker for fragrance contact allergy. It was introduced in the patch test baseline series in Denmark in 2005 and contains six different fragrance chemicals commonly present in cosmetic products and which are known allergens....

Microstructural examination of 12% Cr martensitic stainless steel after irradiation at elevated temperatures in FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Hsu, Chen-Yih; Gelles, D.S.; Lechtenberg, T.A.

1986-06-01

A remelted 12% Cr martensitic stainless steel (HT-9) has been examined by transmission electron microscopy before and after irradiation in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The irradiation temperatures were 365,420, 520, and 600 degree C with the fluences as high as 7.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 34 dpa. The extracted precipitates from each specimen were identified using x-ray microanalysis and selected area diffraction. The precipitates in the unirradiated condition were primarily M 23 C 6 carbides, which formed at martensite lath and prior austenite grain boundaries. During irradiation at elevated temperatures, small amounts of other phases formed, which were tentatively identified as the chromium-rich α', the nickel-silicon rich G-phase, and the intermetallic Chi phase. Irradiation-induced voids were observed only in specimens irradiated at 420 degree C to a dose of 34 dpa; no voids were found for specimens irradiated at 365, 520, and 600 degree C (∼11, ∼34, and ∼34 dpa). These results are not in agreement with previous experiments in that voids have not been reported in this alloy at relatively high fluence level (∼67 dpa) following irradiation in another fast-spectrum reactor (EBR.II). This is, however, the first observation following FFTF irradiation. The present results indicate that cavities can form in HT-9 at modest fluence levels even without significant generation of helium. Hence, the cavity formation in this class of ferritic alloys is not simply caused by helium generation but rather more complex mechanisms. 12 refs., 2 figs., 3 tabs

Long term low dose rate irradiation causes recovery from type II diabetes and suppression of aging in type II diabetes-prone mice

International Nuclear Information System (INIS)

Namura, T.; Oda, T.

2003-01-01

The effects of low dose rate gamma irradiation on model C57BL/KsJ-db/db mice with Type II diabetes mellitus was investigated. These mice develop Type II diabetes by 10 weeks of age, due to obesity, and are characterized by hyperinsulinemia. A group of 12 female 10-week old mice were irradiated at 0.65 mGy/hr in the low dose rate irradiation facility in the Low Dose Radiation Research Center. The urine glucose levels of all of the mice were strongly positive at the beginning of the irradiation. In the irradiated group, a decrease in the glucose level was observed in three mice, one in the 35th week, another in the 52nd week and the third in the 80th week. No recovery from the diabetes was observed in the 12 mice of non-irradiated control group. There was no systematic change of body weight or consumption of food and drinking water between the irradiated group and the non-irradiated group or between the recovered mice and the non-recovered mice. Survival was better in the irradiated group. The surviving fraction at the age of 90 weeks was 75 % in the irradiated group but only 40 % in the non-irradiated. A marked difference was also observed in the appearance of the coat hair, skin and tail. The irradiated group was in much better condition. Mortality was delayed and the healthy appearance was prolonged in the irradiated mice by about 20-30 weeks compared with the control mice. These results suggest that the low dose irradiation modified the condition of the diabetic mice, leading not only to recovery from diabetes, but also to suppression of the aging process

Effects of low-dose rate irradiation on two types of type II diabetes model mice

International Nuclear Information System (INIS)

Nomura, Takaji; Sakai, Kazuo

2004-01-01

The effects of low-dose rate gamma-irradiation were investigated in two mouse strains - C57BL/KsJ-db/db (db mouse) and AKITA (AKITA mouse)-for type II diabetes mellitus. Both strains develop the developed type II diabetes by about 8 weeks of age due to dysfunction of the insulin/insulin receptor. The db Mouse' shows obese and exhibits hyperinsulinism, and the onset of Type II diabetes like resembles that for Westerners. On the other hand, the AKITA mouse has exhibits disordered insulin secretion, and the diabetes such as resembles that of Asians. Ten-week old female mice, in groups of 8 or 12, were irradiated at 0.65 mGy/hr in the low-dose rate irradiation facility in the Low Dose Radiation Research Center. The level of urine glucose was measured with test slips. The urine glucose levels of all of the mice were highly elevated the beginning of the irradiation. In the irradiated group of db mice, three mice showed decrease in glucose level compare to the level of non-irradiated diabetes mice after 35, 52 or 80 weeks of irradiation. All had maintained a normal level thereafter. No such improvement in diabetes was ever observed in the 12 mice of in the non-irradiated control group. The AKITA mice, however, did not decrease the glucose level regardless of the irradiation. Both the db mice and AKITA mice had their lives prolonged their life by the irradiation. The survival rate of db mice at the age of 90 weeks was 75% in the irradiated group, but 50% in the non-irradiated group. The average life span was 104 weeks in the irradiated group and 87 weeks in the control group. Furthermore, a marked difference was furthermore observed in the appearance of the coat hair, skin, and tail; appearances were well preserved in the irradiated group. The average life span in the irradiated AKITA mice was also longer than that for the non-irradiated mice, 51 weeks and 41 weeks in the irradiated and non-irradiated group respectively. These results suggest that the low-dose irradiation

Development of a swelling equation for 20%-CW 316 in a fusion device

International Nuclear Information System (INIS)

1980-01-01

The difficulties involved in the development of swelling correlations for AISI 316 in fusion environments are discussed. A set of void and bubble-swelling correlations has been developed which incorporates the limited available data from EBR-II and HFIR irradiations. It appears that at high fluences helium may play a minor role in the determination of total swelling over a considerable temperature range

Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

International Nuclear Information System (INIS)

Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

1992-01-01

The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

Extension of the RPV irradiation surveillance program of NPP GKN II by T0 approach

International Nuclear Information System (INIS)

Barthelmes, J.; Keim, E.; Hein, H.; Koenig, G.

2015-01-01

The nuclear power plant (NPP) Neckarwestheim II (GKN II) started operation in 1989 and was designed for 40 years of operation. During the plant life time the reactor pressure vessel (RPV) integrity is a main aspect for nuclear safety since the RPV is exposed to neutron irradiation affecting the mechanical material properties, in particular toughness. In this context the ductile to brittle transition reference temperature of the RPV materials can be determined either indirectly according to the RT(NDT) concept by means of comparative examinations of irradiated and unirradiated notched-bar impact specimens or directly according to the Master Curve concept by means of examination of irradiated fracture mechanic specimens and determination of an alternative reference temperature RT(T0). With the implementation and evaluation of the first irradiation surveillance program consisting of three sets, one unirradiated reference set (set 1) and two irradiated sets (set 2 and 3), the RPV safety could be proven for the assessment fluence (AF) of 8*10 18 cm -2 (E > 1 MeV) using the RT(NDT) concept. Against the background of a possible long term operation and the state-of-the-art of science and technology in 1998 the NPP GKN II initiated a supplemental irradiation surveillance program with two irradiation sets (set 4 and 5) containing fracture mechanic specimens for complementary proof of safety according to the Master Curve concept. The results of the first irradiated set 4 are presented and assessed by means of the reference temperatures according to the Master Curve concept and compared to the results of the irradiation sets 1 to 3 of the conventional irradiation surveillance program. As an important outcome the existing RPV integrity assessment could be ensured by the Master Curve results. The applied approach adapts to the state-of-the-art of science and technology and is best practice to ensure the safe operation of RPV supplementary. (authors)

The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

International Nuclear Information System (INIS)

Freund, D.; Geithoff, D.

1988-04-01

The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Uncommon Mixed Type I and II Choledochal Cyst: An Indonesian Experience

Directory of Open Access Journals (Sweden)

Fransisca J. Siahaya

2013-01-01

Full Text Available Bile duct cyst is an uncommon disease worldwide; however, its incidence is remarkably high in Asian population, primarily in children. Nevertheless, the mixed type choledochal cysts are extremely rare especially in adults. A case report of a 20-year-old female with a history of upper abdominal pain that was diagnosed with cholecystitis with stone and who underwent laparoscopic cholecystectomy is discussed. Choledochal malformation was found intraoperatively. Magnetic resonance cholangiography (MRCP and USG after first surgery revealed extrahepatic fusiform dilatation of the CBD; therefore, provisional diagnosis of type I choledochal cyst was made. Complete resection of the cyst was performed, and a mixed type I and II choledochal cyst was found intraoperatively. Bile duct reconstruction was carried out with Roux-en-Y hepaticojejunostomy. The mixed type I and II choledochal cysts are rare in adults, and this is the third adult case that has been reported. The mixed type can be missed on radiology imaging, and diagnosing the anomaly is only possible after a combination of imaging and intraoperative findings. Mixed type choledochal cyst classification should not be added to the existing classification since it does not affect the current operative techniques.

Precipitate evolution in low-nickel austenitic stainless steels during neutron irradiation at very low dose rates

International Nuclear Information System (INIS)

Isobe, Y.; Sagisaka, M.; Garner, F.; Okita, T.

2007-01-01

Full text of publication follows: Not all components of a fusion reactor will be subjected to high atomic displacement rates. Some components outside the plasma containment may experience relatively low displacement rates but data generated under long-term irradiation at low dpa rates is hard to obtain. In another study the neutron-induced microstructural evolution in response to long term irradiation at very low dose rates was studied for a Russian low-nickel austenitic stainless steel that is analogous to AISI 304. The irradiated samples were obtained from an out-of-core anti-crush support column for the BN-600 fast reactor with doses ranging from 1.5 to 22 dpa generated at 3x10 -9 to 4x10 -8 dpa/s. The irradiation temperatures were in a very narrow range of 370-375 deg. C. Microstructural observation showed that in addition to voids and dislocations, an unexpectedly high density of small carbide precipitates was formed that are not usually observed at higher dpa rates in this temperature range. These results required us to ask if such unexpected precipitation was anomalous or was a general feature of low-flux, long-term irradiation. It is shown in this paper that a similar behavior was observed in a western stainless steel, namely AISI 304 stainless steel, irradiated at similar temperatures and dpa rates in the EBR-II fast reactor, indicating that irradiation at low dpa rates for many years leads to a different precipitate microstructure and therefore different associated changes in matrix composition than are generated at higher dpa rates. One consequence of this precipitation is a reduced lattice parameter of the alloy matrix, leading to densification that increases in strength with increasing temperature and dose. A. non-destructive method to evaluate these precipitates is under development and is also discussed in this paper. (authors)

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

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

United States Department of Energy breeder reactor staff training domestic program

International Nuclear Information System (INIS)

1984-01-01

Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described

Polystyrene with pendant mixed functional ruthenium(II)-terpyridine complexes

NARCIS (Netherlands)

Heller, M.; Schubert, U.S.

2002-01-01

A vinyl substituted 2,2:6,2-terpyridine and a mixed, bifunctional ruthenium(II)-terpyridine complex bearing a vinyl and a hydroxymethyl group are utilized as comonomers for radical copolymerization with styrene. The resulting polymers are characterized by means of UV-vis spectroscopy and gel

Mixed-Ligand Complexes Of Nickel (II) With 2-Acetylpyridine ...

African Journals Online (AJOL)

The preparation and spectral properties of five nickel (II) mixed-ligands complexes (Ni [2-Actsc.Y]CI2), derived from 2-acetylpyridinethiosermicarbazones and some nitrogen/sulphur monodentate ligands such as thiophene, ammonia, picoline, pyridine and aniline are described. The complexes have been characterized on ...

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

Synergistic effect of mixed neutron and gamma irradiation in bipolar operational amplifier OP07

Energy Technology Data Exchange (ETDEWEB)

Yan, Liu, E-mail: liuyan@nint.ac.cn [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O.Box 69-10, Xi’an 710024 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wei, Chen; Shanchao, Yang; Xiaoming, Jin [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O.Box 69-10, Xi’an 710024 (China); Chaohui, He [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China)

2016-09-21

This paper presents the synergistic effects in bipolar operational amplifier OP07. The radiation effects are studied by neutron beam, gamma ray, and mixed neutron/gamma ray environments. The characterateristics of the synergistic effects are studied through comparison of different experiment results. The results show that the bipolar operational amplifier OP07 exhibited significant synergistic effects in the mixed neutron and gamma irradiation. The bipolar transistor is identified as the most radiation sensitive unit of the operational amplifier. In this paper, a series of simulations are performed on bipolar transistors in different radiation environments. In the theoretical simulation, the geometric model and calculations based on the Medici toolkit are built to study the radiation effects in bipolar components. The effect of mixed neutron and gamma irradiation is simulated based on the understanding of the underlying mechanisms of radiation effects in bipolar transistors. The simulated results agree well with the experimental data. The results of the experiments and simulation indicate that the radiation effects in the bipolar devices subjected to mixed neutron and gamma environments is not a simple combination of total ionizing dose (TID) effects and displacement damage. The data suggests that the TID effect could enhance the displacement damage. The synergistic effect should not be neglected in complex radiation environments.

Mixed ligand complexes of alkaline earth metals: Part XII. Mg(II, Ca(II, Sr(II and Ba(II complexes with 5-chlorosalicylaldehyde and salicylaldehyde or hydroxyaromatic ketones

Directory of Open Access Journals (Sweden)

MITHLESH AGRAWAL

2002-04-01

Full Text Available The reactions of alkaline earth metal chlorides with 5-chlorosalicylaldehyde and salicylaldehyde, 2-hydroxyacetophenone or 2-hydroxypropiophenone have been carried out in 1 : 1 : 1 mole ratio and the mixed ligand complexes of the type MLL’(H2O2 (where M = Mg(II, Ca(II, Sr(II and Ba(II, HL = 5-chlorosalicylaldehyde and HL’ = salicylaldehyde, 2-hydroxyacetophenone or 2-hydroxypropiophenone have been isolated. These complexes were characterized by TLC, conductance measurements, IR and 1H-NMR spectra.

Hexacoordinated mixed-ligand complexes of vanadium(IV) and copper(II)

International Nuclear Information System (INIS)

Islam, M.S.; Motahera Begum; Roy, H.N.; Haroon, S.A.Q.M.

1996-01-01

The literature reports simple complexes of metal ions with Schiff bases derived from amino acids. But their mixed-ligand complexes are very rare. Keeping this fact in mind, some new mixed ligand complexes of V IV and Cu II with tridentate Schiff bases derived from glycine, salicylaldehyde and amino bases, e.g. quinoline (Q), isoquinoline (IQ), 2-picoline (2-pic), 4-picoline (4-pic) and pyridine (Py) were prepared and studied. 6 refs., 1 tab

Models of mixed irradiation with a 'reciprocal-time' pattern of the repair function

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Shozo; Miura, Yuri; Mizuno, Shoichi [Tokyo Metropolitan Inst. of Gerontology (Japan); Furusawa, Yoshiya [National Inst. of Radiological Sciences, Chiba (Japan)

2002-09-01

Suzuki presented models for mixed irradiation with two and multiple types of radiation by extending the Zaider and Rossi model, which is based on the theory of dual radiation action. In these models, the repair function was simply assumed to be semi-logarithmically linear (i.e., monoexponential), or a first-order process, which has been experimentally contradicted. Fowler, however, suggested that the repair of radiation damage might be largely a second-order process rather than a first-order one, and presented data in support of this hypothesis. In addition, a second-order repair function is preferred to an n-exponential repair function for the reason that only one parameter is used in the former instead of 2n-1 parameters for the latter, although both repair functions show a good fit to the experimental data. However, according to a second-order repair function, the repair rate depends on the dose, which is incompatible with the experimental data. We, therefore, revised the models for mixed irradiation by Zaider and Rossi and by Suzuki, by substituting a 'reciprocal-time' pattern of the repair function, which is derived from the assumption that the repair rate is independent of the dose in a second-order repair function, for a first-order one in reduction and interaction factors of the models, although the underlying mechanism for this assumption cannot be well-explained. The reduction factor, which reduces the contribution of the square of a dose to cell killing in the linear-quadratic model and its derivatives, and the interaction factor, which also reduces the contribution of the interaction of two or more doses of different types of radiation, were formulated by using a 'reciprocal-time' patterns of the repair function. Cell survivals calculated from the older and the newly modified models were compared in terms of the dose-rate by assuming various types of single and mixed irradiation. The result implies that the newly modified models for

Fragrance mix II in the baseline series contributes significantly to detection of fragrance allergy

DEFF Research Database (Denmark)

Heisterberg, Maria S Vølund; Andersen, Klaus E.; Avnstorp, Christian

2010-01-01

Background: Fragrance mix II (FM II) is a relatively new screening marker for fragrance contact allergy. It was introduced in the patch test baseline series in Denmark in 2005 and contains six different fragrance chemicals commonly present in cosmetic products and which are known allergens. Aim......: To investigate the diagnostic contribution of including FM II in the baseline series by comparing it with other screening markers of fragrance allergy: fragrance mix I (FM I), Myroxylon pereirae and hydroxyisohexyl 3-cyclohexene carboxaldehyde (HICC). Method: Retrospective study of 12 302 patients consecutively...

Study on the effect of food irradiation on some blood serum enzymes in rats

International Nuclear Information System (INIS)

Metwalli, O.M.

1977-01-01

Rats were fed irradiated diets as part of a study of screening tests for the safety of irradiated food. The diet consisted (g/100 g) of casein, 8.5; skim-milk, 9.4; potato starch, 50.00; wheat flour, 16.50; sucrose, 5.00; sunflower oil, 6.00; choline chloride, 0.10; salt mixture, 3.50; and vitamin mix.ure, 1.00. Diets were irradiated at 2.5 or 4.5 Mrad and were fed ad lib. for 4 months. Levels of serum (i) glutanic-pyruvic transaminase, (ii) glutamic-oxalacetic transaminase, and (iii) lactic dehydrogenase were determined. No significant changes were observed in (i) or (iii) on ieeding irradiated diets, or in (ii) for male rats. Significant decreases (P [de

Microstructure of irradiated Inconel 706 fuel pin cladding

International Nuclear Information System (INIS)

Yang, W.J.S.; Makenas, B.J.

1983-08-01

A fuel pin from the HEDL-P-60 experiment with a cladding of solution-annealed Inconel 706 breached in an apparently brittle manner at a position 12.7 cm above the bottom of the fuel column with a crack of 5.72 cm in length after 5.0 atomic percent burnup in EBR-II. Temperatures (time-averaged midwall) and fast fluences for the fractured area range from 447 0 C and 5.5 x 10 22 n/cm 2 to 526 0 C and 6.1 x 10 22 n/cm 2 (E > 0.1 MeV). Specimens of the fractured fuel pin section were successfully prepared and examined in both a scanning electron microscope and a transmission electron microscope. The fracture surfaces of the breached section showed brittle intergranular fracture characteristics for both the axial and circumferential cracks. Formation of γ' in the matrix near the breach confirmed that the irradiation temperature at the breached area was below 500 0 C, in agreement with other estimates of the temperature for the area, 447 to 526 0 C. A hexagonal eta-phase, Ni 3 (Ti,Nb), precipitated at boundaries near the breach. A more extensive eta-phase coating at grain boundaries was found in a section irradiated at 650 0 C. The eta-phase plates at grain boundaries are expected to have a detrimental effect on alloy ductility. A plane of weakness in this region along the (111) slip planes will develop in Inconel 706 because the eta-plates have a (111) habit relationship with the matrix

Neutron self-shielding with k0-NAA irradiations

International Nuclear Information System (INIS)

Chilian, C.; Chambon, R.; Kennedy, G.

2010-01-01

A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

Tensile properties in zircaloy-II after 590 MeV proton irradiation

Energy Technology Data Exchange (ETDEWEB)

Dai, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Victoria, M. [Ecole Polytechnique Federale, Lausanne (Switzerland)

1997-09-01

In order to investigate radiation potential damage effects on the SINQ Zircaloy-rod target, four Zircaloy-II tensile specimens were irradiated at the PIREX facility in 1995 to a proton fluence about 3x10{sup 20} p/cm{sup 2}, which produced a radiation damage of about 1.35 displacements per atom (dpa). Tensile test results show that, although there is some reduction in tensile elongation, substantial ductility still exists after such irradiation dose which corresponds to the peak value obtained in the SINQ target for 23 days operation at 1 mA. (author) 1 fig., 2 refs.

Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.

1968-01-01

Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

Chapter 2: Irradiators

Energy Technology Data Exchange (ETDEWEB)

NONE

2018-04-01

The chapter 2 presents the subjects: 1) gamma irradiators which includes: Category-I gamma irradiators (self-contained); Category-II gamma irradiators (panoramic and dry storage); Category-III gamma irradiators (self-contained in water); Category-IV gamma irradiators (panoramic and wet storage); source rack for Category-IV gamma irradiators; product transport system for Category-IV gamma irradiators; radiation shield for gamma irradiators; 2) accelerators which includes: Category-I Accelerators (shielded irradiator); Category-II Accelerators (irradiator inside a shielded room); Irradiation application examples.

Preparation and Spectral Properties of Mixed-Ligand Complexes of VO(IV, Ni(II, Zn(II, Pd(II, Cd(II and Pb(II with Dimethylglyoxime and N-acetylglycine

Directory of Open Access Journals (Sweden)

Shayma A. Shaker

2010-01-01

Full Text Available A number of mixed-ligand complexes of the general formula [M(D(G] where D=dimethylglyoximato monoanion, G=N-acetylglycinato and M=VO(IV, Ni(II, Zn(II, Pd(II, Cd(II and Pb(II were prepared. Each complex was characterized by elemental analysis, determination of metal, infrared spectra, electronic spectra, (1H and 13C NMR spectra, conductivity and magnetic moments. All these complexes were not soluble in some of the organic solvent but highly soluble in dimethylformamide. The conductivity data showed the non-electrolytic nature of the complexes. The electronic spectra exhibited absorption bands in the visible region caused by the d-d electronic transition such as VO(IV, Ni(II and Pd(II. The IR and (1H, 13C NMR spectra which have indicate that the dimethylglyoxime was coordinated with the metal ions through the N and O atoms of the oxime group and N-acetylglycine was coordinated with metal ions through the N atom and terminal carboxyl oxygen atom.

Mixed-ligand complexes of ruthenium(II) incorporating a diazo ...

Indian Academy of Sciences (India)

Unknown

Dedicated to the memory of the late Professor Bhaskar G Maiya. *For correspondence. Mixed-ligand complexes of ruthenium(II) incorporating a diazo ligand: Synthesis .... water (1 : 1) for 5 h to give a dark red solution. The solution was cooled to room temperature. After eva- poration of the solvent, the solid was collected,.

U-target irradiation at FRM II aiming the production of Mo-99 - A feasibility study

International Nuclear Information System (INIS)

Gerstenberg, H.; Mueller, C.; Neuhaus, I.; Roehrmoser, A.

2010-01-01

Following the shortage in radioisotope availability the Technische Unversitaet Muenchen and the Belgian Institut National des Radioelements conducted a common study on the suitability of the FRM II reactor for the generation of Mo-99 as a fission product. A suitable irradiation channel was determined and neutronic calculations resulted in sufficiently high neutron flux densities to make FRM II a promising candidate for Mo-99 production. In addition the feasibility study provides thermohydraulic calculations as input for the design and integration of the additional cooling circuit into the existing heat removal systems of FRM II. The required in-house processes for a regular uranium target irradiation programme have been defined and necessary upgrades identified. Finally the required investment cost was estimated and a possible time schedule was given. (author)

Shelf-life extension and improving micro-biological quality of mixed peas with diced carrot by gamma irradiation

International Nuclear Information System (INIS)

Taha, S. M.; Hammad, A. A.; Amal, S. M.; Gebreel, H. M.

2010-01-01

Mixed peas with diced carrot were collected and examined for their microbiological quality. All the examined samples had high level of microbial load. All examined samples contained Escherichia coli (E. coli) and Enterococcus faecalis (Ent. faecalis). Staphylococcus aureus (S. aureus) was detected in only 4 of samples (26.3%). The tested samples were free from Aeromonas hydrophila (A. hydrophila), Listeria monocytogenes (L monocytogenes) and Salmonella species. Gamma irradiation caused a great reduction in all microbial loads. During refrigerated storage, the counts of all microorganisms increased, but the rate of increase was slower as the irradiation dose increased. Irradiation dose of 3 kGy was the optimum dose for preservation of mixed peas with diced carrot which extended the refrigeration shelf-life up to 21 days and it was sufficient in eliminating pathogenic bacteria without affecting their sensory quality and with negligible effect on chemical quality.

Lithium zirconate elements fabricated by industrial scale processes

International Nuclear Information System (INIS)

Roux, N.

1991-01-01

Lithium metazirconate Li 2 ZrO 3 is one of the leading tritium breeding ceramics contemplated in solid blanket concepts for fusion reactors. Among its merits are fair physical properties, satisfactory compatibility with structural materials and beryllium, satisfactory mechanical strength, excellent irradiation behaviour as shown by a comparative irradiation of ceramics in the EBR II reactor, and very good tritium release performance as evidenced in the MOZART and EXOTIC neutron irradiations. Pechiney and the CEA are jointly involved in developing industrial fabrication of Li 2 ZrO 3 elements to the microstructural and geometrical specifications required for their use in the solid blankets as conceived in the European Program

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

Mantle irradiation alone for pathologic stage I and II Hodgkin's disease: long-term follow-up and patterns of failure

International Nuclear Information System (INIS)

Liao Zhongxing; Ha, Chul S.; Vlachaki, Maria T.; Hagemeister, Frederick; Cabanillas, Fernando; Hess, Mark; Tucker, Susan; Cox, James D.

2001-01-01

Purpose: We performed a retrospective study to determine the long-term outcome, patterns of failure, and prognostic factors for patients with pathologic Stage I or II Hodgkin's disease (HD) who were treated with mantle irradiation alone. Methods and Materials: The medical records of 145 patients with pathologic Stage I or II supradiaphragmatic Hodgkin's disease treated with mantle irradiation alone between June 1967 and June 1991 were reviewed. Patterns of failure, overall survival (OS) rate, and progression-free survival (PFS) rate were determined. Univariate and multivariate analyses were performed to identify adverse prognostic factors for OS and PFS. The number of adverse prognostic factors per patient was counted, and a prognostic score was assigned to each patient. The log-rank test was used to compare the OS or PFS rates among patients with prognostic scores 0, 1, and 2. Results: The median patient age was 27 years (range 10-66), with almost even male to female distribution. Every patient had splenectomy and negative laparotomy (LAP). Fifty-one patients had Stage I disease (IA-49, IB-2) and 94 Stage II (IIA-89, IIB-5). The histologic subtypes were nodular sclerosing in 110, mixed cellularity in 28, lymphocyte predominance in 5, lymphocyte depleted in 1, and unclassified in 1. Twelve patients with Stage II disease had ≥ 3 sites of nodal involvement. Fifty-four patients had a prognostic score of 0, 70 of 1, and 21 of 2. The median follow-up time for the 109 surviving patients was 146 months (range 25-381). The 10- and 20-year actuarial OS rates for the whole group were 87.6% and 65.3%, respectively. The corresponding actuarial PFS rates were 75.3% and 74.2%, respectively. Thirty-six patients (9 Stage I, 27 Stage II) had relapses in a total of 41 sites. Failures by histology were 29 patients with nodular sclerosing, 6 with mixed cellularity, and 1 with lymphocyte predominance. Failures by sites were: trans-diaphragmatic, 22 (para-aortic nodes, 15; as the only

Atomic mixing of metallic bilayers Ni/Ti irradiated with high energy heavy ions; Etude du melange ionique de bicouches metalliques Ni/Ti irradiees avec des ions lourds de haute energie

Energy Technology Data Exchange (ETDEWEB)

Leguay, R

1994-09-26

We have studied the ionic mixing of Nl(105 angstrom) bilayers irradiated, at 80 and 300 K. with GeV heavy ions. In this energy range, the energy transfer from the incident ions to the target occurs mainly through electronic excitations. We have shown that this energy transfer induces a strong ionic mixing at the Nl/Ti interface. The thickness of the mixed interlayer increases with the fluence. At low fluences (10{sup 12} ions/cm{sup 2}), the Nl/Ti interface is rough ; at higher fluences (10{sup 13} ions/cm{sup 2}) a homogeneous mixed interlayer appears ; and at even higher fluences (some 10{sup 13} ions/cm{sup 2}) a preferential diffusion of Ni into Ti is clearly seen. The characterization techniques used are: (1) electrical resistivity measurements which allow to follow in situ the damage kinetic. (II) neutron and X-ray reflectometry. (III) elaboration of transverse cuts on which was performed energy loss spectroscopy. (II) and (III) allow the determination of the concentration profiles of the different species present in the sample. (IV) transmission electron microscopy on the transverse cuts which gives a direct image of the different layers. (author). 11 refs., 103 figs., 23 tabs., 2 appends.

DNA binding and biological activity of mixed ligand complexes of Cu(II, Ni(II and Co(II with quinolones and N donor ligand

Directory of Open Access Journals (Sweden)

S.M M Akram

2015-10-01

Full Text Available  AbstractMixed ligand complexes of  Cu(II, Ni(II and Co(II have been synthesized by using levofloxacin and bipyridyl and characterized using spectral and analytical techniques. The binding behavior of the Ni(II and Cu(II complexes with herring sperm DNA(Hs-DNA were determined using electronic absorption titration, viscometric measurements and cyclic voltammetry measurements. The binding constant calculated  for Cu(II and Ni(II complexes are 2.0 x 104 and 4.0 x 104 M-1 respectively. Detailed analysis reveals that these metal complexes interact with DNA through intercalative binding mode. The nuclease activity of  Cu(II and Ni(II complexes with ct-DNA was carried out using agarose gel electrophoresis technique. The antioxidant activities for the synthesized complexes have been tested and the antibacterial activity for Ni(II complex was also checked.Key words: Intercalation, hypochromism, red shift and  peak potential.

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

Observation of oscillatory radiation induced segregation profiles at grain boundaries in neutron irradiated 316 stainless steel using atom probe tomography

Science.gov (United States)

Barr, Christopher M.; Felfer, Peter J.; Cole, James I.; Taheri, Mitra L.

2018-06-01

Radiation induced segregation in austenitic Fe-Ni-Cr stainless steels is a key detrimental microstructural modification experienced in the current generation of light water reactors. In particular, Cr depletion at grain boundaries can be a significant factor in irradiation-assisted stress corrosion cracking. Therefore, having a complete knowledge and mechanistic understanding of radiation induced segregation at high dose and after a long thermal history is desired for continued sustainability of existing reactors. Here, we examine a 12% cold worked AISI 316 stainless steel hexagonal duct exposed in the lower dose, outer blanket region of the EBR-II reactor, by using advanced characterization and analysis techniques including atom probe tomography and analytical scanning transmission electron microscopy. Contrary to existing literature, we observe an oscillatory w-shape Cr and M-shape Ni concentration profile at 31 dpa. The presence and characterization through advanced atom probe tomography analysis of the w-shape Cr RIS profile is discussed in the context of the localized GB plane interfacial excess of the other major and minor alloying elements. The key finding of a co-segregation phenomena coupling Cr, Mo, and C is discussed in the context of the existing solute segregation literature under irradiation with emphasis on improved spatial and chemical resolution of atom probe tomography.

Detection of irradiation history of seasoning mixes composed of dried fish and its extract. TL analysis and application considerations for mineral separation from foods

International Nuclear Information System (INIS)

Sekiguchi, Masayuki; Nakagawa, Seiko; Yunoki, Syunji

2009-01-01

In the present study, the thermoluminescence (TL) method (EN1788) was used to detect the irradiation treatment of Japanese traditional seasoning mixes mainly composed of dried fish (bonito etc.) and its extract. The Glow 1 curves of minerals separated from the seasoning mixes using a heavy liquid showed significant single peaks at temperatures between 146.5degC and 175.4degC. The peaks are typical for irradiated food, despite the samples being not irradiated. The Glow 2 curves showed single peaks at temperatures higher than that of the Glow 1 curves (175.4degC to 217.9degC). The peak temperature of Glow 2 is usually lower than that of Glow 1 because the peaks from irradiated silicate minerals shift to higher temperatures with time. The TL glow ratios (Glow1/Glow2) calculated in the temperature ranges (167-232degC) defined by means of irradiated TLD-100 were above 0.1, suggesting that the mineral samples were contaminated with organic materials such as protein or bio-inorganic materials such as bone. In order to remove the possible contaminants, acid hydrolysis and subsequent heavy liquid separation were employed. The minerals thus obtained showed no Glow 1 peaks. A significant peak was observed at 213degC instead of peaks at lower temperature for the case of irradiated seasoning mixes (2.45 kGy). These results suggest that the TL method may provide false positives for the Japanese traditional seasoning mixes because of luminescence from some components other than silicate minerals. It appears that EN1788 needs some modification to precisely detect food irradiation for the seasoning mixes. (author)

Effects of 24-epibrassinolide pre-treatment on UV-B-induced changes in the pigment content of pea leaves

International Nuclear Information System (INIS)

Dobrikova, A.; Vladkova, R.; Stanoeva, D.; Popova, A.; Velitchkova, M.

2013-01-01

In the present work, the effects of 24-epibrassinolide (EBR) on the UV-B-induced changes in the pigment content of pea leaves were studied. Control (non-EBR-treated) and EBR-treated plants were irradiated with UV-B for 3 h and pigment analysis was performed after 24 and 48 h. The results show that EBR spraying of plants 48 h prior to UV-B exposure alleviates its detrimental effect on chlorophyll a and b (Chl a and Chl b) content in comparison with control pea leaves. An increase in carotenoids (Car) and UV-B absorbing compounds was also observed at low dose of UV-B radiation. For the first time, it is shown that UV-B damage effect on control leaves is accompanied by a significant (more than 50%) increase in their pheophytin a (Pheo a) content 48 h after the UV-B exposure and that the EBR pre-treatment prevents the increase of Pheo a content in UV-B irradiated leaves. In addition, it is demonstrated that EBR application modifies UV-B-induced alterations of energy distribution between the main pigment-protein complexes in pea thylakoid membranes

Development of advanced blanket performance under irradiation and system integration through JUPITER-II project

Energy Technology Data Exchange (ETDEWEB)

Abe, Katsunori; Kohyama, Akira; Tanaka, Satoru; Namba, C.; Terai, T.; Kunugi, T.; Muroga, Takeo; Hasegawa, Akira; Sagara, A.; Berk, S.; Zinkle, Steven J.; Sze, Dai Kai; Petti, D. A.; Abdou, Mohamed A.; Morley, Neil B.; Kurtz, Richard J.; Snead, Lance L.; Ghoniem, Nasr M.

2008-12-01

This report describes an outline of the activities of the JUPITER-II collaboration (japan-USA program of Irradiation/Integration test for Fusion Research-II), Which has bee carried out through six years (2001-2006) under Phase 4 of the collabroation implemented by Amendment 4 of Annex 1 to the DOE (United States Department of Energy)-MEXT (Ministry of Education ,Culture,Sports,Science and Technology) Cooperation. This program followed the RTNS-II Program (Phase1:1982-4986), the FFTF/MOTA Program (Phase2:1987-1994) and the JUPITER Program (Phase 3: 1995-2000) [1].

Swelling behavior of manganese-bearing AISI 216 steel

International Nuclear Information System (INIS)

Gelles, D.S.; Garner, F.A.

1984-01-01

The inclusion of 8.5 wt % manganese in AISI 216 does not appear to alter the swelling behavior from that found to be typical of austenitic alloys with comparable levels of other austentite-stabilizing elements. The swelling in AISI 216 in EBR-II is quite insensitive to irradiation temperature in the range 400-650 0 C. Microscopy reveals that this may arise from the low level of precipitation that occurs in the alloy

Laser pulse heating of steel mixing with WC particles in a irradiated region

Science.gov (United States)

Shuja, S. Z.; Yilbas, B. S.; Ali, H.; Karatas, C.

2016-12-01

Laser pulse heating of steel mixing with tungsten carbide (WC) particles is carried out. Temperature field in the irradiated region is simulated in line with the experimental conditions. In the analysis, a laser pulse parameter is introduced, which defines the laser pulse intensity distribution at the irradiated surface. The influence of the laser parameter on the melt pool size and the maximum temperature increase in the irradiated region is examined. Surface temperature predictions are compared with the experimental data. In addition, the distribution of WC particles and their re-locations in the treated layer, due to combination of the natural convection and Marangoni currents, are predicted. The findings are compared to the experimental data. It is found that surface temperature predictions agree well with the experimental data. The dislocated WC particles form a streamlining in the near region of the melt pool wall, which agree with the experimental findings. The Gaussian distribution of the laser pulse intensity results in the maximum peak temperature and the maximum flow velocity inside the melt pool. In this case, the melt pool depth becomes the largest as compared to those corresponding to other laser pulse intensity distributions at the irradiated surface.

SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

International Nuclear Information System (INIS)

Wilson, D.R.; Dutt, D.S.

1979-05-01

SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

Flavor democracy and type-II seesaw realization of bilarge neutrino mixing

International Nuclear Information System (INIS)

Rodejohann, Werner; Xing Zhizhong

2004-01-01

We generalize the democratic neutrino mixing ansatz by incorporating the type-II seesaw mechanism with S(3) flavor symmetry. For only the triplet mass term or only the conventional seesaw term large neutrino mixing can be achieved only by assuming an unnatural suppression of the flavor democracy contribution. We show that bilarge neutrino mixing can naturally appear if the flavor democracy term is strongly suppressed due to significant cancellation between the conventional seesaw and triplet mass terms. Explicit S(3) symmetry breaking yields successful neutrino phenomenology and various testable correlations between the neutrino mass and mixing parameters. Among the results are a normal neutrino mass ordering, 0.005= e3 vertical bar = 2 2θ 23 >=0.005, positive J CP and moderate cancellation in the effective mass of the neutrinoless double beta decay

Tunneling and migration of the dumbbell defect in electron-irradiated aluminum-zinc alloys

International Nuclear Information System (INIS)

Wallace, P.W.

1983-01-01

Ultrasonic attenuation and velocity measurements on irradiated Al-Zn alloys (0.01, 0.1, and 0.5 at %) indicate a tunneling relaxation of the predominant mixed dumbbell defect at low temperatures, and mixed dumbbell migration at the Stage II anneal temperature. The effect of an internal strain varying with the zinc concentration on the measured decrement and modulus change is striking. Evaluated in the framework of a six-level system, this reveals the simultaneous action of resonance and nonclassical relaxation processes. Using Fe as a probe atom, it is shown that mixed dumbbell dissociation is in an insignificant component of the annealing of this defect. The decrease of the annealing temperature at higher zinc concentrations provides evidence that the mixed dumbbell migrates as a unit during annealing. The energies associated with dumbbell migration and interstitial escape are derived. Further evidence for the migration mechanism is obtained from successive irradiation and annealing

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

The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1988-01-01

The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

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

Unilateral irradiation of pigs in a mixed neutrons+gamma field. Early results

International Nuclear Information System (INIS)

Lemaitre, Guy; Maas, Jean.

1982-08-01

Pigs (16-20kg) were irradiated with 60 Co gamma or in a mixed field (neutron + gamma from the pulsed reactor SILENE). Pigs were unilaterally exposed by the left side. Each experimental group was composed of twelve animals and one control. Within the dose range explored (reference dose is mid-line tissue dose): 4-9.8 Gy of gamma rays only; 4.6 - 5.7 Gy of neutrons and gamma rays, pigs presented the haematopioetic form of the acute radiation sickness. At 5 Gy mixed field was more harmful than gamma rays only. Therefore the numerical value of neutron RBE (lethality 50 p cent within 30 days) is more than one. Experiments will be carried out in order to determine RBE values more accurately. Bone marrow dose will also be determined [fr

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Effects on the glucose metabolism in type II diabetes model mice treated with dose-rates irradiation

International Nuclear Information System (INIS)

Nomura, Takaharu; Sakai, Kazuo

2004-01-01

The effects of low-dose rate gamma-irradiation on the type II diabetes mellitus were investigated in C57BL/KsJ-ab/db (db mouse). This mouse develops the type II diabetes within 8 weeks of the birth due to a dysfunction of the insulin receptors. As a result the db mouse shows obese and exhibits hyperinsulinism. Ten-week old female mice (12 mice in each group) were irradiated with gamma-rays at 0.35 mGy/hr, 0.65 mGy/hr or 1.2 mGy/hr in the low-dose rate irradiation facility in the Low Dose Radiation Research Center. The level of plasma glucose and insulin was measured. After 2 weeks irradiation, the glucose level slightly increased, however the difference between the irradiated mice and non-irradiated groups was not significant. The plasma insulin concentration decreased in the non-irradiated group to half of the initial level. In the irradiated group, it also decreased but in the group of 0.65 mGy/hr and 0.35 mGy/hr, it was significantly differed from that in the non-irradiated group. In the glucose tolerance test, plasma glucose level increased shortly after 0.1 mg/head glucose injection by mouth and reached to a peak at 90-120 min after the injection. The glucose level of the non-irradiated mice was slightly higher than that of irradiated mice. The plasma insulin level of non-irradiated group was enhanced after the injection and maintained the level during the test. However the levels of irradiated mice were decreased at 30-60 min after the injection. Both the level of non-irradiated an irradiated was almost same but the non-irradiated one was a little high. In all of mice, the plasma insulin level was highly elevated right after the 0.05 units/head insulin injection by i.p. and the levels were also gradually decreased. The level of the non-irradiated group was slowly decreased and was higher than the irradiated mice. The plasma glucose levels of all mice did not change after the test; however, the levels of irradiated mice were slightly lower than that of non-irradiated

Formation of Mixed-Ligand Complexes of Metals(II) with Monoamine Complexones and Amino Acids in Solution

Science.gov (United States)

Pyreu, D. F.; Gridchin, S. N.

2018-05-01

The formation of mixed-ligand complexes in the M(II)-Nta, Ida-L (M = Cu(II), Ni, Zn, Co(II), L = Ser, Thr, Asp, Arg, Asn) systems, where Ida and Nta are the residues of iminodiacetic and nitrilotriacetic acids, respectively, is studied using pH measurements, calorimetry and spectrophotometry. The thermodynamic parameters (log K, Δr G 0, Δr H, Δr S) of their formation at 298.15 K and ionic strength I = 0.5 (KNO3) are determined. The most likely scenario of amino acid residue coordination in the composition of mixed complexes is discussed.

Status of RBCB testing of LMR oxide fuel in EBR-II

International Nuclear Information System (INIS)

Strain, R.V.; Bottcher, J.H.; Gross, K.C.; Lambert, J.D.B.; Ukai, S.; Nomura, S.; Shikakura, S.; Katsuragawa, M.

1991-01-01

The status is given of the the American-Japanese collaborative program in Experimental Breeder Reactor 2 to determine the run-beyond-cladding-breach performance of (UPu)O 2 fuel pins for liquid-metal cooled reactors. Phase 1 of the collaboration involved eighteen irradiation tests over 1981--86 with 5.84-mm pins in 316 or D9 stainless steel. Emphasis in Phase 2 tests from 1989 onwards is with larger diameter (7.5mm) pins in advanced claddings. Results include delayed neutron and fission gas release data from breached pins, the impact of fuel-sodium reaction product formation on pin performance, and fuel and fission product contamination from failures. 13 refs, 1 fig., 4 tabs

Recovery Effect and Life Prolong Effect of Long Term Low-Dose Rate Irradiation on Type II Diabetes Model Mice

International Nuclear Information System (INIS)

Nomura, T.; Makino, N.; Oda, T.; Suzuki, I.; Sakai, K

2004-01-01

The effects of low-dose rate gamma-irradiation were investigated on model mice for type II diabetes mellitus, C57BL/KsJ-db/db. The mice develop the type II diabetes by 10 weeks of age due to obesity and are characterized by hyperinsulinemia. Female 10-week old mice, a group of 12 mice, were irradiated at 0.65 mGy/hr from 137-Cs (370 GBq). The urine glucose levels of all of the mice were strongly positive at the beginning of the irradiation. In the irradiated group, the decrease in the glucose level was observed in 3 mice. Such recovery from the diabetes was never observed in 12 mice of non-irradiated control group. There is no systematic difference in the change of body weight, food assumption, and amount of drinking water, between the irradiated group and the non-irradiated group or between the recovered mice and the non-recovered mice. The survival was better in the irradiated group: the surviving fraction at the age of 90 weeks was 75% in the irradiated group, while 40% in the non-irradiated. Marked difference was also observed in the appearance of the coat hair, skin, and tail; better condition was kept in the irradiated group. In the irradiated mice mortality was delayed and the healthy appearance was prolonged in the irradiated mice by about 20 ? 30 weeks compared with the non-irradiated mice. These results suggest that the low-dose irradiation modified the condition of the diabetic mice, which lead not only to the recovery of the diabetes, but also to the suppression of the aging process. (Author)

Effectiveness of thermoluminescence analysis to detect low quantity of gamma-irradiated component in non-irradiated mushroom powders

International Nuclear Information System (INIS)

Akram, Kashif; Ahn, Jae-Jun; Shahbaz, Hafiz Muhammad; Jo, Deokjo; Kwon, Joong-Ho

2013-01-01

Gamma-irradiated (0–10 kGy) dried mushrooms (Lentinus edodes) powders were mixed at different ratios (1–10%) in the non-irradiated samples and investigated using photostimulated-luminescence (PSL), electron spin resonance (ESR) and thermoluminescence (TL) techniques. The PSL results were negative for all samples at 1% mixing ratio, whereas intermediate results were observed for the samples containing 5% or 10% irradiated component with the exception (positive) of 10% mixing of 10 kGy-irradiated sample. The ESR analysis showed the presence of crystalline sugar radicals in the irradiated samples but the radiation-specific spectral features were absent in the mixed samples. TL analysis showed the radiation-specific TL glow curves; however, the complicated results were observed at 1% mixing of 2 and 5 kGy-irradiated samples, which required careful evaluations to draw the final conclusion about the irradiation status of the samples. TL ratios could only confirm the results of samples with 5% and 10% mixing of 10 kGy, and 10% mixing of 5 kGy-irradiated components. SEM-EDX analysis showed that feldspar and quartz were major contaminating minerals, responsible for the radiation-specific luminescence characteristics. -- Highlights: ► Detection of irradiated food is important to enforce the applied regulations. ► The effectiveness of TL analysis was investigated to detect irradiated component. ► The TL results were compared with those from PSL and ESR analysis. ► TL analysis was most effective to characterize the irradiation status of samples. ► SEM-EDX analysis showed feldspar and quartz as the main source of TL properties

Effectiveness of thermoluminescence analysis to detect low quantity of gamma-irradiated component in non-irradiated mushroom powders

Energy Technology Data Exchange (ETDEWEB)

Akram, Kashif [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Institute of Food Science and Nutrition, University of Sargodha, Sargodha 40100 (Pakistan); Ahn, Jae-Jun [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shahbaz, Hafiz Muhammad [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Institute of Food Science and Nutrition, University of Sargodha, Sargodha 40100 (Pakistan); Jo, Deokjo [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kwon, Joong-Ho, E-mail: jhkwon@knu.ac.kr [School of Food Science and Biotechnology, Kyungpook National University, Daegu 702-701 (Korea, Republic of)

2013-04-15

Gamma-irradiated (0–10 kGy) dried mushrooms (Lentinus edodes) powders were mixed at different ratios (1–10%) in the non-irradiated samples and investigated using photostimulated-luminescence (PSL), electron spin resonance (ESR) and thermoluminescence (TL) techniques. The PSL results were negative for all samples at 1% mixing ratio, whereas intermediate results were observed for the samples containing 5% or 10% irradiated component with the exception (positive) of 10% mixing of 10 kGy-irradiated sample. The ESR analysis showed the presence of crystalline sugar radicals in the irradiated samples but the radiation-specific spectral features were absent in the mixed samples. TL analysis showed the radiation-specific TL glow curves; however, the complicated results were observed at 1% mixing of 2 and 5 kGy-irradiated samples, which required careful evaluations to draw the final conclusion about the irradiation status of the samples. TL ratios could only confirm the results of samples with 5% and 10% mixing of 10 kGy, and 10% mixing of 5 kGy-irradiated components. SEM-EDX analysis showed that feldspar and quartz were major contaminating minerals, responsible for the radiation-specific luminescence characteristics. -- Highlights: ► Detection of irradiated food is important to enforce the applied regulations. ► The effectiveness of TL analysis was investigated to detect irradiated component. ► The TL results were compared with those from PSL and ESR analysis. ► TL analysis was most effective to characterize the irradiation status of samples. ► SEM-EDX analysis showed feldspar and quartz as the main source of TL properties.

Mixed features in patients with a major depressive episode: the BRIDGE-II-MIX study.

Science.gov (United States)

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

2015-03-01

To estimate the frequency of mixed states in patients diagnosed with major depressive episode (MDE) according to conceptually different definitions and to compare their clinical validity. This multicenter, multinational cross-sectional Bipolar Disorders: Improving Diagnosis, Guidance and Education (BRIDGE)-II-MIX study enrolled 2,811 adult patients experiencing an MDE. Data were collected per protocol on sociodemographic variables, current and past psychiatric symptoms, and clinical variables that are risk factors for bipolar disorder. The frequency of mixed features was determined by applying both DSM-5 criteria and a priori described Research-Based Diagnostic Criteria (RBDC). Clinical variables associated with mixed features were assessed using logistic regression. Overall, 212 patients (7.5%) fulfilled DSM-5 criteria for MDE with mixed features (DSM-5-MXS), and 818 patients (29.1%) fulfilled diagnostic criteria for a predefined RBDC depressive mixed state (RBDC-MXS). The most frequent manic/hypomanic symptoms were irritable mood (32.6%), emotional/mood lability (29.8%), distractibility (24.4%), psychomotor agitation (16.1%), impulsivity (14.5%), aggression (14.2%), racing thoughts (11.8%), and pressure to keep talking (11.4%). Euphoria (4.6%), grandiosity (3.7%), and hypersexuality (2.6%) were less represented. In multivariate logistic regression analysis, RBDC-MXS was associated with the largest number of variables including diagnosis of bipolar disorder, family history of mania, lifetime suicide attempts, duration of the current episode > 1 month, atypical features, early onset, history of antidepressant-induced mania/hypomania, and lifetime comorbidity with anxiety, alcohol and substance use disorders, attention-deficit/hyperactivity disorder, and borderline personality disorder. Depressive mixed state, defined as the presence of 3 or more manic/hypomanic features, was present in around one-third of patients experiencing an MDE. The valid symptom, illness

Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

International Nuclear Information System (INIS)

Fenneker; Steinmetz; Toebbe

1986-07-01

The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

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

Transient redistribution of intragranular fission gas in irradiated mixed oxide

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1981-01-01

Safety analyses for an LMFBR require a knowledge of the fuel and fission gas behavior under transient conditions. Analyses of microstructural data derived from transiently heated, irradiated, mixed oxide fuel specimens have allowed the calculation of the degree of nonequilibrium of intragranular bubbles formed during the transient. It is hypothesized that the observed over-pressurization of the intragranular bubbles mechanically loads the fuel within the grain, leading to a stress gradient derived force upon near-grain-surface bubbles, driving them preferentially to the grain boundaries. Using existing models for forced diffusion it can be estimated that the stress derived forces on bubbles are within the same magnitude, and possibly greater, than the forces derived from the thermal gradient

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

Influence of Mixed Mode I-Mode II Loading on Fatigue Delamination Growth Characteristics of a Graphite Epoxy Tape Laminate

Science.gov (United States)

Ratcliffe, James G.; Johnston, William M., Jr.

2014-01-01

Mixed mode I-mode II interlaminar tests were conducted on IM7/8552 tape laminates using the mixed-mode bending test. Three mixed mode ratios, G(sub II)/G(sub T) = 0.2, 0.5, and 0.8, were considered. Tests were performed at all three mixed-mode ratios under quasi-static and cyclic loading conditions, where the former static tests were used to determine initial loading levels for the latter fatigue tests. Fatigue tests at each mixed-mode ratio were performed at four loading levels, Gmax, equal to 0.5G(sub c), 0.4G(sub c), 0.3G(sub c), and 0.2G(sub c), where G(sub c) is the interlaminar fracture toughness of the corresponding mixed-mode ratio at which a test was performed. All fatigue tests were performed using constant-amplitude load control and delamination growth was automatically documented using compliance solutions obtained from the corresponding quasi-static tests. Static fracture toughness data yielded a mixed-mode delamination criterion that exhibited monotonic increase in Gc with mixed-mode ratio, G(sub II)/G(sub T). Fatigue delamination onset parameters varied monotonically with G(sub II)/G(sub T), which was expected based on the fracture toughness data. Analysis of non-normalized data yielded a monotonic change in Paris law exponent with mode ratio. This was not the case when normalized data were analyzed. Fatigue data normalized by the static R-curve were most affected in specimens tested at G(sub II)/G(sub T)=0.2 (this process has little influence on the other data). In this case, the normalized data yielded a higher delamination growth rate compared to the raw data for a given loading level. Overall, fiber bridging appeared to be the dominant mechanism, affecting delamination growth rates in specimens tested at different load levels and differing mixed-mode ratios.

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

Biological and Irradiation Treatment of Mix Industrial Wastewater in Flood Mitigation Pond at Prai Industrial Zone

International Nuclear Information System (INIS)

Khomsaton Abu Bakar; Jamaliah Sharif; Selambakkanu, S.; Ming, T.M.; Natasha Isnin; Hasnul Nizam Osman; Khasmidatul Akma Mohd Khairul Azmi

2014-01-01

In this work, activated sludge process and E-Beam was used to treat mixed industrial waste water from mitigation pond A. The objectives of this study to analyze the effect of mix liquor volatile suspended solid (MLVSS) concentration on the properties of wastewater and duration of time taken to achieve steady stage condition for biological treatment. Besides that, effect of electron beam energy on the characteristic of wastewater after irradiation with electron beam machine EPS 3000 was studied as well. The result shows removal percentage of COD, suspended solid and color was linearly proportional with MLVSS. Maximum reduction values recorded for COD, suspended solid and color removal was 69.4, 73.0 and 43.7 % respectively with 3500 mg/l MLVSS at 48 h HRT. In irradiation treatment, significant reduction of COD was obtained with the increase of electron beam energy but the results for suspended solid and color was not favorable. (author)

Sesquiterpene lactone mix as a diagnostic tool for Asteraceae allergic contact dermatitis: chemical explanation for its poor performance and Sesquiterpene lactone mix II as a proposed improvement.

Science.gov (United States)

Jacob, Mathias; Brinkmann, Jürgen; Schmidt, Thomas J

2012-05-01

Two preparations are currently in use for the diagnosis of allergic contact dermatitis caused by Asteraceae: (i) Sesquiterpene lactone (SL) mix [three pure sesquiterpene lactones (STLs)], whose use has been questioned, owing to an insufficient rate of true-positive results; and (ii) Compositae mix, consisting of five Asteraceae extracts, which is problematic because of lack of standardization and questionable reproducibility. To analyse the reasons for the narrow sensitivity of SL mix from a chemoinformatic point of view, and to propose a solution by rational selection of alternative constituents for a new SL mix II covering a broader cohort of allergic patients. Structural and biological information on allergenic STLs was retrieved from databases and the literature, and molecular modelling and chemoinformatic computations were performed. An explanation for the insufficient hit rate of SL mix is that the three constituents possess extremely similar molecular structures/properties and do not represent well the structural diversity of allergenic STLs. STLs that are known as constituents of Compositae mix plants show much a wider diversity, which explains the higher positive rate. On the basis of their positions in chemical property space, a new collection of STLs that more evenly cover the overall structural diversity spectrum is proposed. SL mix II is likely to detect a larger number of patients sensitized to Asteraceae. © 2012 John Wiley & Sons A/S.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Determination of mixed stability constants of lead(II/uranyl(II-NTA-cysteine complexes by paper electrophoresis

Directory of Open Access Journals (Sweden)

Brij Bhushan Tewari

2005-12-01

Full Text Available A method involving the use of paper ionophoresis is described for the study of equilibria in mixed – ligand complex systems in solution. The technique is based on the movement of a spot of metal ion under an electric field with the complexants added to the background electrolyte at pH 8.5. The stability constants of the complexes Pb(II – nitrilotriacetate – cysteine and UO2(II – nitrilotriacetate – cysteine are found to be 5.35 plus or minus 0.02 and 6.27 plus or minus 0.07 (logarithm of stability constant values at ionic strength 0.1 M and a temperature of 35 0C.

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.

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

Effects of excited state mixing on transient absorption spectra in dimers Application to photosynthetic light-harvesting complex II

CERN Document Server

Valkunas, L; Trinkunas, G; Müller, M G; Holzwarth, A R

1999-01-01

The excited state mixing effect is taken into account considering the difference spectra of dimers. Both the degenerate (homo) dimer as well as the nondegenerate (hetero) dimer are considered. Due to the higher excited state mixing with the two-exciton states in the homodimer, the excited state absorption (or the difference spectrum) can be strongly affected in comparison with the results obtained in the Heitler-London approximation. The difference spectrum of the heterodimer is influenced by two resonance effects (i) mixing of the ground state optical transitions of both monomers in the dimer and (ii) mixing of the excited state absorption of the excited monomer with the ground state optical transition in the nonexcited monomer. These effects have been tested by simulating the difference absorption spectra of the light-harvesting complex of photosystem II (LHC II) experimentally obtained with the 60 fs excitation pulses at zero delay times and various excitation wavelengths. The pairs of coupled chlorophylls...

Recovery in stages I and II of thermal and fission neutron irradiated molybdenum

International Nuclear Information System (INIS)

Coltman, R.R. Jr.; Klabunde, C.E.; Redman, J.K.

1975-01-01

The influence of initial dose and irradiation doping upon the recovery of Mo was studied for the markedly different types of damage produced by thermal and fission neutrons. The features of the Stage I recovery commonly seen for several fcc metals can be identified, including a I/sub D/ peak at 40 0 K. The typical dose-dependent behavior of a I/sub E/ subpeak was observed at approximately 47 0 K, and evidence for free interstitial migration is further supported by irradiation doping results. Stage II shows a first-order peak at 120 0 K in which the population percentage increases with increasing initial dose in opposite fashion to fcc impurity detrapping peaks. (auth)

The influence of the UV irradiation intensity on photocatalytic activity of ZnAl layered double hydroxides and derived mixed oxides

Directory of Open Access Journals (Sweden)

Hadnađev-Kostić Milica S.

2012-01-01

Full Text Available Layered double hydroxides (LDHs have been studied to a great extent as environmental-friendly complex materials that can be used as photocatalysts or photocatalyst supports. ZnAl layered double hydroxides and their derived mixed oxides were chosen for the investigation of photocatalytic performances in correlation with the UV intensities measured in the South Pannonia region. Low supersaturation coprecipitation method was used for the ZnAl LDH synthesis. For the characterization of LDH and thermal treated samples powder X-ray diffraction (XRD, scanning electron microscopy (SEM, electron dispersive spectroscopy (EDS, nitrogen adsorption-desorption were used. The decomposition of azodye, methylene blue was chosen as photocatalytic test reaction. The study showed that the ZnAl mixed oxide obtained by thermal decomposition of ZnAl LDH has stable activity in the broader UV light irradiation range characterizing the selected region. Photocatalytic activity could be mainly attributed to the ZnO phase, detected both in LDH and thermally treated samples. The study showed that the ZnAl mixed oxide obtained by the calcination of ZnAl LDH has a stable activity within the measured UV light irradiation range; whereas the parent ZnAl LDH catalyst did not perform satisfactory when low UV irradiation intensity is implied.

Determination of D10 values of single and mixed cultures of bacteria after gamma irradiation

International Nuclear Information System (INIS)

Adu-Gyamfi, A.; Nketsia-Tabiri, J.; Boatin, R.

2009-01-01

The D 10 value of bacteria represents the absorbed radiation dose required to inactivate 90 % of a viable population or reduce the population by a factor of 10. D 10 values of 3 bacterial isolates (Escherichia coli, Staphylococcus aureus and Salmonella parathyphi B) were determined using single and mixed cultures to assess the effect of microbial competition on radiosensitivity. The isolates were inoculated into wakye substrate and exposed to γ-radiation doses of 0, 100, 300, 450, 600, 750, 850 Gy from a 6O Co source at a dose rate of 2.20 kGy/h in air. Enumeration of survivors of the isolates was carried out using serial dilution and pour plate methods. The surviving fraction of isolates decreased with increased irradiation doses. D 10 values of E. coli, S. aureus and S. parathyphi B were respectively 0.27, 0.33 and 0.44 kGy when inoculated as single cultures, and 0.24, 0.28 and 0.32 kGy respectively when inoculated as mixed cultures. D 10 values were lower for mixed cultures compared to single cultures, which might indicate reduced resistance to γ-radiation as a result of competition among the isolates. Microbiological challenge tests based on the D 10 values may result in delivery of higher irradiation doses, but the extra dose could serve as safety margin to enhance the food preservative capacity of radiation processing. (au)

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.

Effect of Compressive Mode I on the Mixed Mode I/II Fatigue Crack Growth Rate of 42CrMo4

Science.gov (United States)

Heirani, Hasan; Farhangdoost, Khalil

2018-01-01

Subsurface cracks in mechanical contact loading components are subjected to mixed mode I/II, so it is necessary to evaluate the fatigue behavior of materials under mixed mode loading. For this purpose, fatigue crack propagation tests are performed with compact tension shear specimens for several stress intensity factor (SIF) ratios of mode I and mode II. The effect of compressive mode I loading on mixed mode I/II crack growth rate and fracture surface is investigated. Tests are carried out for the pure mode I, pure mode II, and two different mixed mode loading angles. On the basis of the experimental results, mixed mode crack growth rate parameters are proposed according to Tanaka and Richard with Paris' law. Results show neither Richard's nor Tanaka's equivalent SIFs are very useful because these SIFs depend strongly on the loading angle, but Richard's equivalent SIF formula is more suitable than Tanaka's formula. The compressive mode I causes the crack closure, and the friction force between the crack surfaces resists against the crack growth. In compressive loading with 45° angle, d a/d N increases as K eq decreases.

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

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

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

Evidence for D0-D(0) mixing using the CDF II detector.

Science.gov (United States)

Aaltonen, T; Adelman, J; Akimoto, T; Albrow, M G; González, B Alvarez; Amerio, S; Amidei, D; Anastassov, A; Annovi, A; Antos, J; Aoki, M; Apollinari, G; Apresyan, A; Arisawa, T; Artikov, A; Ashmanskas, W; Attal, A; Aurisano, A; Azfar, F; Azzi-Bacchetta, P; Azzurri, P; Bacchetta, N; Badgett, W; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Baroiant, S; Bartsch, V; Bauer, G; Beauchemin, P-H; Bedeschi, F; Bednar, P; Behari, S; Bellettini, G; Bellinger, J; Belloni, A; Benjamin, D; Beretvas, A; Beringer, J; Berry, T; Bhatti, A; Binkley, M; Bisello, D; Bizjak, I; Blair, R E; Blocker, C; Blumenfeld, B; Bocci, A; Bodek, A; Boisvert, V; Bolla, G; Bolshov, A; Bortoletto, D; Boudreau, J; Boveia, A; Brau, B; Bridgeman, A; Brigliadori, L; Bromberg, C; Brubaker, E; Budagov, J; Budd, H S; Budd, S; Burkett, K; Busetto, G; Bussey, P; Buzatu, A; Byrum, K L; Cabrera, S; Campanelli, M; Campbell, M; Canelli, F; Canepa, A; Carlsmith, D; Carosi, R; Carrillo, S; Carron, S; Casal, B; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chang, S H; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Chlebana, F; Cho, K; Chokheli, D; Chou, J P; Choudalakis, G; Chuang, S H; Chung, K; Chung, W H; Chung, Y S; Ciobanu, C I; Ciocci, M A; Clark, A; Clark, D; Compostella, G; Convery, M E; Conway, J; Cooper, B; Copic, K; Cordelli, M; Cortiana, G; Crescioli, F; Cuenca Almenar, C; Cuevas, J; Culbertson, R; Cully, J C; Dagenhart, D; Datta, M; Davies, T; de Barbaro, P; De Cecco, S; Deisher, A; De Lentdecker, G; De Lorenzo, G; Dell'orso, M; Demortier, L; Deng, J; Deninno, M; De Pedis, D; Derwent, P F; Di Giovanni, G P; Dionisi, C; Di Ruzza, B; Dittmann, J R; D'Onofrio, M; Donati, S; Dong, P; Donini, J; Dorigo, T; Dube, S; Efron, J; Erbacher, R; Errede, D; Errede, S; Eusebi, R; Fang, H C; Farrington, S; Fedorko, W T; Feild, R G; Feindt, M; Fernandez, J P; Ferrazza, C; Field, R; Flanagan, G; Forrest, R; Forrester, S; Franklin, M; Freeman, J C; Furic, I; Gallinaro, M; Galyardt, J; Garberson, F; Garcia, J E; Garfinkel, A F; Gerberich, H; Gerdes, D; Giagu, S; Giakoumopolou, V; Giannetti, P; Gibson, K; Gimmell, J L; Ginsburg, C M; Giokaris, N; Giordani, M; Giromini, P; Giunta, M; Glagolev, V; Glenzinski, D; Gold, M; Goldschmidt, N; Golossanov, A; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González, O; Gorelov, I; Goshaw, A T; Goulianos, K; Gresele, A; Grinstein, S; Grosso-Pilcher, C; Grundler, U; Guimaraes da Costa, J; Gunay-Unalan, Z; Haber, C; Hahn, K; Hahn, S R; Halkiadakis, E; Hamilton, A; Han, B-Y; Han, J Y; Handler, R; Happacher, F; Hara, K; Hare, D; Hare, M; Harper, S; Harr, R F; Harris, R M; Hartz, M; Hatakeyama, K; Hauser, J; Hays, C; Heck, M; Heijboer, A; Heinemann, B; Heinrich, J; Henderson, C; Herndon, M; Heuser, J; Hewamanage, S; Hidas, D; Hill, C S; Hirschbuehl, D; Hocker, A; Hou, S; Houlden, M; Hsu, S-C; Huffman, B T; Hughes, R E; Husemann, U; Huston, J; Incandela, J; Introzzi, G; Iori, M; Ivanov, A; Iyutin, B; James, E; Jayatilaka, B; Jeans, D; Jeon, E J; Jindariani, S; Johnson, W; Jones, M; Joo, K K; Jun, S Y; Jung, J E; Junk, T R; Kamon, T; Kar, D; Karchin, P E; Kato, Y; Kephart, R; Kerzel, U; Khotilovich, V; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, S B; Kim, S H; Kim, Y K; Kimura, N; Kirsch, L; Klimenko, S; Klute, M; Knuteson, B; Ko, B R; Koay, S A; Kondo, K; Kong, D J; Konigsberg, J; Korytov, A; Kotwal, A V; Kraus, J; Kreps, M; Kroll, J; Krumnack, N; Kruse, M; Krutelyov, V; Kubo, T; Kuhlmann, S E; Kuhr, T; Kulkarni, N P; Kusakabe, Y; Kwang, S; Laasanen, A T; Lai, S; Lami, S; Lammel, S; Lancaster, M; Lander, R L; Lannon, K; Lath, A; Latino, G; Lazzizzera, I; Lecompte, T; Lee, J; Lee, J; Lee, Y J; Lee, S W; Lefèvre, R; Leonardo, N; Leone, S; Levy, S; Lewis, J D; Lin, C; Lin, C S; Linacre, J; Lindgren, M; Lipeles, E; Lister, A; Litvintsev, D O; Liu, T; Lockyer, N S; Loginov, A; Loreti, M; Lovas, L; Lu, R-S; Lucchesi, D; Lueck, J; Luci, C; Lujan, P; Lukens, P; Lungu, G; Lyons, L; Lys, J; Lysak, R; Lytken, E; Mack, P; Macqueen, D; Madrak, R; Maeshima, K; Makhoul, K; Maki, T; Maksimovic, P; Malde, S; Malik, S; Manca, G; Manousakis, A; Margaroli, F; Marino, C; Marino, C P; Martin, A; Martin, M; Martin, V; Martínez, M; Martínez-Ballarín, R; Maruyama, T; Mastrandrea, P; Masubuchi, T; Mattson, M E; Mazzanti, P; McFarland, K S; McIntyre, P; McNulty, R; Mehta, A; Mehtala, P; Menzemer, S; Menzione, A; Merkel, P; Mesropian, C; Messina, A; Miao, T; Miladinovic, N; Miles, J; Miller, R; Mills, C; Milnik, M; Mitra, A; Mitselmakher, G; Miyake, H; Moed, S; Moggi, N; Moon, C S; Moore, R; Morello, M; Movilla Fernandez, P; Mülmenstädt, J; Mukherjee, A; Muller, Th; Mumford, R; Murat, P; Mussini, M; Nachtman, J; Nagai, Y; Nagano, A; Naganoma, J; Nakamura, K; Nakano, I; Napier, A; Necula, V; Neu, C; Neubauer, M S; Nielsen, J; Nodulman, L; Norman, M; Norniella, O; Nurse, E; Oh, S H; Oh, Y D; Oksuzian, I; Okusawa, T; Oldeman, R; Orava, R; Osterberg, K; Pagan Griso, S; Pagliarone, C; Palencia, E; Papadimitriou, V; Papaikonomou, A; Paramonov, A A; Parks, B; Pashapour, S; Patrick, J; Pauletta, G; Paulini, M; Paus, C; Pellett, D E; Penzo, A; Phillips, T J; Piacentino, G; Piedra, J; Pinera, L; Pitts, K; Plager, C; Pondrom, L; Portell, X; Poukhov, O; Pounder, N; Prakoshyn, F; Pronko, A; Proudfoot, J; Ptohos, F; Punzi, G; Pursley, J; Rademacker, J; Rahaman, A; Ramakrishnan, V; Ranjan, N; Redondo, I; Reisert, B; Rekovic, V; Renton, P; Rescigno, M; Richter, S; Rimondi, F; Ristori, L; Robson, A; Rodrigo, T; Rogers, E; Rolli, S; Roser, R; Rossi, M; Rossin, R; Roy, P; Ruiz, A; Russ, J; Rusu, V; Saarikko, H; Safonov, A; Sakumoto, W K; Salamanna, G; Saltó, O; Santi, L; Sarkar, S; Sartori, L; Sato, K; Savoy-Navarro, A; Scheidle, T; Schlabach, P; Schmidt, E E; Schmidt, M A; Schmidt, M P; Schmitt, M; Schwarz, T; Scodellaro, L; Scott, A L; Scribano, A; Scuri, F; Sedov, A; Seidel, S; Seiya, Y; Semenov, A; Sexton-Kennedy, L; Sfyria, A; Shalhout, S Z; Shapiro, M D; Shears, T; Shepard, P F; Sherman, D; Shimojima, M; Shochet, M; Shon, Y; Shreyber, I; Sidoti, A; Sinervo, P; Sisakyan, A; Slaughter, A J; Slaunwhite, J; Sliwa, K; Smith, J R; Snider, F D; Snihur, R; Soderberg, M; Soha, A; Somalwar, S; Sorin, V; Spalding, J; Spinella, F; Spreitzer, T; Squillacioti, P; Stanitzki, M; Denis, R St; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Stuart, D; Suh, J S; Sukhanov, A; Sun, H; Suslov, I; Suzuki, T; Taffard, A; Takashima, R; Takeuchi, Y; Tanaka, R; Tecchio, M; Teng, P K; Terashi, K; Thom, J; Thompson, A S; Thompson, G A; Thomson, E; Tipton, P; Tiwari, V; Tkaczyk, S; Toback, D; Tokar, S; Tollefson, K; Tomura, T; Tonelli, D; Torre, S; Torretta, D; Tourneur, S; Trischuk, W; Tu, Y; Turini, N; Ukegawa, F; Uozumi, S; Vallecorsa, S; van Remortel, N; Varganov, A; Vataga, E; Vázquez, F; Velev, G; Vellidis, C; Veszpremi, V; Vidal, M; Vidal, R; Vila, I; Vilar, R; Vine, T; Vogel, M; Volobouev, I; Volpi, G; Würthwein, F; Wagner, P; Wagner, R G; Wagner, R L; Wagner-Kuhr, J; Wagner, W; Wakisaka, T; Wallny, R; Wang, S M; Warburton, A; Waters, D; Weinberger, M; Wester, W C; Whitehouse, B; Whiteson, D; Wicklund, A B; Wicklund, E; Williams, G; Williams, H H; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, C; Wright, T; Wu, X; Wynne, S M; Yagil, A; Yamamoto, K; Yamaoka, J; Yamashita, T; Yang, C; Yang, U K; Yang, Y C; Yao, W M; Yeh, G P; Yoh, J; Yorita, K; Yoshida, T; Yu, G B; Yu, I; Yu, S S; Yun, J C; Zanello, L; Zanetti, A; Zaw, I; Zhang, X; Zheng, Y; Zucchelli, S

2008-03-28

We measure the time dependence of the ratio of decay rates for the rare decay D{0}-->K{+}pi{-} to the Cabibbo-favored decay D{0}-->K{-}pi;{+}. A signal of 12.7x10;{3} D{0}-->K{+}pi{-} decays was obtained using the Collider Detector at Fermilab II detector at the Fermilab Tevatron with an integrated luminosity of 1.5 fb;{-1}. We measure the D0-D[over ]{0} mixing parameters (R_{D},y{'},x{'2}), and find that the data are inconsistent with the no-mixing hypothesis with a probability equivalent to 3.8 Gaussian standard deviations.

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

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

Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

International Nuclear Information System (INIS)

Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

1990-07-01

A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

Radical unique to gamma-irradiated allspice and cinnamon and its utiliy for detection of irradiated foods

Energy Technology Data Exchange (ETDEWEB)

Uchiyama, S. [National Inst. of Hygienic Sciences, Tokyo (Japan); Sugiki, A.; Kawamura, Y.; Murayama, M.; Saito, Y.

1993-04-15

Gamma-Irradiation at a practical dose level of allspice and cinnamon generates a principal signal (signal I, g-value: 2.0048 approx 2.0050) and a minor signal (signal II) at 30 G lower field from signal I in the electron spin resonance spectrum. Signal I, which was not increased in red pepper by photo-exposure, was increased in allspice and cinnamon by gamma-irradiation, heating and even photo-exposure. Signal II was generated only by gamma-irradiation, was little influenced by humidity and was stable for a long time. The ESR method with signal II was applicable to detection of allspice and cinnamon irradiated at 5 kGy or more for up to 6 months after irradiation, as well as allspice irradiated at 10 kGy or more and cinnamon at 5 kGy or more for up to a year. However, signal intensities of signal II differed to some extent between allspice and cinnamon, and even between varieties of cinnamon. Signals I and II were both enhanced after extraction with methanol. Since the rate of increase in signal I was obviously distinct from that of signal II, the radicals corresponding to these signals were presumed to be located at different positions of the matrix of the spice. The methanolic extracts did not yield a major component common to the spices giving signal II.

Significance of postoperative irradiation for breast cancer

International Nuclear Information System (INIS)

Murai, Nobuko; Ogami, Koji; Nishikawa, Kiyoshi; Koga, Kenji; Waki, Norio; Higashi, Hidefumi; Hayashi, Asami; Shibata, Koichiro; Watanabe, Katsuji

1986-01-01

From 1978 through 1983, 27 patients were treated with surgery followed by irradiation (irradiated group) and 29 with surery alone (non-irradiated group). In the irradiated group, 10 had stage II and 17 stage III; in the non-irradiated group, 25 had stage II and 4 stage III. The most common histology was medullary tubular carcinoma (MTC). There was no significant difference in survivals at 3 years and 5 years between the groups. Similarly, no significant difference was seen among stage II patients. Patients with MTC tended to have worse survivals in the irradiated group than in the non-irradiated group, with no statistically significant difference. Among stage II patients, no major differences in local recurrence were seen between the groups; the incidence of distant metastases tended to be high in the irradiated group. The incidence of both local recurrence and distant metastases for stage III patients showed a tendency to be higher in the irradiated group than in the non-irradiated group. The results indicated no apparent benifit of postoperative irradiation for breast cancer. A randomized clinical trial is needed for the evaluation of postoperative irradiation for breast cancer. (Namekawa, K.)

Secretory activity and cell cycle alteration of alveolar type II cells in the early and late phase after irradiation

International Nuclear Information System (INIS)

Willner, Jochen; Vordermark, Dirk; Schmidt, Michael; Gassel, Andreamaria; Flentje, Michael; Wirtz, Hubert

2003-01-01

Purpose: Type II cells and the surfactant system have been proposed to play a central role in pathogenesis of radiation pneumonitis. We analyzed the secretory function and proliferation parameters of alveolar type II cells in the early (until 24 h) and late phase (1-5 weeks) after irradiation (RT) in vitro and in vivo. Methods and Materials: Type II cells were isolated from rats according to the method of Dobbs. Stimulation of secretion was induced with terbutaline, adenosine triphosphate (ATP), and 12-O-tetradecanoylphorbol-13-acetate (TPA) for a 2-h period. Determination of secretion was performed using 3 H-labeled phosphatidylcholine. For the early-phase analysis, freshly isolated and adherent type II cells were irradiated in vitro with 9-21 Gy (stepwise increase of 3 Gy). Secretion stimulation was initiated 1, 6, 24, and 48 h after RT. For late-phase analysis, type II cells were isolated 1-5 weeks after 18 Gy whole lung or sham RT. Each experiment was repeated at least fivefold. Flow cytometry was used to determine cell cycle distribution and proliferating cell nuclear antigen index. Results: During the early-phase (in vitro) analysis, we found a normal stimulation of surfactant secretion in irradiated, as well as unirradiated, cells. No change in basal secretion and no dose effect were seen. During the late phase, 1-5 weeks after whole lung RT, we observed enhanced secretory activity for all secretagogues and a small increase in basal secretion in Weeks 3 and 4 (pneumonitis phase) compared with controls. The total number of isolated type II cells, as well as the rate of viable cells, decreased after the second post-RT week. Cell cycle alterations suggesting an irreversible G 2 /M block occurred in the second post-RT week and did not resolve during the observation period. The proliferating cell nuclear antigen index of type II cells from irradiated rats did not differ from that of controls. Conclusion: In contrast to literature data, we observed no direct

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

International Nuclear Information System (INIS)

Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

2013-01-01

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm -3 ) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio. (author)

Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

International Nuclear Information System (INIS)

Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

2002-01-01

An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

2010-01-31

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

Fabrication of lithium ceramic pellets, rings and single crystals for irradiation in BEATRIX-II

International Nuclear Information System (INIS)

Slagle, O.D.; Noda, K.; Takahashi, T.

1989-04-01

BEATRIX-II is an IEA sponsored experiment of lithium ceramic solid breeder materials in the FFTF/MOTA. Li 2 O solid pellets and annular ring specimens were fabricated for in-situ tritium release tests. In addition, a series of single crystal and polycrystalline lithium ceramic samples were fabricated to determine the irradiation behavior and beryllium compatibility. 6 refs., 10 figs., 4 tabs

Synthesis, characterization, DNA interaction and antimicrobial screening of isatin-based polypyridyl mixed-ligand Cu(II and Zn(II complexes

Directory of Open Access Journals (Sweden)

NATARAJAN RAMAN

2010-06-01

Full Text Available Several mixed ligand Cu(II/Zn(II complexes using 3-(phenyl-imino-1,3-dihydro-2H-indol-2-one (obtained by the condensation of isatin and aniline as the primary ligand and 1,10-phenanthroline (phen/2,2’-bipyridine (bpy as an additional ligand were synthesized and characterized analytically and spectroscopically by elemental analyses, magnetic susceptibility and molar conductance measurements, as well as by UV–Vis, IR, NMR and FAB mass spectroscopy. The interaction of the complexes with calf thymus (CT DNA was studied using absorption spectra, cyclic voltammetric and viscosity measurements. They exhibit absorption hypochromicity, and the specific viscosity increased during the binding of the complexes to calf thymus DNA. The shifts in the oxidation–reduction potential and changes in peak current on addition of DNA were shown by CV measurements. The Cu(II/Zn(II complexes were found to promote cleavage of pUC19 DNA from the supercoiled form I to the open circular form II and linear form III. The complexes show enhanced antifungal and antibacterial activities compared with the free ligand.

Supported liquid membrane based removal of lead(II) and cadmium(II) from mixed feed: Conversion to solid waste by precipitation.

Science.gov (United States)

Bhatluri, Kamal Kumar; Manna, Mriganka Sekhar; Ghoshal, Aloke Kumar; Saha, Prabirkumar

2015-12-15

Simultaneous removal of two heavy metals, lead(II) and cadmium(II), from mixed feed using supported liquid membrane (SLM) based technique is investigated in this work. The carrier-solvent combination of "sodium salt of Di-2-ethylhexylphosphoric acid (D2EHPA) (4% w/w) in environmentally benign coconut oil" was immobilized into the pores of solid polymeric polyvinylidene fluoride (PVDF) support. Sodium carbonate (Na2CO3) was used as the stripping agent. Carbonate salts of lead(II) and cadmium(II) were formed in the stripping side interface and they were insoluble in water leading to precipitation inside the stripping solution. The transportation of solute is positively affected due to the precipitation. Lead(II) removal was found to be preferential due to its favorable electronic configuration. The conversion of the liquid waste to the solid one was added advantage for the final removal of hazardous heavy metals. Copyright © 2015 Elsevier B.V. All rights reserved.

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

Optimized total body irradiation for induction of renal allograft tolerance through mixed chimerism in cynomolgus monkeys

International Nuclear Information System (INIS)

Kimikawa, Masaaki; Kawai, Tatsuo; Ota, Kazuo

1996-01-01

We previously demonstrated that a nonmyeloablative preparative regimen can induce mixed chimerism and renal allograft tolerance between MHC-disparate non-human primates. The basic regimen includes anti-thymocyte globulin (ATG), total body irradiation (TBI, 300 cGy), thymic irradiation (TI, 700 cGy), splenectomy, donor bone marrow (DBM) infusion, and posttransplant cyclosporine therapy (CYA, discontinued after 4 weeks). To evaluate the importance and to minimize the toxicity of irradiation, kidney allografts were transplanted with various manipulations of the irradiation protocol. Monkeys treated with the basic protocol without TBI and TI did not develop chimerism or long-term allograft survival. In monkeys treated with the full protocol, all six monkeys treated with two fractionated dose of 150 cGy developed chimerism and five monkeys appeared tolerant. In contrast, only two of the four monkeys treated with fractionated doses of 125 cGy developed chimerism and only one monkey survived long term. The degree of lymphocyte depletion in all recipients was proportional to the TBI dose. The fractionated TBI regimen of 150 cGy appears to be the most consistently effective regimen for establishing donor bone marrow cell engraftment and allograft tolerance. (author)

Optimized total body irradiation for induction of renal allograft tolerance through mixed chimerism in cynomolgus monkeys

Energy Technology Data Exchange (ETDEWEB)

Kimikawa, Masaaki; Kawai, Tatsuo; Ota, Kazuo [Tokyo Women`s Medical Coll. (Japan)

1996-12-01

We previously demonstrated that a nonmyeloablative preparative regimen can induce mixed chimerism and renal allograft tolerance between MHC-disparate non-human primates. The basic regimen includes anti-thymocyte globulin (ATG), total body irradiation (TBI, 300 cGy), thymic irradiation (TI, 700 cGy), splenectomy, donor bone marrow (DBM) infusion, and posttransplant cyclosporine therapy (CYA, discontinued after 4 weeks). To evaluate the importance and to minimize the toxicity of irradiation, kidney allografts were transplanted with various manipulations of the irradiation protocol. Monkeys treated with the basic protocol without TBI and TI did not develop chimerism or long-term allograft survival. In monkeys treated with the full protocol, all six monkeys treated with two fractionated dose of 150 cGy developed chimerism and five monkeys appeared tolerant. In contrast, only two of the four monkeys treated with fractionated doses of 125 cGy developed chimerism and only one monkey survived long term. The degree of lymphocyte depletion in all recipients was proportional to the TBI dose. The fractionated TBI regimen of 150 cGy appears to be the most consistently effective regimen for establishing donor bone marrow cell engraftment and allograft tolerance. (author)

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)

Differences in TLD 600 and TLD 700 glow curves derived from distict mixed gamma/neutron field irradiations

International Nuclear Information System (INIS)

Cavalieri, Tassio A.; Castro, Vinicius A.; Siqueira, Paulo T.D.

2013-01-01

In Neutron Capture Therapy, a thermal neutron beam shall impinge on a specific nuclide, such as 10 B, to promote a nuclear reaction which releases the useful therapeutic energy. A nuclear reactor is usually used as the neutron source, and therefore field contaminants such as gamma and high energy neutrons are also present in the field. However, mixed field dosimetry still stands as a challenge in some cases, due to the difficulty to experimentally discriminate the dose from each field component. For the mixed field dosimetry, the International Commission on Radiation end Units (ICRU) recommends the use of detector pairs with different responses for each beam component. The TLD 600/700 pair meets this need, because these LiF detectors have different Li isotopes concentration, with distinct thermal neutron responses because 6 Li presents a much higher neutron capture cross section than does 7 Li for low energy neutrons. TLD 600 is 6 Li enriched while TLD 700 is 7 Li enriched. However, depending on the neutron spectrum presented in the mixed field, TLD 700 response to thermal neutrons cannot be disregarded. This work aims to study the difference in TLD 600 and TLD 700 glow curves when these TLDs are submitted to mixed fields of different energy spectra and components balance. The TLDs were irradiated in a pure gamma source, and in mixed fields from an AmBe sealed source and from the IPEN/MB-01 reactor. These TLDs were read and had their two main dosimetric regions analyzed to observe the differences in the glow curves of these TLDs in each irradiation. Field components discrimination was achieved through Monte Carlo simulations run with MCNP radiation transport code. (author)

Effects of Particulate Organic Matter Complexation and Photo-Irradiation on the Fate and Toxicity of Mercury(II) in Aqueous Systems

Science.gov (United States)

Gelfond, C. E.; Kocar, B. D.; Carrasquillo, A. J.

2015-12-01

This project investigates how interactions between mercury (Hg) and particulate organic matter (POM) affect the fate, transport, and toxicity of Hg in the environment. Previous studies have evaluated the coordination of dissolved organic matter (DOM) with Hg, but the coordination of POM with Hg has not been thoroughly addressed. Owing to a high density of reactive functional groups, POM will sorb appreciable quantities of Hg, resulting in a large pool of Hg susceptible to organic matter dependent transformations. Particulate organic carbon is also susceptible photolysis, hence chemical changes induced by irradiation by natural sunlight is also important. Further, photo-reduction of Hg(II) to elemental mercury in the presence of DOM has been observed, yet studies examining this process with Hg(II) complexed to POM are less exhaustive. Here, we illustrate that POM derived from fresh plant detritus is a powerful sorbent of Hg(II), and sorbent properties are altered during POM photolysis. Further, we examine redox transformations of Hg(II), and examine functional groups that contribute to mercury association with POM. Batch sorption isotherms of Hg to dark and irradiated POM from ground Phragmites australis ("common reed") were performed and data was collected using ICP-MS. Coordination of Hg to POM was lower in the irradiated samples, resulting from the decrease in Hg-associated (reduced) sulfur bearing functional groups as measured using X-ray adsorption near-edge spectroscopy (XANES) and extended x-ray adsorption fine structure (EXAFS). Further analysis of the dark and irradiated POM was performed using FT-IR microscopy and STXM to determine changes in distribution and alteration of functional groups responsible for Hg sorption to POM.

COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

International Nuclear Information System (INIS)

Rhodes, C.A.

1984-12-01

This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

Replacement of milk fat by mixed vegetable oils in manufacturing soft cheese treated by gamma irradiation

International Nuclear Information System (INIS)

Afifi, E.A.; Anwar, M.M.

2007-01-01

This investigation aimed to study the possibility of substituting milk fat by using blended vegetable oils in manufacturing soft cheese with low salt content, in addition, lo utilize gamma irradiation to prolong the shelf-life of the new manufactured product. Therefore, one hundred (lOOKg) from fresh buffaloes milk containing 5 % milk fal and 3 % salt were divided into tow parts , the first part was used for manufacturing control soft cheese sample (containing milk fat ), while the second part was skimmed, blended with blended vegetable oils and homogenized. The skim homogenized milk containing 5% mixed vegetable oils used for manufacturing soft cheese ( new product filled ). The obtained soft cheese was subjected to 1, 2 and 3 kGy y-irradiation, and stored at refrigerator temperature. During cold storage, the sensory, microbial and chemical properties of control soft cheese and treated one were evaluated. The obtained results indicated that the replacement of milk fat by mixed vegetable oils in the manufacturing soft cheese had no effect on chemical composition and sensory properties except white color and slight oily flavor which have been noticed in treated filled cheese. In addition, irradiation dose of 3 kGy prolonged the shelf-life of treated filled cheese to 42 days compared to 18 days for control sample and scqueiitly, the new product high percentage of iinsaluraled fatly acid and no cholesterol compared with cheese made from natural milk and can be recommended as a healthy food especially for those who need to low or free cholesterol foods

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.

Use of an isotope separator at the INEL

International Nuclear Information System (INIS)

Anderl, R.A.

1977-01-01

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

Tunneling and migration of the dumbbell defect in electron irradiated aluminum-zinc alloys

International Nuclear Information System (INIS)

Wallace, P.W.

1983-01-01

Ultrasonic attenuation and velocity measurements have been made on irradiated Al-Zn alloys (Zn concentrations of .01%, .1%, and .5% atomic). They provide strong evidence for a tunneling relaxation of the predominant mixed dumbbell defect at low temperatures and for mixed dumbbell migration at the Stage II anneal temperature. The effect of an internal strain varying with the zinc concentration of the measured decrement and modulus change is striking. Evaluated in the framework of a six level system, this reveals the simultaneous action of resonance and non-classical relaxation processes. Using Fe as a probe atom, it is shown that mixed dumbbell dissociation is in an insignificant component of the annealing of this defect. The decrease of the annealing temperature at higher zinc concentrations provides evidence that the mixed dumbbell migrates as a unit during annealing. The energies associated with dumbbell migration and interstitial escape are derived. Further evidence for the migration mechanism is obtained from successive irradiation and annealing measurements on the Al-Zn .01% alloy, and from a comparison of these results with published radiation damage rate measurements of dilute Al-Zn alloys

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

Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

International Nuclear Information System (INIS)

Mingliang Chen

2006-01-01

R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

PIREX II, a new irradiation facility for testing fusion first wall materials

International Nuclear Information System (INIS)

Marmy, P.; Daum, M.; Gavillet, D.; Green, S.; Green, W.V.; Hegedues, F.; Pronnecke, S.; Rohrer, U.; Stiefel, U.; Victoria, M.

1988-12-01

A new irradiation facility, PIREX II, became operational in March 1987. It is located on a dedicated beam line split from the main beam of the 590 MeV proton accelerator at the Paul Scherrer Institute (PSI). Irradiation with protons of this energy introduces simultaneously displacement damage, helium and other impurities. Because of the penetration range of 590 MeV protons, both damage and impurities are homogeneously distributed in the target. The installation has its own beam line optics that can support a proton current of up to 50 μA. At a typical beam density of 4 μA/mm 2 , the damage rate in steels is 0.7 x 10 -5 dpa/sec (dpa: displacements per atom) and the helium production rate is 170 appm He/dpa. Both flat tensile specimens of up to 0.4 mm thickness and tubular fatigue samples of 3 mm diameter can be irradiated. Cooling of the temperatures can be controlled between 100 o and 800 o C. Installation of an in situ low cycle fatigue device is foreseen. Beams of up to 20 μA have been obtained, the beam having approximately a gaussian distribution of elliptical cross section with 4 σ between 0.8 and 3 mm by 10 mm. Irradiations for a dosimetry program have been completed on samples of Al, Cu, Fe, Ni, Au, W, and the 1.4914 ferritic steel. The evaluation of results allows the correct choice of reactions to be used for determining total dose, from the standpoint of half life and gamma energy. A program of irradiations on candidate materials for the Next European Torus (NET) design (Cu and Cu alloys, the 1.4914 ferritic martensitic steel, W and W-Re alloys and Mo alloys), where the above mentioned characteristics of this type of irradiation can be used advantageously, is now under way. (author) 11 figs., 4 tabs., 20 refs

Observation of D⁰-D¯⁰ mixing using the CDF II detector.

Science.gov (United States)

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2013-12-06

We measure the time dependence of the ratio of decay rates for D0→K(+)π(-) to the Cabibbo-favored decay D(0)→K(-)π(+). The charge conjugate decays are included. A signal of 3.3×10(4) D(*+)→π(+)D(0), D(0)→K(+)π(-) decays is obtained with D0 proper decay times between 0.75 and 10 mean D0 lifetimes. The data were recorded with the CDF II detector at the Fermilab Tevatron and correspond to an integrated luminosity of 9.6  fb(-1) for pp¯ collisions at √s=1.96  TeV. Assuming CP conservation, we search for D0-D¯0 mixing and measure the mixing parameters to be R(D)=(3.51±0.35)×10(-3), y'=(4.3±4.3)×10(-3), and x'2=(0.08±0.18)×10(-3). We report Bayesian probability intervals in the x'2-y' plane and find that the significance of excluding the no-mixing hypothesis is equivalent to 6.1 Gaussian standard deviations, providing the second observation of D0-D¯0 mixing from a single experiment.

DNA-repair after irradiation of cells with gamma-rays and neutrons

International Nuclear Information System (INIS)

Altmann, H.

1975-11-01

The structural alterations of calf thymus DNA produced by neutron or gamma irradiation were observed by absorption spectra, sedimentation rate and viscosity measurements. Mixed neutron-gamma irradiation produced fewer single and double strand breaks compared with pure gamma irradiation. RBE-values for mixed neutron-gamma radiation were less than 1, and DNA damage decreased with increasing neutron dose rate. Repair processes of DNA occuring after irradiation were measured in mouse spleen suspensions and human lymphocytes using autoradiographic methods and gradient centrifugations. The number of labelled cells was smaller after mixed neutron-gamma irradiation than after gamma irradiation. The rejoining of strand breaks in alkaline and neutral sucrose was more efficient after gamma irradiation than after mixed neutron-gamma irradiation. Finally, the effect of detergents Tween 80 and Nonident P40 on unscheduled DNA synthesis was studied by autoradiography after mixed neutron-gamma irradiation (Dn=5 krad). The results showed that the DNA synthesis was inhibited by detergent solutions of 0.002%