Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

J. L. Smith

2001-01-01

This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS).

Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

International Nuclear Information System (INIS)

Smith, J. L.

2001-01-01

This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS)

Nuclear fuel bundle disassembly and assembly tool

International Nuclear Information System (INIS)

Yates, J.; Long, J.W.

1975-01-01

A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

Preliminary topical report on comparison reactor disassembly calculations

International Nuclear Information System (INIS)

McLaughlin, T.P.

1975-11-01

Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2-POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherent in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident

Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

International Nuclear Information System (INIS)

Viebrock, J.M.

1979-09-01

A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems

Treatment of spent nuclear fuel L-basin sludge

International Nuclear Information System (INIS)

Westover, B.L.; Oji, L.N.; Martin, H.L.; Nichols, D.M.

1997-01-01

Each production reactor at the DOE Savannah River Site (SRS) has a disassembly basin whose primary purpose is to cool irradiated production fuel and target. The disassembly basins also provide a shielded environment for personnel. Material has historically resided in the basins for 6 to 12 months. Increases in basin storage time have occurred, and have caused the buildup of a sludge layer on the basin floors to be greater than historical levels. The sludge is composed primarily of inorganic oxide and hydroxide corrosion products. The sludge layer has increased the turbidity and conductivity of the basin water, contributed to fuel corrosion, and has impacted fuel handling operations. Initial characterization of the sludge indicates that it is a low-level radioactive aqueous waste. This evaluation looked at methods to separate the sludge into its liquid and solid phases. The experimental data obtained during this evaluation clearly shows that a filtration-based approach to dewatering using an Oberlin pressure filtration unit at SRS is possible. This research task was to identify and optimize filtration and settling parameters pursuant sludge processing. The research specifically addressed: choice of filter aid, filter aid to sludge ratio, choice and dosage of polymer flocculation and settling agents, and the determination of Kynch curve settling parameters. Two commercial perlite filter-aids were identified as the most suitable. Of 11 water soluble flocculating polymers evaluated, 3 cationic commercial types formed stable flocs in the screening tests. In low doses, the flocculating polymers also enhanced sludge particle settling and decreased filtrate turbidity. The filtration cake from the sludge can be solidified to meet waste acceptance and storage criteria. However, the conductivity of the remaining filtrate does not meet Reactor Area Return Water criteria and may require a secondary filtration process. 2 refs., 14 figs., 5 tabs

Assessment of the requirements for placing and maintaining Savannah River Site spent fuel storage basins under International Atomic Energy Agency safeguards

International Nuclear Information System (INIS)

Amacker, O.P. Jr.; Curtis, M.M.; Delegard, C.H.; Hsue, S.T.; Whitesel, R.N.

1997-03-01

The United States is considering the offer of irradiated research reactor spent fuel (RRSF) for international safeguards applied by the International Atomic Energy Agency (IAEA). The offer would be to add one or more spent fuel storage basins to the list of facilities eligible for IAEA safeguards. The fuel to be safeguarded would be stored in basins on the Savannah River Site (SRS). This RRSF potentially can include returns of Material Test Reactor (MTR) VAX fuel from Argentina, Brazil, and Chile (ABC); returns from other foreign research reactors; and fuel from domestic research reactors. Basins on the SRS being considered for this fuel storage are the Receiving Basin for Offsite Fuel (RBOF) and the L-Area Disassembly Basin (L-Basin). A working group of SRS, U.S. Department of Energy International Safeguards Division (NN-44), and National Laboratory personnel with experience in IAEA safeguards was convened to consider the requirements for applying the safeguards to this material. The working group projected the safeguards requirements and described alternatives

AGC-2 Disassembly Report

Energy Technology Data Exchange (ETDEWEB)

William Windes

2014-05-01

The Next Generation Nuclear Plant (NGNP) Graphite Research and Development (R&D) Program is currently measuring irradiated material properties for predicting the behavior and operating performance of new nuclear graphite grades available for use within the cores of new very high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate irradiated graphite performance data for NGNP reactor operating conditions. The AGC experiment is designed to determine the changes to specific material properties such as thermal diffusivity, thermal expansion, elastic modulus, mechanical strength, irradiation induced dimensional change rate, and irradiation creep for a wide variety of nuclear grade graphite types over a range of high temperature, and moderate doses. A series of six capsules containing graphite test specimens will be used to expose graphite test samples to a dose range from 1 to 7 dpa at three different temperatures (600, 900, and 1200°C) as described in the Graphite Technology Development Plan. Since irradiation induced creep within graphite components is considered critical to determining the operational life of the graphite core, some of the samples will also be exposed to an applied load to determine the creep rate for each graphite type under both temperature and neutron flux. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR). AGC-1 and AGC-2 will be irradiated in the south flux trap and AGC-3–AGC-6 will be irradiated in the east flux trap. The change in flux traps is due to NGNP irradiation priorities requiring the AGC experiment to be moved to accommodate Fuel irradiation experiments. After irradiation, all six AGC capsules will be cooled in the ATR Canal, sized for shipment, and shipped to the Materials and Fuels Complex (MFC) where the capsule will be disassembled in the Hot Fuel Examination Facility (HFEF). During disassembly, the metallic

L-Reactor 186-basin cleaning alternatives

International Nuclear Information System (INIS)

Turcotte, M.D.S.

1983-01-01

Operation of L Reactor will necessitate annual cleaning of the L Area 186 basins. Alternatives are presented for sediment discharge due to 186-basin cleaning activities as a basis for choosing the optimal cleaning method. Current cleaning activities (i.e. removal of accumulated sediments) for the P, C and K-Area 186 basins result in suspended solids concentrations in the effluent waters above the NPDES limits, requiring an exemption from the NPDES permit for these short-term releases. The objective of mitigating the 186-basin cleaning activities is to decrease the suspended solids concentrations to within permit limits while continuing satisfactory operation of the basins

Simulation-based disassembly systems design

Science.gov (United States)

Ohlendorf, Martin; Herrmann, Christoph; Hesselbach, Juergen

2004-02-01

Recycling of Waste of Electrical and Electronic Equipment (WEEE) is a matter of actual concern, driven by economic, ecological and legislative reasons. Here, disassembly as the first step of the treatment process plays a key role. To achieve sustainable progress in WEEE disassembly, the key is not to limit analysis and planning to merely disassembly processes in a narrow sense, but to consider entire disassembly plants including additional aspects such as internal logistics, storage, sorting etc. as well. In this regard, the paper presents ways of designing, dimensioning, structuring and modeling different disassembly systems. Goal is to achieve efficient and economic disassembly systems that allow recycling processes complying with legal requirements. Moreover, advantages of applying simulation software tools that are widespread and successfully utilized in conventional industry sectors are addressed. They support systematic disassembly planning by means of simulation experiments including consecutive efficiency evaluation. Consequently, anticipatory recycling planning considering various scenarios is enabled and decisions about which types of disassembly systems evidence appropriateness for specific circumstances such as product spectrum, throughput, disassembly depth etc. is supported. Furthermore, integration of simulation based disassembly planning in a holistic concept with configuration of interfaces and data utilization including cost aspects is described.

Advanced disassembling technique of irradiated driver fuel assembly for continuous irradiation of fuel pins

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

2012-01-01

It was necessary to carry out continuous irradiation tests in order to obtain the irradiation data of high burn-up fuel and high neutron dose material for FaCT (Fast Reactor Cycle Technology Development) project. There, the disassembling technique of an irradiated fuel assembly was advanced in order to realize further continuous irradiation tests. Although the conventional disassembling technique had been cutting a lower end-plug of a fuel pin needed to fix fuel pins to an irradiation vehicle, the advanced disassembling technique did not need cutting a lower end-plug. As a result, it was possible to supply many irradiated fuel pins to various continuous irradiation tests for FaCT project. (author)

Restart of R reactor at SRP

International Nuclear Information System (INIS)

McDonell, W.R.

1983-01-01

Restart of the Savannah River R-Reactor is an alternative to L-Reactor operation for increased production of defense nuclear material. R-Reactor was shut down in 1964 after 11-years operation and has been on standby for 19 years. This report presents a description of R-Reactor operation to serve as a basis for analysis of environmental impacts after restoration to meet current SRP performance standards. R-Reactor operation would differ from L-Reactor operation principally in discharge and recycle of effluent cooling water to Par Pond, rather than direct discharge to the Savannah River by way of Steel Creek. Significant differences in environmental effects could result. A costly renovation program would be required to restore R-Reactor to operability, and the reactor could not contribute to material production before about 1989

T1R3 homomeric sweet taste receptor regulates adipogenesis through Gαs-mediated microtubules disassembly and Rho activation in 3T3-L1 cells.

Directory of Open Access Journals (Sweden)

Yosuke Masubuchi

Full Text Available We previously reported that 3T3-L1 cells express a functional sweet taste receptor possibly as a T1R3 homomer that is coupled to Gs and negatively regulates adipogenesis by a Gαs-mediated but cAMP-independent mechanism. Here, we show that stimulation of this receptor with sucralose or saccharin induced disassembly of the microtubules in 3T3-L1 preadipocytes, which was attenuated by overexpression of the dominant-negative mutant of Gαs (Gαs-G226A. In contrast, overexpression of the constitutively active mutant of Gαs (Gαs-Q227L as well as treatment with cholera toxin or isoproterenol but not with forskolin caused disassembly of the microtubules. Sweetener-induced microtubule disassembly was accompanied by activation of RhoA and Rho-associated kinase (ROCK. This was attenuated with by knockdown of GEF-H1, a microtubule-localized guanine nucleotide exchange factor for Rho GTPase. Furthermore, overexpression of the dominant-negative mutant of RhoA (RhoA-T19N blocked sweetener-induced dephosphorylation of Akt and repression of PPARγ and C/EBPα in the early phase of adipogenic differentiation. These results suggest that the T1R3 homomeric sweet taste receptor negatively regulates adipogenesis through Gαs-mediated microtubule disassembly and consequent activation of the Rho/ROCK pathway.

Integrating Safeguards into the Pit Disassembly and Conversion Facility

International Nuclear Information System (INIS)

Clark, T.G.

2002-01-01

In September 2000, the United States and the Russian Federation entered into an agreement which stipulates each country will irreversibly transform 34 metric tons of weapons-grade plutonium into material which could not be used for weapon purposes. Supporting the Department of Energy's (DOE) program to dispose of excess nuclear materials, the Pit Disassembly and Conversion Facility (PDCF) is being designed and constructed to disassemble the weapon ''pits'' and convert the nuclear material to an oxide form for fabrication into reactor fuel at the separate Mixed Oxide Fuel Fabrication Facility. The PDCF design incorporates automation to the maximum extent possible to facilitate material safeguards, reduce worker dose, and improve processing efficiency. This includes provisions for automated guided vehicle movements for shipping containers, material transport via automated conveyor between processes, remote process control monitoring, and automated Nondestructive Assay product systems

Engineering scale tests of mechanical disassembly and short stroke shearing systems for FBR fuel assembly

International Nuclear Information System (INIS)

Higuchi, Hidetoshi; Kitagaki, Toru; Koizumi, Kenji; Hirano, Hiroyasu; Takeuchi, Masayuki; Washiya, Tadahiro; Kawabe, Yukinari; Kobayashi, Tsuguyuki

2011-01-01

Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called 'NEXT' (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called 'slit-cut' for cutting the wrapper tube and 'crop-cut' for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor defueling and disassembly. Summary status report. Volume 3

International Nuclear Information System (INIS)

Doerge, D.H.; Miller, R.L.; Scotti, K.S.

1986-05-01

This document summarizes information relating to the preparations for defueling the Three Mile Island Unit 2 (TMI-2) reactor and disassembly activities being performed concurrently with decontamination of the facility. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data sysem which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during the period of April 23, 1979 to April 16, 1985, in the completion of activities related to preparation for reactor defueling. Support activities conducted outside of radiation areas are not included within the scope of this report. Computerized reports included in this document are: A chronological summary listing work performed for the period; and summary reports for each major task undertaken in connection with the specific scope of this report. Presented in chronological order for the referenced time period. Manually-assembled table summaries are included for: Labor and exposures by department; and labor and exposures by major activity

Disassembling and modification of RA-3

International Nuclear Information System (INIS)

Tarizzo, R.D.

1990-01-01

The objective of this paper is to describe the partial disassembling and modification of RA-3, called the Modernization Project. It comprises all the technical and administrative steps directly related with this task. The improvement of RA-3 is a result of the lack of 90% enriched uranium obliging a change over to 20% enriched uranium. This brought about design modifications both in fuel elements and the reactor. The presentation of documents for the licensing authority as well as are detailed separately. The modernization project was divided in 25 tasks: 1) changing fuel element support table, 2) changing heat exchanger, 3) repairing of cooling towers, 4) repairing of primary circuit valves, 5) repairing of irradiation channels, 6) construction of a new sampler, 7) changing tangential channel, 8) cleaning and disassembling of reactor (inside), 9) changing continuous demineralizer (ion exchange column), 10) detection of failure in fuel elements, 11) modification of nuclear instrumentation, 12) modification of conventional instrumentation, 13) modification of electrical system, 14) changing telemanipulators, 15) construction of mechanism bridge, 16) changing a primary circuit valve when the heat exchanger is changed too, 17) painting ground floor, hall floor, and pump room floor with epoxy resin levelling, 18) installation of fire alarm system, 19) radioactive liquid discharge, 20) modification of secondary circuit (This task involves: a) installation of a third secondary pump, b) extension of this piping, c) installation of two 12 inch valves to the present cooling towers pools independent, d) installation of filtering system), 21) optimization hot water bed, 22) changing detector support table, 23) removal, decontamination and reinstallation of shielding, 25) changing pneumatic system

Disassembly and physical separation of electric/electronic components layered in printed circuit boards (PCB).

Science.gov (United States)

Lee, Jaeryeong; Kim, Youngjin; Lee, Jae-chun

2012-11-30

Although printed circuit boards (PCBs) contain various elements, only the major elements (i.e., those with content levels in wt% or over grade) of and precious metals (e.g., Ag, Au, and platinum groups) contained within PCBs can be recycled. To recover other elements from PCBs, the PCBs should be properly disassembled as the first step of the recycling process. The recovery of these other elements would be beneficial for efforts to conserve scarce resources, reuse electric/electronic components (EECs), and eliminate environmental problems. This paper examines the disassembly of EECs from wasted PCBs (WPCBs) and the physical separation of these EECs using a self-designed disassembling apparatus and a 3-step separation process of sieving, magnetic separation, and dense medium separation. The disassembling efficiencies were evaluated by using the ratio of grinding area (E(area)) and the weight ratio of the detached EECs (E(weight)). In the disassembly treatment, these efficiencies were improved with an increase of grinder speed and grinder height. 97.7% (E(area)) and 98% (E(weight)) could be accomplished ultimately by 3 repetitive treatments at a grinder speed of 5500 rpm and a grinder height of 1.5mm. Through a series of physical separations, most groups of the EECs (except for the diode, transistor, and IC chip groups) could be sorted at a relatively high separation efficiency of about 75% or more. To evaluate the separation efficiency with regard to the elemental composition, the distribution ratio (R(dis)) and the concentration ratio (R(conc)) were used. 15 elements could be separated with the highest R(dis) and R(conc) in the same separated division. This result implies that the recyclability of the elements is highly feasible, even though the initial content in EECs is lower than several tens of mg/kg. Copyright © 2012 Elsevier B.V. All rights reserved.

Modeling operational behavior of a disassembly line

Science.gov (United States)

Kizilkaya, Elif A.; Gupta, Surendra M.

2004-12-01

In this paper we present a dynamic kanban (pull) system specifically developed for disassembly lines. This type of kanban system is much more complex than the traditional kanban system used in assembly lines. For instance, unlike the assembly line where the external demand occurs only at the last station, the demands in the disassembly case also occur at any of the intermittent stations. The reason is that as a product moves on the disassembly line, various parts are disassembled at every station and accumulated at that station. Therefore, there are as many demand sources as there are number of parts. We consider a case example involving the end-of-life products. Based on the precedence relationships and other criteria such as hazardous properties of the parts, we balance the disassembly line. The results of the disassembly line-balancing problem (DLBP) are used as input to the proposed dynamic kanban system for disassembly line (DKSDL). We compare the performance of the DKSDL to the modified kanban system for disassembly line (MKSDL), which was previously introduced by the authors. We show, via simulation, that the DKSDL is far superior to MKSDL considered.

Task-failure-driven rebalancing of disassembly lines

OpenAIRE

Altekin, Fatma Tevhide; Akkan, Can

2011-01-01

Many reverse-logistics systems that collect and reprocess end-of-life products require a disassembly stage. The nature of variability in incoming products, and damages, which are more likely to occur during disassembly than assembly, create a significant uncertainty in disassembly tasks, namely, possibility of failed tasks. Such failures may lead to some successor tasks being infeasible, which changes work contents of downstream stations. To improve the profitability of such a disassembly lin...

Generation of control sequences for a pilot-disassembly system

Science.gov (United States)

Seliger, Guenther; Kim, Hyung-Ju; Keil, Thomas

2002-02-01

Closing the product and material cycles has emerged as a paradigm for industry in the 21st century. Disassembly plays a key role in a life cycle economy since it enables the recovery of resources. A partly automated disassembly system should adapt to a large variety of products and different degrees of devaluation. Also the amounts of products to be disassembled can vary strongly. To cope with these demands an approach to generate on-line disassembly control sequences will be presented. In order to react on these demands the technological feasibility is considered within a procedure for the generation of disassembly control sequences. Procedures are designed to find available and technologically feasible disassembly processes. The control system is formed by modularised and parameterised control units in the cell level within the entire control architecture. In the first development stage product and process analyses at the sample product washing machine were executed. Furthermore a generalized disassembly process was defined. Afterwards these processes were structured in primary and secondary functions. In the second stage the disassembly control at the technological level was investigated. Factors were the availability of the disassembly tools and the technological feasibility of the disassembly processes within the disassembly system. Technical alternative disassembly processes are determined as a result of availability of the tools and technological feasibility of processes. The fourth phase was the concept for the generation of the disassembly control sequences. The approach will be proved in a prototypical disassembly system.

UNDERWATER ANALYSIS OF IRRADIATED REACTOR SLUGS FOR Co-60 AND OTHER RADIONUCLIDES AT THE SAVANNAH RIVER SITE

International Nuclear Information System (INIS)

CASELLA, VITO

2004-01-01

Co-60 was produced in the Savannah River Site (SRS) reactors in the 1970s, and the irradiated cobalt reactor slugs were stored in a reactor basin at SRS. Since the activity rates of these slugs were not accurately known, assaying was required. A sodium iodide gamma detector was placed above a specially designed air collimator assembly, so that the slug was eight to nine feet from the detector and was shielded by the basin water. Also, 18 curium sampler slugs, used to produce Cm-244 from Pu-239, were to be disposed of with the cobalt slugs. The curium slugs were also analyzed with a High Purity Germanium (HPGE) detector in an attempt to identify any additional radionuclides produced from the irradiation. Co-60 concentrations were determined for reactor disassembly basin cobalt slugs and the 18 curium sampler slugs. The total Co-60 activity of all of the assayed slugs in this work summed to 31,783 curies on 9/15/03. From the Co-60 concentrations of the curium sampler slugs, the irradiation flux was determined for the known irradiation time. The amounts of Pu-238,-239,-240,-241,-242; Am-241,-243; and Cm-242,-244 produced were then obtained based on the original amount of Pu-239 irradiated

Pit disassembly and conversion demonstration environmental assessment and research and development activities

International Nuclear Information System (INIS)

1998-08-01

A significant portion of the surplus plutonium is in the form of pits, a nuclear weapons component. Pits are composed of plutonium which is sealed in a metallic shell. These pits would need to be safely disassembled and permanently converted to an unclassified form that would be suitable for long-term disposition and international inspection. To determine the feasibility of an integrated pit disassembly and conversion system, a Pit Disassembly and Conversion Demonstration is proposed to take place at the Los Alamos National Laboratory (LANL). This demonstration would be done in existing buildings and facilities, and would involve the disassembly of up to 250 pits and conversion of the recovered plutonium to plutonium metal ingots and plutonium dioxide. This demonstration also includes the conversion of up to 80 kilograms of clean plutonium metal to plutonium dioxide because, as part of the disposition process, some surplus plutonium metal may be converted to plutonium dioxide in the same facility as the surplus pits. The equipment to be used for the proposed demonstration addressed in this EA would use some parts of the Advanced Recovery and Integrated Extraction System (ARIES) capability, other existing equipment/capacities, plus new equipment that was developed at other sites. In addition, small-scale R and D activities are currently underway as part of the overall surplus plutonium disposition program. These R and D activities are related to pit disassembly and conversion, MOX fuel fabrication, and immobilization (in glass and ceramic forms). They are described in Section 7.0. On May 16, 1997, the Office of Fissile Materials Disposition (MD) notified potentially affected states and tribes that this EA would be prepared in accordance with NEPA. This EA has been prepared to provide sufficient information for DOE to determine whether a Finding of No Significant Impact (FONSI) is warranted or whether an EIS must be prepared

Impact of different disassembly line balancing algorithms on the performance of dynamic kanban system for disassembly line

Science.gov (United States)

Kizilkaya, Elif A.; Gupta, Surendra M.

2005-11-01

In this paper, we compare the impact of different disassembly line balancing (DLB) algorithms on the performance of our recently introduced Dynamic Kanban System for Disassembly Line (DKSDL) to accommodate the vagaries of uncertainties associated with disassembly and remanufacturing processing. We consider a case study to illustrate the impact of various DLB algorithms on the DKSDL. The approach to the solution, scenario settings, results and the discussions of the results are included.

Assembly and disassembly of the nucleolus during the cell cycle.

Science.gov (United States)

Hernandez-Verdun, Danièle

2011-01-01

The nucleolus is a large nuclear domain in which transcription, maturation and assembly of ribosomes take place. In higher eukaryotes, nucleolar organization in three sub-domains reflects the compartmentation of the machineries related to active or inactive transcription of the ribosomal DNA, ribosomal RNA processing and assembly with ribosomal proteins of the two (40S and 60S) ribosomal subunits. The assembly of the nucleoli during telophase/early G(1) depends on pre-existing machineries inactivated during prophase (the transcription machinery and RNP processing complexes) and on partially processed 45S rRNAs inherited throughout mitosis. In telophase, the 45S rRNAs nucleate the prenucleolar bodies and order the dynamics of nucleolar assembly. The assembly/disassembly processes of the nucleolus depend on the equilibrium between phosphorylation/dephosphorylation of the transcription machinery and on the RNP processing complexes under the control of the CDK1-cyclin B kinase and PP1 phosphatases. The dynamics of assembly/disassembly of the nucleolus is time and space regulated.

Multikanban model for disassembly line with demand fluctuation

Science.gov (United States)

Udomsawat, Gun; Gupta, Surendra M.; Al-Turki, Yousef A. Y.

2004-02-01

In recent years, the continuous growth in consumer waste and dwindling natural resources has seriously threatened the environment. Realizing this, several countries have passed regulations that force manufacturers not only to manufacture environmentally conscious products, but also to take back their used products from consumers so that the components and materials recovered from the products may be reused and/or recycled. Disassembly plays an important role in product recovery. A disassembly line is perhaps the most suitable setting for disassembly of products in large quantities. Because a disassembly line has a tendency to generate excessive inventory, employing a kanban system can reduce the inventory level and let the system run more efficiently. A disassembly line is quite different from an assembly line. For example, not only can the demand arrive at the last station, it can also arrive at any of the other stations in the system. The demand for a component on the disassembly line could fluctuate widely. In fact, there are many other complicating matters that need to be considered to implement the concept of kanbans in such an environment. In this paper, we discuss the complications that are unique to a disassembly line. We discuss the complications in utilizing the conventional production control mechanisms in a disassembly line setting. We then show how to overcome them by implementing kanbans in a disassembly line setting with demand fluctuation and introduce the concept of multi-kanban mechanism. We demonstrate its effectiveness using a simulation model. An example is presented to illustrate the concept.

Workshop on instrumentation of the disassembled BER II

International Nuclear Information System (INIS)

Jauch, W.; Steiner, M.

1982-01-01

A workshop on the instrumentation of a disassembled BER II-reactor took place in the Hahn-Meitner-Institute in Berlin on April 19 and 20, 1982. Invited were all groups that are promoted by the associations 'Neutron Scattering for Investigation of Condensed Substance' and 'Neutron Scattering and Complementary Methods in Chemistry and Biology', along with experts for neutron spectrometers. 40 foreign scientists from 22 different institutes had accepted the invitation. The actual questions were treated in 13 presentations and a certain number of posters, with the latter also comprising activation analysis. The present report contains the presentations submitted, the final discussion minutes and a summary from HMI-view. (orig./RW) [de

Centralized disassembly and packaging of spent fuel in the DOE spent fuel management system

International Nuclear Information System (INIS)

Johnson, E.R.

1986-01-01

In October 1984, E.R. Johnson Associates, Inc. (JAI) initiated a study of the prospective use of a centralized facility for the disassembly and packaging of spent fuel to support the various elements of the US Dept. of Energy (DOE) spent fuel management system, including facilities for monitored retrievable storage (MRS) and repositories. It was DOE's original plan to receive spent fuel at each repository where it would be disassembled and packaged (overpacked) for disposal purposes. Subsequently, DOE considered the prospective use of MRS of spent fuel as an option for providing safe and reliable management of spent fuel. This study was designed to consider possible advantages of the use of centralized facilities for disassembly and packaging of spent fuel at whose location storage facilities could be added as required. The study was divided into three principal technical tasks that covered: (a) development of requirements and criteria for the central disassembly and packaging facility and associated systems. (2) Development of conceptual designs for the central disassembly and packaging facility and associated systems. (3) Estimation of capital and operating costs involved for all system facilities and determination of life cycle costs for various scenarios of operation - for comparison with the reference system

Disassembly automation automated systems with cognitive abilities

CERN Document Server

Vongbunyong, Supachai

2015-01-01

This book presents a number of aspects to be considered in the development of disassembly automation, including the mechanical system, vision system and intelligent planner. The implementation of cognitive robotics increases the flexibility and degree of autonomy of the disassembly system. Disassembly, as a step in the treatment of end-of-life products, can allow the recovery of embodied value left within disposed products, as well as the appropriate separation of potentially-hazardous components. In the end-of-life treatment industry, disassembly has largely been limited to manual labor, which is expensive in developed countries. Automation is one possible solution for economic feasibility. The target audience primarily comprises researchers and experts in the field, but the book may also be beneficial for graduate students.

Manipulator and materials handling systems for reactor decommissioning -Cooperation between the university and the plant operator

International Nuclear Information System (INIS)

Schreck, G.; Bach, F. W.; Haferkamp, H.

1995-01-01

Nuclear reactor dismantling requires suitable handling systems for tools and disassembled components, as well as qualified and reliable disassembly and cutting techniques. From the angle of radiation protection, remote-controlled handling techniques and underwater techniques are the methods of choice, the latter particularly in continuation of plant operating conditions, and this all the more the more disassembly work proceeds towards the reactor core. With the experience accumulated for 20 years now by the Institut fuer Werkstoffkunde (materials science) of Hannover University by basic research and application-oriented development work in the field of thermal cutting technology, especially plasma arc cutting techniques, as well as development work in the field of remote-controlled materials handling systems, the institute is the cut-out partner for disassembly tasks in reactor decommissioning. (Orig./DG) [de

pH-Sensitive Reversible Programmed Targeting Strategy by the Self-Assembly/Disassembly of Gold Nanoparticles.

Science.gov (United States)

Ma, Jinlong; Hu, Zhenpeng; Wang, Wei; Wang, Xinyu; Wu, Qiang; Yuan, Zhi

2017-05-24

A reversible programmed targeting strategy could achieve high tumor accumulation due to its long blood circulation time and high cellular internalization. Here, targeting ligand-modified poly(ethylene glycol) (PEG-ligand), dibutylamines (Bu), and pyrrolidinamines (Py) were introduced on the surface of gold nanoparticles (Au NPs) for reversible shielding/deshielding of the targeting ligands by pH-responsive self-assembly. Hydrophobic interaction and steric repulsion are the main driving forces for the self-assembly/disassembly of Au NPs. The precise self-assembly (pH ≥ 7.2) and disassembly (pH ≤ 6.8) of Au NPs with different ligands could be achieved by fine-tuning the modifying molar ratio of Bu and Py (R m ), which followed the formula R m = 1/(-0.0013X 2 + 0.0323X + 1), in which X is the logarithm of the partition coefficient of the targeting ligand. The assembled/disassembled behavior of Au NPs at pH 7.2 and 6.8 was confirmed by transmission electron microscopy and dynamic light scattering. Enzyme-linked immunosorbent assays and cellular uptake studies showed that the ligands could be buried inside the assembly and exposed when disassembled. More importantly, this process was reversible, which provides the possibility of prolonging blood circulation by shielding ligands associated with the NPs that were effused from tumor tissue.

Conceptual designs of tokamak reactor and R D

International Nuclear Information System (INIS)

Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

1983-01-01

The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

Use of Cementitious Materials for SRS Reactor Facility In-Situ Decommissioning

International Nuclear Information System (INIS)

Langton, C.A.; Stefanko, D.B.; Serrato, M.G.; Blankenship, J.K.; Griffin, W.G.; Long, J.T.

2013-01-01

properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2011. The focus of this paper is to describe the 1) grout mixes for filling the reactor vessels, and 2) a specialty grout mix to fill a selected portion of the P-Reactor Disassembly Basin, and 3) materials used to fill the below-grade facilities. Material property test results, placement strategies, full-scale production and delivery systems will also be described. (author)

Modelling disassembled fuel bundles using CATHENA MOD-3.5a under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Lei, Q M; Sanderson, D B; Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

CATHENA MOD-3.5a is a multipurpose thermalhydraulic computer code developed primarily to analyse postulated loss-of-coolant scenarios for CANDU nuclear reactors. The code contains a generalized heat transfer package that enables it to model the behaviour of a fuel channel in great detail. Throughout the development of the CATHENA code, considerable effort has been devoted to evaluating, validating and documenting its overall capability as a design and safety assessment tool. Specific attention has focused on its ability to predict fuel channel behaviour under postulated accident conditions. This paper describes an investigation of CATHENA`s ability to predict the thermal-chemical responses of a fuel channel in which the 37-element bundles were assumed to disassemble and rearrange into a closed-packed stack of elements at the bottom of the pressure tube. A representative disassembled bundle geometry was modelled during a simulated loss-of-coolant accident scenario using CATHENA MOD-3.5a/Rev 0, with superheated steam being the only coolant available. Thermal conduction in the radial and circumferential directions was calculated for individual fuel elements, the pressure tube, and the calandria tube. Radiation view factors for the intact and disassembled bundle geometries were calculated using a CATHENA utility program. Inter-element metal-to-metal contact was accounted for using the CATHENA solid-solid contact model. An offset pressure-tube configuration, representing a partially sagged pressure tube, and the effect of steam starvation on the exothermic zirconium-steam reaction, were included in the CATHENA model. The CATHENA-predicted results show a dramatic suppression of heat generation from the zirconium-steam reaction when bundle disassembly is initiated. The predicted results show a smaller temperature increase in the fuel sheaths and the pressure tube for the disassembled bundle geometry, compared to the temperature excursion for the intact bundle. (author

Recycling of electrical motors by automatic disassembly

Science.gov (United States)

Karlsson, Björn; Järrhed, Jan-Ove

2000-04-01

This paper presents a robotized workstation for end-of-life treatment of electrical motors with an electrical effect of about 1 kW. These motors can, for example, be found in washing machines and in industry. There are two main steps in the work. The first step is an inspection whereby the functionality of the motor is checked and classification either for re-use or for disassembly is done. In the second step the motors classified for disassembly are disassembled in a robotized automatic station. In the initial step measurements are performed during a start-up sequence of about 1 s. By measuring the rotation speed and the current and voltage of the three phases of the motor classification for either reuse or disassembly can be done. During the disassembly work, vision data are fused in order to classify the motors according to their type. The vision system also feeds the control system of the robot with various object co-ordinates, to facilitate correct operation of the robot. Finally, tests with a vision system and eddy-current equipment are performed to decide whether all copper has been removed from the stator.

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

Energy Technology Data Exchange (ETDEWEB)

Bergren, Christopher L.; Long, J. Tony; Blankenship, John K. [Savannah River Nuclear Solutions, LLC, Bldg. 730-4B, Aiken, SC 29808 (United States); Adams, Karen M. [United States Department of Energy, Bldg. 730-B, Aiken, SC 29808 (United States)

2013-07-01

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

International Nuclear Information System (INIS)

Bergren, Christopher L.; Long, J. Tony; Blankenship, John K.; Adams, Karen M.

2013-01-01

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal

Illustrating the disassembly of 3D models

KAUST Repository

Guo, Jianwei; Yan, Dongming; Li, Er; Dong, Weiming; Wonka, Peter; Zhang, Xiaopeng

2013-01-01

We present a framework for the automatic disassembly of 3D man-made models and the illustration of the disassembly process. Given an assembled 3D model, we first analyze the individual parts using sharp edge loops and extract the contact faces

Disassembly and removal of sodium instrumentation test loop

International Nuclear Information System (INIS)

Ishikawa, Okinobu; Onojima, Takamitu; Nagai, Keiichi

2000-07-01

In 1999, the Sodium Instrumentation Test Loop was disassembled and removed. This report describes the tasks and experiences obtained in removing sodium from a storage tank, disassembling, and cleansing components and related activities. Overall the disassembly, handling and cleansing tasks proceeded as planned and the activities were carried out efficiently and safely. Documentation of the process is meant to establish not only a procedure, but also a guideline for future similar tasks. (author)

Multi-kanban mechanism for appliance disassembly

Science.gov (United States)

Udomsawat, Gun; Gupta, Surendra M.

2005-11-01

The use of household appliances continues to rise every year. A significant number of End-Of-Life (EOL) appliances are generated because of the introduction of newer models that are more attractive, efficient and affordable. Others are, of course, generated when they become non-functional. Many regulations encourage recycling of EOL appliances to reduce the amount of waste sent to landfills. In addition, EOL appliances offer the appliance manufacturing and remanufacturing industries a source of less expensive raw materials and components. For this reason product recovery has become a subject of interest during the past decade. In this paper, we study the disassembly line for appliance disassembly. We discuss and incorporate some of the complications that are inherent in disassembly line including product arrival, demand arrival, inventory fluctuation and production control mechanisms. We show how to overcome such complications by implementing a multi-kanban system in the appliance disassembly line setting. The multi-kanban system (MKS) relies on dynamic routing of kanbans according to the state of the system. We investigate the multi-kanban mechanism using simulation and explore the effect of product mix on performance of the traditional push system (TPS) and MKS in terms of controlling the system's inventory while attempting to achieve a decent customer service level.

Methods for otpimum and near optimum disassembly sequencing

NARCIS (Netherlands)

Lambert, A.J.D.; Gupta, S.M.

2008-01-01

This paper considers disassembly sequencing problems subjected to sequence dependent disassembly costs. In practice, the methods for dealing with such problems rely mainly on metaheuristic and heuristic methods, which intrinsically generate suboptimum solutions. Exact methods are NP-hard and

The Swedish Zero Power Reactor R0

Energy Technology Data Exchange (ETDEWEB)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-15

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

The Swedish Zero Power Reactor R0

International Nuclear Information System (INIS)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-01

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

An Intelligent Agent-Controlled and Robot-Based Disassembly Assistant

Science.gov (United States)

Jungbluth, Jan; Gerke, Wolfgang; Plapper, Peter

2017-09-01

One key for successful and fluent human-robot-collaboration in disassembly processes is equipping the robot system with higher autonomy and intelligence. In this paper, we present an informed software agent that controls the robot behavior to form an intelligent robot assistant for disassembly purposes. While the disassembly process first depends on the product structure, we inform the agent using a generic approach through product models. The product model is then transformed to a directed graph and used to build, share and define a coarse disassembly plan. To refine the workflow, we formulate “the problem of loosening a connection and the distribution of the work” as a search problem. The created detailed plan consists of a sequence of actions that are used to call, parametrize and execute robot programs for the fulfillment of the assistance. The aim of this research is to equip robot systems with knowledge and skills to allow them to be autonomous in the performance of their assistance to finally improve the ergonomics of disassembly workstations.

Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

Energy Technology Data Exchange (ETDEWEB)

Weber, J.W.

1994-09-28

The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

Corrosion surveillance for research reactor spent nuclear fuel in wet basin storage

International Nuclear Information System (INIS)

Howell, J.P.

1999-01-01

Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the ''Atoms for Peace'' program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on ''Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water'' and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the Receiving Basin for Offsite Fuels (RBOF) at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993

A symbolic methodology to improve disassembly process design.

Science.gov (United States)

Rios, Pedro; Blyler, Leslie; Tieman, Lisa; Stuart, Julie Ann; Grant, Ed

2003-12-01

Millions of end-of-life electronic components are retired annually due to the proliferation of new models and their rapid obsolescence. The recovery of resources such as plastics from these goods requires their disassembly. The time required for each disassembly and its associated cost is defined by the operator's familiarity with the product design and its complexity. Since model proliferation serves to complicate an operator's learning curve, it is worthwhile to investigate the benefits to be gained in a disassembly operator's preplanning process. Effective disassembly process design demands the application of green engineering principles, such as those developed by Anastas and Zimmerman (Environ. Sci. Technol. 2003, 37, 94A-101A), which include regard for product complexity, structural commonality, separation energy, material value, and waste prevention. This paper introduces the concept of design symbolsto help the operator more efficiently survey product complexity with respect to location and number of fasteners to remove a structure that is common to all electronics: the housing. With a sample of 71 different computers, printers, and monitors, we demonstrate that appropriate symbols reduce the total disassembly planning time by 13.2 min. Such an improvement could well make efficient the separation of plastic that would otherwise be destined for waste-to-energy or landfill. The symbolic methodology presented may also improve Design for Recycling and Design for Maintenance and Support.

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

Present status of design, research and development of nuclear fusion reactors and problems

International Nuclear Information System (INIS)

1983-04-01

Seven years have elapsed since the publication of ''Progress of nuclear fusion research and perspective toward the development of power reactors'' by the Atomic Energy Society of Japan in August, 1976. During this period, the research and development of nuclear fusion have changed from plasma physics to reactor technology, being conscious of the realization of fusion reactors. There are the R project in the Institute of Plasma Physics, Nagoya University, and the design and construction of JT-60 in Japan Atomic Energy Research Institute, to put it concretely. Now the research and development taking the economical efficiency into account are adopted. However, the type of fusion reactors is not reduced to tokamak type, accordingly the research and development to meet the diverse possibilities are forwarded. The progress of tokamak reactor research, core plasma design, nuclear design and shielding design, thermal structure design, the design of superconducting magnets, disassembling and repair, safety, economical efficiency, the conceptual design of other types than tokamak and others are reported. (Kako, I.)

Illustrating the disassembly of 3D models

KAUST Repository

Guo, Jianwei

2013-06-11

We present a framework for the automatic disassembly of 3D man-made models and the illustration of the disassembly process. Given an assembled 3D model, we first analyze the individual parts using sharp edge loops and extract the contact faces between each pair of neighboring parts. The contact faces are then used to compute the possible moving directions of each part. We then present a simple algorithm for clustering the sets of the individual parts into meaningful sub-assemblies, which can be used for a hierarchical decomposition. We take the stability of sub-assemblies into account during the decomposition process by considering the upright orientation of the input models. Our framework also provides a user-friendly interface to enable the superimposition of the constraints for the decomposition. Finally, we visualize the disassembly process by generating an animated sequence. The experiments demonstrate that our framework works well for a variety of complex models. © 2013 Elsevier Ltd.

Coordination of substrate binding and ATP hydrolysis in Vps4-mediated ESCRT-III disassembly.

Science.gov (United States)

Davies, Brian A; Azmi, Ishara F; Payne, Johanna; Shestakova, Anna; Horazdovsky, Bruce F; Babst, Markus; Katzmann, David J

2010-10-01

ESCRT-III undergoes dynamic assembly and disassembly to facilitate membrane exvagination processes including multivesicular body (MVB) formation, enveloped virus budding, and membrane abscission during cytokinesis. The AAA-ATPase Vps4 is required for ESCRT-III disassembly, however the coordination of Vps4 ATP hydrolysis with ESCRT-III binding and disassembly is not understood. Vps4 ATP hydrolysis has been proposed to execute ESCRT-III disassembly as either a stable oligomer or an unstable oligomer whose dissociation drives ESCRT-III disassembly. An in vitro ESCRT-III disassembly assay was developed to analyze Vps4 function during this process. The studies presented here support a model in which Vps4 acts as a stable oligomer during ATP hydrolysis and ESCRT-III disassembly. Moreover, Vps4 oligomer binding to ESCRT-III induces coordination of ATP hydrolysis at the level of individual Vps4 subunits. These results suggest that Vps4 functions as a stable oligomer that acts upon individual ESCRT-III subunits to facilitate ESCRT-III disassembly.

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Mathematical and Simulation Modelling of Moisture Diffusion Mechanism during Plastic IC Packages Disassembly

OpenAIRE

Peng Mou; Dong Xiang; Guanghong Duan

2013-01-01

Reuse of plastic IC packages disassembled from printed circuit boards (PCBs) has significant environmental benefits and economic value. The interface delamination caused by moisture diffusion is the main failure mode of IC packages during the disassembling process, which greatly reduces the reusability and reliability of disassembled IC packages. Exploring moisture diffusion mechanism is a prerequisite to optimize prebaking processes before disassembling that is an effective way to avoid the ...

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Movement and equipositioning of plasmids by ParA filament disassembly

DEFF Research Database (Denmark)

Ringgaard, Simon; van Zon, Jeroen; Howard, Martin

2009-01-01

, plasmids consistently migrate behind disassembling ParA cytoskeletal structures, suggesting that ParA filaments pull plasmids by depolymerization. The perpetual cycles of ParA assembly and disassembly result in continuous relocation of plasmids, which, on time averaging, results in equidistribution...

Desmosome Assembly and Disassembly Are Membrane Raft-Dependent

Science.gov (United States)

Faundez, Victor; Koval, Michael; Mattheyses, Alexa L.; Kowalczyk, Andrew P.

2014-01-01

Strong intercellular adhesion is critical for tissues that experience mechanical stress, such as the skin and heart. Desmosomes provide adhesive strength to tissues by anchoring desmosomal cadherins of neighboring cells to the intermediate filament cytoskeleton. Alterations in assembly and disassembly compromise desmosome function and may contribute to human diseases, such as the autoimmune skin blistering disease pemphigus vulgaris (PV). We previously demonstrated that PV auto-antibodies directed against the desmosomal cadherin desmoglein 3 (Dsg3) cause loss of adhesion by triggering membrane raft-mediated Dsg3 endocytosis. We hypothesized that raft membrane microdomains play a broader role in desmosome homeostasis by regulating the dynamics of desmosome assembly and disassembly. In human keratinocytes, Dsg3 is raft associated as determined by biochemical and super resolution immunofluorescence microscopy methods. Cholesterol depletion, which disrupts rafts, prevented desmosome assembly and adhesion, thus functionally linking rafts to desmosome formation. Interestingly, Dsg3 did not associate with rafts in cells lacking desmosomal proteins. Additionally, PV IgG-induced desmosome disassembly occurred by redistribution of Dsg3 into raft-containing endocytic membrane domains, resulting in cholesterol-dependent loss of adhesion. These findings demonstrate that membrane rafts are required for desmosome assembly and disassembly dynamics, suggesting therapeutic potential for raft targeting agents in desmosomal diseases such as PV. PMID:24498201

Multi-kanban mechanism for personal computer disassembly

Science.gov (United States)

Udomsawat, Gun; Gupta, Surendra M.; Kamarthi, Sagar V.

2004-12-01

The use of personal computers (PCs) continues to increase every year. According to a 1999 figure, 50 percent of all US households owned PCs, a figure that continues to rise every year. With continuous development of sophisticated software, PCs are becoming increasingly powerful. In addition, the price of a PC continues to steadily decline. Furthermore, the typical life of a PC in the workplace is approximately two to three years while in the home it is three to five years. As these PCs become obsolete, they are replaced and the old PCs are disposed of. It is estimated that between 14 and 20 million PCs are retired annually in the US. While 20 to 30% of the units may be resold, the others are discarded. These discards represent a significant potential source of lead for the waste stream. In some communities, waste cathode ray tubes (CRTs) represent the second largest source of lead in the waste stream after vehicular lead acid batteries. PCs are, therefore, not suitable for dumping in landfills. Besides, several components of a PC can be reused and then there are other valuable materials that can also be harvested. And with the advent of product stewardship, product recovery is the best solution for manufacturers. Disassembly line is perhaps the most suitable set up for disassembling PCs. However, planning and scheduling of disassembly on a disassembly line is complicated. In this paper, we discuss some of the complications including product arrival, demand arrival, inventory fluctuation and production control mechanisms. We then show how to overcome them by implementing a multi-kanban mechanism in the PC disassembly line setting. The multi-kanban mechanism relies on dynamic routing of kanbans according to the state of the system. We investigate the multi-kanban mechanism using simulation and demonstrate that this mechanism is superior to the traditional push system in terms of controlling the system"s inventory while maintaining a decent customer service level.

Maintenance features of the Compact Ignition Tokamak fusion reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Hager, E.R.

1987-01-01

The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

KADIS: a program to analyse the disassembly phase of hypothetical accidents in LMFBRs

International Nuclear Information System (INIS)

Schmuck, P.; Jacobs, G.; Arnecke, G.

1977-11-01

The program KADIS models the disassembly phase during power excursions in LMFBR hypothetical accidents. KADIS is based on point kinetics in the neutronics part and on a 2-dimensional representation of the reactor core in the hydrodynamics part. The core is modeled as an ideal, compressible fluid which is heated up adiabatically during the excursion. KADIS was built up with the help of the VENUS program of Argonne National Laboratory. Several important features were added to the basic VENUS model. Therefore we give first a complete description of the mathematical models used. Secondly we provide the user with the necessary information to handle the input/output of KADIS. (orig.) [de

Lead-based Fast Reactor Development Plan and R&D Status in China

International Nuclear Information System (INIS)

Wu Yican

2013-01-01

• Lead-based fast reactors have good potential for waste transmutation, fuel breeding and energy production, which has been selected by CAS as the advanced reactor development emphasis with the support of ADS program and MFE program. Sharing of technologies R&D is possible among GIF/ADS/Fusion. • The concepts and test strategy of series China lead-based fast reactors (CLEAR) have been developed. The preliminary engineering design and safety analysis of CLEAR-I are underway. • Technology R&D on CLEAR with series lead alloy loops and accelerator-based neutron generator have been constructed or under construction. • CLEAR series reactor design and construction have big challenges, widely international cooperation on reactor design and technology R&D is welcome

Accomplishment of JT-60U disassembly work dealing with radioactive components

International Nuclear Information System (INIS)

Ikeda, Yoshitaka

2015-01-01

The upgrade of the JT-60U to the superconducting tokamak 'JT-60SA' has been carried out to contribute the early realization of fusion energy by addressing key physics issues relevant for ITER and DEMO. Disassembly of the JT-60U tokamak was required so as to newly install the JT-60SA torus at the same position in the torus hall. The JT-60U tokamak was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium-deuterium (D-D) reactions of 1.5x10"2"0 (n) in total. Since this work was the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. About 13,000 components with a total weight of more than 5,400 tonnes were removed from the torus hall and stored safely in storage facilities. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. It was confirmed that the main radioactive material of the disassembly components was the stainless steel and that its dose rate was almost background level (∼0.1 μSv/h) at ∼10 m far from the vacuum vessel. It seems that the disassembly components with background dose level are in the clearance level. The assembly of JT-60SA tokamak has started in January 2013 after this disassembly of the JT-60U tokamak. (author)

1982 annual status report: reactor safety

International Nuclear Information System (INIS)

1982-01-01

This report presents the projects of the Reactor Safety Program at the JRC: 1) Reliability and risk evolution; 2) LWR loss of coolant accident studies; 3) Primary system integrity; 4) LMFBR core accident initiation and transition phase; and, 5) LMFBR accident post disassembly phase

Disassembly for remanufacturing: A systematic literature review, new model development and future research needs

Directory of Open Access Journals (Sweden)

Anjar Priyono

2016-11-01

Full Text Available Purpose: Disassembly is an important process that distinguishes remanufacturing from conventional manufacturing. It is a unique process that becomes focus of investigation from many scholars. Yet, most scholars investigate disassembly from technical and operational standpoint that lack of strategic perspective. This paper attempts to fill this gap by looking at disassembly from a strategic perspective by considering organisational characteristics, process choices and product attributes. To be more specific, this paper has three objectives. First, to gain understanding what has been done, and what need to be done in the field of disassembly in remanufacturing. Second, to conduct a systematic literature review for identifying the factors affecting disassembly for remanufacturing. Third, to propose a new model of disassembly for remanufacturing and also to provide avenues for future research. Design/methodology/approach: This study used a systematic literature review method. A series of steps were undertaken during the review. The study was started with determining the purpose of the study, selecting appropriate keywords, and reducing the selected papers using a number of criteria. A deeper analysis was carried out on the final paper that meets the criteria for this review. Findings: There are two main findings of this study. First, a list of factors affecting disassembly in remanufacturing is identified. The factors can be categorised into three groups: organisational factors, process choices and product attributes. Second, using factors that have been identified, a new model of disassembly process for remanufacturing is developed. Current studies only consider disassembly as a physical activity to break down products into components. In the new model, disassembly is viewed as a process that converts into into output, which consist of a series of steps. Research limitations/implications: The opportunities for future research include: the need to

Disassembly for remanufacturing: A systematic literature review, new model development and future research needs

International Nuclear Information System (INIS)

Priyono, A.; Ijomah, W.; Bititci, U.

2016-01-01

Purpose: Disassembly is an important process that distinguishes remanufacturing from conventional manufacturing. It is a unique process that becomes focus of investigation from many scholars. Yet, most scholars investigate disassembly from technical and operational standpoint that lack of strategic perspective. This paper attempts to fill this gap by looking at disassembly from a strategic perspective by considering organisational characteristics, process choices and product attributes. To be more specific, this paper has three objectives. First, to gain understanding what has been done, and what need to be done in the field of disassembly in remanufacturing. Second, to conduct a systematic literature review for identifying the factors affecting disassembly for remanufacturing. Third, to propose a new model of disassembly for remanufacturing and also to provide avenues for future research. Design/methodology/approach: This study used a systematic literature review method. A series of steps were undertaken during the review. The study was started with determining the purpose of the study, selecting appropriate keywords, and reducing the selected papers using a number of criteria. A deeper analysis was carried out on the final paper that meets the criteria for this review. Findings: There are two main findings of this study. First, a list of factors affecting disassembly in remanufacturing is identified. The factors can be categorised into three groups: organisational factors, process choices and product attributes. Second, using factors that have been identified, a new model of disassembly process for remanufacturing is developed. Current studies only consider disassembly as a physical activity to break down products into components. In the new model, disassembly is viewed as a process that converts into into output, which consist of a series of steps. Research limitations/implications: The opportunities for future research include: the need to develop an index of

Disassembly for remanufacturing: A systematic literature review, new model development and future research needs

Energy Technology Data Exchange (ETDEWEB)

Priyono, A.; Ijomah, W.; Bititci, U.

2016-07-01

Purpose: Disassembly is an important process that distinguishes remanufacturing from conventional manufacturing. It is a unique process that becomes focus of investigation from many scholars. Yet, most scholars investigate disassembly from technical and operational standpoint that lack of strategic perspective. This paper attempts to fill this gap by looking at disassembly from a strategic perspective by considering organisational characteristics, process choices and product attributes. To be more specific, this paper has three objectives. First, to gain understanding what has been done, and what need to be done in the field of disassembly in remanufacturing. Second, to conduct a systematic literature review for identifying the factors affecting disassembly for remanufacturing. Third, to propose a new model of disassembly for remanufacturing and also to provide avenues for future research. Design/methodology/approach: This study used a systematic literature review method. A series of steps were undertaken during the review. The study was started with determining the purpose of the study, selecting appropriate keywords, and reducing the selected papers using a number of criteria. A deeper analysis was carried out on the final paper that meets the criteria for this review. Findings: There are two main findings of this study. First, a list of factors affecting disassembly in remanufacturing is identified. The factors can be categorised into three groups: organisational factors, process choices and product attributes. Second, using factors that have been identified, a new model of disassembly process for remanufacturing is developed. Current studies only consider disassembly as a physical activity to break down products into components. In the new model, disassembly is viewed as a process that converts into into output, which consist of a series of steps. Research limitations/implications: The opportunities for future research include: the need to develop an index of

Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

International Nuclear Information System (INIS)

1995-07-01

The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

International Nuclear Information System (INIS)

Arien, B.; Daniels, J.

1986-12-01

CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

R and D directions for the development of CANDU reactors

International Nuclear Information System (INIS)

Torgerson, D.F.

1998-01-01

Full text: AECL is carrying out a comprehensive R and D programme to advance all aspects of CANDU reactor technology. These programs are focusing on three main strategic directions: improved economics, enhanced safety, and fuel cycle flexibility. R and D areas include fuel cycle development, heavy water technology, fuel channel development, safety technology, control and instrumentation, reactor chemistry, systems and components, and health and environment. In each case, the R and D programs have short, medium, and long-term goals to achieve the overall strategic directions. Most of the programs seek to further develop and exploit some of the unique characteristics of pressurized heavy water reactors. Examples of this include high neutron economy and on-power fueling which allow several different fuel cycles, the presence of large water heat sinks for enhanced safety, and modular components that can be easily replaced for plant life extension. This presentation reviews AECL's product development directions and the R and D programs that have been begun for their development

An Entropy-Based Adaptive Hybrid Particle Swarm Optimization for Disassembly Line Balancing Problems

Directory of Open Access Journals (Sweden)

Shanli Xiao

2017-11-01

Full Text Available In order to improve the product disassembly efficiency, the disassembly line balancing problem (DLBP is transformed into a problem of searching for the optimum path in the directed and weighted graph by constructing the disassembly hierarchy information graph (DHIG. Then, combining the characteristic of the disassembly sequence, an entropy-based adaptive hybrid particle swarm optimization algorithm (AHPSO is presented. In this algorithm, entropy is introduced to measure the changing tendency of population diversity, and the dimension learning, crossover and mutation operator are used to increase the probability of producing feasible disassembly solutions (FDS. Performance of the proposed methodology is tested on the primary problem instances available in the literature, and the results are compared with other evolutionary algorithms. The results show that the proposed algorithm is efficient to solve the complex DLBP.

A methodological approach for designing and sequencing product families in Reconfigurable Disassembly Systems

Directory of Open Access Journals (Sweden)

Ignacio Eguia

2011-10-01

Full Text Available Purpose: A Reconfigurable Disassembly System (RDS represents a new paradigm of automated disassembly system that uses reconfigurable manufacturing technology for fast adaptation to changes in the quantity and mix of products to disassemble. This paper deals with a methodology for designing and sequencing product families in RDS. Design/methodology/approach: The methodology is developed in a two-phase approach, where products are first grouped into families and then families are sequenced through the RDS, computing the required machines and modules configuration for each family. Products are grouped into families based on their common features using a Hierarchical Clustering Algorithm. The optimal sequence of the product families is calculated using a Mixed-Integer Linear Programming model minimizing reconfigurability and operational costs. Findings: This paper is focused to enable reconfigurable manufacturing technologies to attain some degree of adaptability during disassembly automation design using modular machine tools. Research limitations/implications: The MILP model proposed for the second phase is similar to the well-known Travelling Salesman Problem (TSP and therefore its complexity grows exponentially with the number of products to disassemble. In real-world problems, which a higher number of products, it may be advisable to solve the model approximately with heuristics. Practical implications: The importance of industrial recycling and remanufacturing is growing due to increasing environmental and economic pressures. Disassembly is an important part of remanufacturing systems for reuse and recycling purposes. Automatic disassembly techniques have a growing number of applications in the area of electronics, aerospace, construction and industrial equipment. In this paper, a design and scheduling approach is proposed to apply in this area. Originality/value: This paper presents a new concept called Reconfigurable Disassembly System

The effect of sudden server breakdown on the performance of a disassembly line

Science.gov (United States)

Udomsawat, Gun; Gupta, Surendra M.

2005-11-01

Product and material recovery relies on the disassembly process to separate target components or materials from the end-of-life (EOL) products. Disassembly line is especially effective when products in large quantity are disassembled. Unlike an assembly line, a disassembly line is more complex and is subjected to numerous uncertainties including stochastic and multi-level arrivals of component demands, stochastic arrival times for EOL products, and process interruption due to equipment failure. These factors seriously impair the control mechanism in the disassembly line. A common production control mechanism is the traditional push system (TPS). TPS responds to the aforementioned complications by carrying substantial amounts of inventories. An alternative control mechanism is a newly developed multi-kanban pull system (MKS) that relies on dynamic routing of kanbans, which tends to minimize the system's inventories while maintaining demand serviceability. In this paper we explore the impact of sudden breakdown of server on the performance of a disassembly line. We compare the overall performances of the TPS and MKS by considering two scenarios. We present the solution procedure and results for these cases.

Design for Disassembly

DEFF Research Database (Denmark)

Beim, Anne

2018-01-01

‘Design for Disassembly’ is becoming a familiar part of discourses on environmental strategies for construction and sustainable design. The concept originates in industrial design and pertains to a wide range of strategies for reuse and recycling, all of which highlight a whole-of-life approach t...... to materials and buildings. Design for Disassembly is particularly relevant to prefabricated and industrialized construction; as such methodologies already contain within them the possibility for easy deconstruction, whether for maintenance, refurbishment or reuse....

Application of industrial robots in automatic disassembly line of waste LCD displays

Science.gov (United States)

Wang, Sujuan

2017-11-01

In the automatic disassembly line of waste LCD displays, LCD displays are disassembled into plastic shells, metal shields, circuit boards, and LCD panels. Two industrial robots are used to cut metal shields and remove circuit boards in this automatic disassembly line. The functions of these two industrial robots, and the solutions to the critical issues of model selection, the interfaces with PLCs and the workflows were described in detail in this paper.

Low cytoplasmic pH reduces ER-Golgi trafficking and induces disassembly of the Golgi apparatus

Energy Technology Data Exchange (ETDEWEB)

Soonthornsit, Jeerawat [Laboratory for Cell and Developmental Biology, Department of Molecular Biosciences, Faculty of Life Sciences, Kyoto Sangyo University, Motoyama, Kamigamo, Kita, Kyoto 603-8555 (Japan); Yamaguchi, Yoko; Tamura, Daisuke [Division of Life Sciences, Graduate School of Natural Science and Technology, Kanazawa University, Kakuma, Kanazawa 920-1192 (Japan); Ishida, Ryuichi; Nakakoji, Yoko; Osako, Shiho [Laboratory for Cell and Developmental Biology, Department of Molecular Biosciences, Faculty of Life Sciences, Kyoto Sangyo University, Motoyama, Kamigamo, Kita, Kyoto 603-8555 (Japan); Yamamoto, Akitsugu [Department of Animal Bioscience, Nagahama Institute of Bio-Science and Technology, 266 Tamura, Nagahama, Shiga, 526‐0829 (Japan); Nakamura, Nobuhiro, E-mail: osaru3@cc.kyoto-su.ac.jp [Laboratory for Cell and Developmental Biology, Department of Molecular Biosciences, Faculty of Life Sciences, Kyoto Sangyo University, Motoyama, Kamigamo, Kita, Kyoto 603-8555 (Japan); Division of Life Sciences, Graduate School of Natural Science and Technology, Kanazawa University, Kakuma, Kanazawa 920-1192 (Japan)

2014-11-01

The Golgi apparatus was dramatically disassembled when cells were incubated in a low pH medium. The cis-Golgi disassembled quickly, extended tubules and spread to the periphery of cells within 30 min. In contrast, medial- and trans-Golgi were fragmented in significantly larger structures of smaller numbers at a slower rate and remained largely in structures distinct from the cis-Golgi. Electron microscopy revealed the complete disassembly of the Golgi stack in low pH treated cells. The effect of low pH was reversible; the Golgi apparatus reassembled to form a normal ribbon-like structure within 1–2 h after the addition of a control medium. The anterograde ER to Golgi transport and retrograde Golgi to ER transport were both reduced under low pH. Phospholipase A{sub 2} inhibitors (ONO, BEL) effectively suppressed the Golgi disassembly, suggesting that the phospholipase A{sub 2} was involved in the Golgi disassembly. Over-expression of Rab1, 2, 30, 33 and 41 also suppressed the Golgi disassembly under low pH, suggesting that they have protective role against Golgi disassembly. Low pH treatment reduced cytoplasmic pH, but not the luminal pH of the Golgi apparatus, strongly suggesting that reduction of the cytoplasmic pH triggered the Golgi disassembly. Because a lower cytoplasmic pH is induced in physiological or pathological conditions, disassembly of the Golgi apparatus and reduction of vesicular transport through the Golgi apparatus may play important roles in cell physiology and pathology. Furthermore, our findings indicated that low pH treatment can serve as an important tool to analyze the molecular mechanisms that support the structure and function of the Golgi apparatus. - Highlights: • The Golgi apparatus reversibly disassembles by low pH treatment. • The cis-Golgi disassembles quickly generating tubular structures. • Both anterograde and retrograde transport between the ER and the Golgi apparatus are reduced. • Phospholipase A{sub 2} inhibitors (ONO

105-KE basin pilot run relocation

International Nuclear Information System (INIS)

Crystal, J.B.

1994-01-01

The purpose of this document is to present the bases for selecting the exact in-facility location for installation of process equipment to support pilot testing activities in the 105-KE Basin at the United States Department of Energy Hanford Site, in southeastern Washington State. The 105-KE Basin was constructed during the early 1950s, as an integralcomponent of the 105-K East reactor building. Similar basins were provided in all Hanford weapons production reactor buildings to receive fuel elements discharged from the reactors and stage them for rail transport to 200 Area fuel reprocessing plants. The 105-KE reactor began operation in 1955. It was shut down in 1971. However, the 105-KE Basin was reactivated several years later to store spent fuel from the N-Reactor basin and permit its continued operation during outages at the Plutonium Uranium Extraction (PUREX) plant in the 200E Area

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

The molten salt reactor: R and D status and perspectives in Europe

International Nuclear Information System (INIS)

Renault, Claude; Delpech, Sylvie; Merle-Lucotte, Elsa; Konings, Rudy; Hron, Miloslav; Ignatiev, Victor

2010-01-01

The paper concentrates on molten salt fast reactor (MSFR) concepts which are receiving most attention in the EU context. It shows the main R and D achievements and some remaining issues to be addressed in such essential areas as (a) reactor conceptual design, (b) molten salt properties, (c) fuel salt clean-up scheme and (d) high temperature materials. The status and perspectives of molten salt reactor R and D efforts in Europe are then discussed

Disassembly Properties of Cementitious Finish Joints Using an Induction Heating Method

Science.gov (United States)

Ahn, Jaecheol; Noguchi, Takafumi; Kitagaki, Ryoma

2015-01-01

Efficient maintenance and upgrading of a building during its lifecycle are difficult because a cementitious finish uses materials and parts with low disassembly properties. Additionally, the reuse and recycling processes during building demolition also present numerous problems from the perspective of environmental technology. In this study, an induction heating (IH) method was used to disassemble cementitious finish joints, which are widely used to join building members and materials. The IH rapidly and selectively heated and weakened these joints. The temperature elevation characteristics of the cementitious joint materials were measured as a function of several resistor types, including wire meshes and punching metals, which are usually used for cementitious finishing. The disassembly properties were evaluated through various tests using conductive resistors in cementitious joints such as mortar. When steel fiber, punching metal, and wire mesh were used as conductive resistors, the cementitious modifiers could be weakened within 30 s. Cementitious joints with conductive resistors also showed complete disassembly with little residual bond strength.

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

Energy Technology Data Exchange (ETDEWEB)

Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

2003-07-01

By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

SP-100 reactor disassembly remote handling test program

International Nuclear Information System (INIS)

Wilson, C.E.; Potter, J.D.; Maiden, G.E.; Vader, D.P.

1991-01-01

This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities

Characterization of radioactive graphite and concrete of the reactor ULYSSE/INSTN at CEA/Saclay to be dismantled

International Nuclear Information System (INIS)

Van Lauwe, Aymeric; Ridikas, Danas; Damoy, Francois; Blideanu, Valentin; Fajardo, Christophe; Aubert, Marie-Cecile; Foulon, Francois

2006-01-01

Decommissioning and dismantling of nuclear installations after their service life are connected with the necessity of the disassembling, handling and disposing of a large amount of radioactive material. In order to optimize the disassembling operations, to reduce the undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the radiological characterisation of the material present in the reactor and around its environment should be accurately evaluated. The present work has been done in the framework of the decommissioning and dismantling of the experimental reactor ULYSSE that is presently operating in INSTN/Saclay and will be closed in the middle of 2006. A methodology, already successfully used for another research reactor, is proposed for determining accurately the long-term induced activity of the materials present in the active reactor core and its surroundings. The comparison of theoretical predictions, based on Monte Carlo technique, with experimental values validated the approach and the methodology used in the present study. The goal is to plan efficiently the disassembling and dismantling of the system and to optimise the mass flow going to different waste repositories. We show that this approach might reduce substantially the total cost of decommissioning. (authors)

Design of first reactor protection system prototype for C A R E M reactor

International Nuclear Information System (INIS)

Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

2006-01-01

In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

Energy Technology Data Exchange (ETDEWEB)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

Spent fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

International Nuclear Information System (INIS)

Townes, G.A.

1979-10-01

Methods of disassembling and canning spent fuel to allow more efficient storage are being investigated at the BNFP. Studies and development programs are aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than full fuel assemblies. Results indicate that doubling existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly could be performed in the BNFP hot cells at rates of about 12 to 15 assemblies per day

R-matrix parameters in reactor applications

International Nuclear Information System (INIS)

Hwang, R.N.

1992-01-01

The key role of the resonance phenomena in reactor applications manifests through the self-shielding effect. The basic issue involves the application of the microscopic cross sections in the macroscopic reactor lattices consisting of many nuclides that exhibit resonance behavior. To preserve the fidelity of such a effect requires the accurate calculations of the cross sections and the neutron flux in great detail. This clearly not possible without viable resonance data. Recently released ENDF/B VI resonance data in the resolved range especially reflect the dramatic improvement in two important areas; namely, the significant extension of the resolved resonance ranges accompanied by the availability of the R-matrix parameters of the Reich-Moore type. Aside from the obvious increase in computing time required for the significantly greater number of resonances, the main concern is the compatibility of the Riech-Moore representation to the existing reactor processing codes which, until now, are based on the traditional cross section formalisms. This purpose of this paper is to summarize our recent efforts to facilitate implementation of the proposed methods into the production codes at ANL

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

Hybrid disassembly system for cellular telephone end-of-life treatment

Energy Technology Data Exchange (ETDEWEB)

Kniebel, M.; Basdere, B.; Seliger, G. [Technical Univ. Berlin, Inst. for Machine Tools and Factory Management, Dept. of Assembly Technology and Factory Management, Berlin (Germany)

2004-07-01

Concern over the negative environmental impacts associated with the production, use, and end-of-life (EOL) of cellular telephones is particularly high due to large production volumes and characteristically short time scales of technological and stylistic obsolescence. Landfilled or incinerated cellular telephones create the potential for release of toxic substances. The European legislation has passed the directive on Waste of Electrical and Electronic Equipment (WEEE) to regulate their collection and appropriate end-of-life treatment. Manufacturers must conduct material recycling or remanufacturing processes to recover resources. While recovery rates can hardly be met economically by material recycling, remanufacturing and reusing cellular phones is developing into a reasonable alternative. Both end-of-life options require disassembly processes for WEEE compliant treatment. Due to the high number of different cell phone variants and their typical design that fits components into tight enclosing spaces, cellular phone disassembly becomes a challenging task. These challenges and the expected high numbers of phones to be returned in the course of the WEEE urges for automated disassembly. A hybrid disassembly system has been developed to ensure the mass-treatment of obsolete cellular phones. It has been integrated into a prototypical remanufacturing factory for cellular phones that has been planned based on market data. (orig.)

Spent LWR fuel storage costs: reracking, AR basins, and AFR basins

International Nuclear Information System (INIS)

1980-01-01

Whenever possible, fuel storage requirements will be met by reracking existing reactor basins and/or transfer of fuel to available space in other reactor basins. These alternatives represent not only the lowest cost storage options but also the most timely. They are recognized to face environmental and regulatory obstacles. However, such obstacles should be less severe than those that would be encountered with AR or AFR basin storage. When storage requirements cannot be met by the first two options, the least costly alternative for most utilities will be use of a Federal AFR. Storage costs of $100,000 to $150,000 MTU at a AFR are less costly than charges of up to $320,000/MTU that could be incurred by the use of AR basins. AFR storage costs do not include transportation from the reactor to the AFR. This cost would be paid by the utility separately. Only when a utility requires annual storage capacity for 100 MTU of spent fuel can self-storage begin to compete with AFR costs. The large reactor complexes discharging these fuel quantities are not currently those that require relief from fuel storage problems

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2002-07-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Low cytoplasmic pH reduces ER-Golgi trafficking and induces disassembly of the Golgi apparatus.

Science.gov (United States)

Soonthornsit, Jeerawat; Yamaguchi, Yoko; Tamura, Daisuke; Ishida, Ryuichi; Nakakoji, Yoko; Osako, Shiho; Yamamoto, Akitsugu; Nakamura, Nobuhiro

2014-11-01

The Golgi apparatus was dramatically disassembled when cells were incubated in a low pH medium. The cis-Golgi disassembled quickly, extended tubules and spread to the periphery of cells within 30 min. In contrast, medial- and trans-Golgi were fragmented in significantly larger structures of smaller numbers at a slower rate and remained largely in structures distinct from the cis-Golgi. Electron microscopy revealed the complete disassembly of the Golgi stack in low pH treated cells. The effect of low pH was reversible; the Golgi apparatus reassembled to form a normal ribbon-like structure within 1-2h after the addition of a control medium. The anterograde ER to Golgi transport and retrograde Golgi to ER transport were both reduced under low pH. Phospholipase A2 inhibitors (ONO, BEL) effectively suppressed the Golgi disassembly, suggesting that the phospholipase A2 was involved in the Golgi disassembly. Over-expression of Rab1, 2, 30, 33 and 41 also suppressed the Golgi disassembly under low pH, suggesting that they have protective role against Golgi disassembly. Low pH treatment reduced cytoplasmic pH, but not the luminal pH of the Golgi apparatus, strongly suggesting that reduction of the cytoplasmic pH triggered the Golgi disassembly. Because a lower cytoplasmic pH is induced in physiological or pathological conditions, disassembly of the Golgi apparatus and reduction of vesicular transport through the Golgi apparatus may play important roles in cell physiology and pathology. Furthermore, our findings indicated that low pH treatment can serve as an important tool to analyze the molecular mechanisms that support the structure and function of the Golgi apparatus. Copyright © 2014 Elsevier Inc. All rights reserved.

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Microcanonical simulation of nuclear disassembly

International Nuclear Information System (INIS)

Koonin, S.E.; Randrup, J.

1986-01-01

There is considerable interest in the disassembly of the hot nuclear matter produced in high-energy nuclear collisions. A particular stimulus has been the prospect of observing a nuclear liquid-gas phase transition. On rather general grounds, such a transition is expected to occur in nuclear matter at subsaturation densities with temperatures of 10-20 MeV. However, virtually all previous discussions of this phenomenon have been based on thermodynamical considerations valid for infinite, non-interacting systems and the qualitative validity of the results has not been ascertained for the relatively small, finite, interacting systems of practical relevance. Nor is it clear how the occurrence of the phase transition will manifest itself in the asymptotically observed fragment distribution. To progress in these matters, the authors have formulated a microcanonical simulation of the disassembly process, including interfragment interactions. It is a natural refinement of the grand canonical model first presented in [1] and further developed in [2] and is also a exact version of the model developed in [3] for the generation of complete multifragment events in medium-energy collisions. In this contribution, the authors give a brief description of the key ingredients in the model and its numerical implementation

Effects of reactive oxygen species on cellular wall disassembly of banana fruit during ripening.

Science.gov (United States)

Cheng, Guiping; Duan, Xuewu; Shi, John; Lu, Wangjin; Luo, Yunbo; Jiang, Weibo; Jiang, Yueming

2008-07-15

Fruit softening is generally attributed to cell wall disassembly. Experiments were conducted to investigate effects of various reactive oxygen species (ROS) on in vitro cellular wall disassembly of harvested banana fruit. The alcohol-extracted insoluble residue (AEIR) was obtained from the pulp tissues of banana fruit at various ripening stages and then used to examine the disassembly of cellular wall polysaccharides in the presence of superoxide anion (O2(-)), hydrogen peroxide (H2O2) or hydroxyl radical (OH) and their scavengers. The presence of OH accelerated significantly disassembly of cellular wall polysaccharides in terms of the increase in contents of total sugars released and uronic acid, and the decrease in molecular mass of soluble polysaccharides, using gel permeation chromatography. However, the treatment with H2O2 or O2(-) showed no significant effect on the disassembly of cellular wall polysaccharides. Furthermore, the degradation of the de-esterified AEIR was more susceptible to OH attack than the esterified AEIR. In addition, the effect of OH could be inhibited in the presence of OH scavenger. This study suggests that disassembly of cellular wall polysaccharides could be initiated by OH as the solublisation of the polysaccharides increased, which, in turn, accelerated fruit softening. Copyright © 2008 Elsevier Ltd. All rights reserved.

Meiotic Clade AAA ATPases: Protein Polymer Disassembly Machines.

Science.gov (United States)

Monroe, Nicole; Hill, Christopher P

2016-05-08

Meiotic clade AAA ATPases (ATPases associated with diverse cellular activities), which were initially grouped on the basis of phylogenetic classification of their AAA ATPase cassette, include four relatively well characterized family members, Vps4, spastin, katanin and fidgetin. These enzymes all function to disassemble specific polymeric protein structures, with Vps4 disassembling the ESCRT-III polymers that are central to the many membrane-remodeling activities of the ESCRT (endosomal sorting complexes required for transport) pathway and spastin, katanin p60 and fidgetin affecting multiple aspects of cellular dynamics by severing microtubules. They share a common domain architecture that features an N-terminal MIT (microtubule interacting and trafficking) domain followed by a single AAA ATPase cassette. Meiotic clade AAA ATPases function as hexamers that can cycle between the active assembly and inactive monomers/dimers in a regulated process, and they appear to disassemble their polymeric substrates by translocating subunits through the central pore of their hexameric ring. Recent studies with Vps4 have shown that nucleotide-induced asymmetry is a requirement for substrate binding to the pore loops and that recruitment to the protein lattice via MIT domains also relieves autoinhibition and primes the AAA ATPase cassettes for substrate binding. The most striking, unifying feature of meiotic clade AAA ATPases may be their MIT domain, which is a module that is found in a wide variety of proteins that localize to ESCRT-III polymers. Spastin also displays an adjacent microtubule binding sequence, and the presence of both ESCRT-III and microtubule binding elements may underlie the recent findings that the ESCRT-III disassembly function of Vps4 and the microtubule-severing function of spastin, as well as potentially katanin and fidgetin, are highly coordinated. Copyright © 2015 Elsevier Ltd. All rights reserved.

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

Capillarity-induced disassembly of virions in carbon nanotubes

Energy Technology Data Exchange (ETDEWEB)

Fan Xiaobin; Peng Wenchao; Li Yang; Li Xianyu; Zhang Guoliang; Zhang Fengbao [School of Chemical Engineering and Technology, Tianjin University, Tianjin (China); Barclay, J Elaine; Evans, David J [Department of Biological Chemistry, John Innes Centre, Norwich Research Park, Norwich NR4 7UH (United Kingdom)], E-mail: fbzhang@tju.edu.cn

2008-04-23

Studying the transport and fate of viruses through nanochannels is of great importance. By using the nanochannel of a carbon nanotube (CNT) as an ideal model, we evaluated the possibility of capillarity-induced viral transport through a closely fitting nanochannel and explored the mechanisms involved. It is shown both experimentally and theoretically that Cowpea mosaic virus can enter CNTs by capillarity. However, when introduced into a nanotube the protein capsid may disassemble. During the initial capillary filling stage, anomalous needle-shaped high pressure exists in the centre of the nanotube's entrance. This high pressure, combining with the significant negative pressure within the nanotube, may account for the disassembly of the virions.

A Label Correcting Algorithm for Partial Disassembly Sequences in the Production Planning for End-of-Life Products

Directory of Open Access Journals (Sweden)

Pei-Fang (Jennifer Tsai

2012-01-01

Full Text Available Remanufacturing of used products has become a strategic issue for cost-sensitive businesses. Due to the nature of uncertain supply of end-of-life (EoL products, the reverse logistic can only be sustainable with a dynamic production planning for disassembly process. This research investigates the sequencing of disassembly operations as a single-period partial disassembly optimization (SPPDO problem to minimize total disassembly cost. AND/OR graph representation is used to include all disassembly sequences of a returned product. A label correcting algorithm is proposed to find an optimal partial disassembly plan if a specific reusable subpart is retrieved from the original return. Then, a heuristic procedure that utilizes this polynomial-time algorithm is presented to solve the SPPDO problem. Numerical examples are used to demonstrate the effectiveness of this solution procedure.

Decontamination of TRIGA Mark II reactor, Indonesia

International Nuclear Information System (INIS)

Suseno, H.; Daryoko, M.; Goeritno, A.

2002-01-01

The TRIGA Mark II Reactor in the Centre for Research and Development Nuclear Technique Bandung has been partially decommissioned as part of an upgrading project. The upgrading project was carried out from 1995 to 2000 and is being commissioned in 2001. The decommissioning portion of the project included disassembly of some components of the reactor core, producing contaminated material. This contaminated material (grid plate, reflector, thermal column, heat exchanger and pipe) will be sent to the Decontamination Facility at the Radioactive Waste Management Development Centre. (author)

Studsvik's R2 reactor - Review of activities

Energy Technology Data Exchange (ETDEWEB)

Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

1993-07-01

A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

Modularized mirror fusion reactor concept with emphasis on fabricability, assembly, and disassembly

International Nuclear Information System (INIS)

Peterson, M.A.; Werner, R.W.; Hoffman, M.A.; Carlson, G.A.

1975-01-01

A progress report on a continuing study directed toward the development of mirror reactor designs which simultaneously satisfy the various engineering, economic, and maintenance consideration is presented. Two new blanket and coil structure designs are presented which satisfy engineering requirements equally as well as previous designs while offering substantial gains in accessibility for maintenance. Because of the commercial requirement for a high duty cycle and the possible high frequency of blanket module removal--for either maintenance replacement--the module removal must be accomplished quickly with a minimum disruption of reactor operations. The blanket and coil structure designs allow the removal of any one of the identical blanket modules without disturbing either the remaining modules or the coil and its associated support structure. With fabricated coil structure costs estimated at $2.50/lbm and the reactor net electrical power calculated from a plasma and reactor system model detailed in the paper, coil and support structure costs of between 100 to 200 $/kwe were estimated. (U.S.)

Shape memory polymer nanocomposites for application of multiple-field active disassembly: experiment and simulation.

Science.gov (United States)

Carrell, John; Zhang, Hong-Chao; Wang, Shiren; Tate, Derrick

2013-11-19

Active disassembly (AD) uses innovative materials that can perform a designed disassembly action by the application of an external field. AD provides improvements over current disassembly processes by limiting machine or manual labor and enabling batch processing for end-of-life products. With improved disassembly operations, more reuse of components and purer recycling streams may be seen. One problem with AD, however, has been with the single-field actuation because of the probability of accidental disassembly. This presentation will discuss the application of shape memory polymer (SMP) nanocomposites in a new AD process. This novel AD process requires multiple-field actuation of the SMP nanocomposite fastener. In the analysis of this AD process, thermal and magnetic field tests were performed on the SMP nanocomposite. From these tests, finite-element analysis was performed to model and simulate the multiple-field AD process. The results of the simulations provide performance variables for the AD process and show a better performance time for the SMP nanocomposite fastener than for a comparable SMP fastener.

Management of European fast reactor R and D

International Nuclear Information System (INIS)

Judd, A.M.; Sheriff, N.

1993-01-01

Since 1984 government-funded fast reactor R and D in France, Germany and the UK has been run as a collaborative activity, and since 1988 as a unified programme in support of the design and construction of the advanced European Fast Reactor. This paper describes the international management structure which has been set up, and the means used to control the work. It is written from the point of view of those engaged in the project, and makes no attempt at a formal analysis of the structure. The main difficulty is that control of funding remains with the three governments. The R and D programme has to be managed so that it meets the needs of each government separately as well as the designers' requirements. To start with the management structure was excessively bureaucratic, but it has become more flexible and efficient. This has happened as the initial nationalistic suspicions have broken down, and the staff engaged in the work have learnt more about each others' ways of working so that an atmosphere of trust and inter-dependence has grown up. (This paper was written before the changes in UK policy on fast reactor development were announced in November 1992). (Author)

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

Energy Technology Data Exchange (ETDEWEB)

Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

Mesquita, Henrique F.A.; Ferreira, Andrea V.

2015-01-01

The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

Fast reactors: R and D targets and outlook for their introduction

International Nuclear Information System (INIS)

Poplavsky, V.; Barre, B.; Aizawa, K.

1997-01-01

In this paper the current status of fast reactors development is briefly outlined, including experimental, demonstration, and commercial installations. Data on the experience gained in development and operation of NPPs with reactors of this type are presented. The issues are discussed in connection with possibilities of fast reactor development in the nuclear power structure for the near (up to 2010-2020) and distant future. In the final part of the paper, an analysis is given of possible ways for R and D development in the field of NPPs with fast neutron reactors. (author)

Mathematical and Simulation Modelling of Moisture Diffusion Mechanism during Plastic IC Packages Disassembly

Directory of Open Access Journals (Sweden)

Peng Mou

2013-01-01

Full Text Available Reuse of plastic IC packages disassembled from printed circuit boards (PCBs has significant environmental benefits and economic value. The interface delamination caused by moisture diffusion is the main failure mode of IC packages during the disassembling process, which greatly reduces the reusability and reliability of disassembled IC packages. Exploring moisture diffusion mechanism is a prerequisite to optimize prebaking processes before disassembling that is an effective way to avoid the interface delamination. To this end, a computational model with variable boundary conditions is developed based on the different combination state of water in IC packages. The distribution characteristics and mechanism of moisture diffusion behavior are analyzed including the humidity distribution field and the relation among baking temperature, water loss rate, and baking time during baking process, and then the results are validated by FEA simulation based on the improved definition of relative moisture concentration. Baking under variable temperature is proposed and compared with the baking process and baking efficiency under constant temperature to find out the optimized baking parameters. Finally, a set of curves which indicate the relation between baking energy consumption and temperature are determined under actual industrial baking experiments, which could be used as references to develop industrial standards for PCB disassembling process.

Development Plan and R and D Status of China Lead-based Reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS [Institute of Nuclear Energy Safety Technology, Beijing (Switzerland)

2013-07-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Development Plan and R and D Status of China Lead-based Reactor

International Nuclear Information System (INIS)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS

2013-01-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

An improved slow neutron spectrometer at nuclear research reactor et-r r-1. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Abu El-Ela, M A [Reactor and Neutron Physics, Nuclear Research Center, AEA, Cairo (Egypt)

1996-03-01

An improved slow neutron selector has been aligned at channel number 6 of the nuclear research reactor ET-R R-1 Inshas. The flight path is 4 meter. The collimator-rotor-collimator system has the dimensions 0.3 x 2.5 x 70 cm with the rotor diameter 16 cm and 3 slits of 0.3 x 2.5 cm cross section. The rotor rotation rate varies between 600 r.p.m. the counting system has one of the best modern high electronic advanced technology time analyzer with minimum dwell time 2 sec, 8192 channels and a double detector inputs of TTL and NEG NIM standard pulses. The analyzer external triggering signals are of TTL standard type. A special design {sup 3} He detector for time of flight spectrometry has been used in the SNS. The reactor bare thermal neutron spectrum has been successfully measured, to show good agreement with the previous data. 6 figs.

Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

International Nuclear Information System (INIS)

Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae

2016-01-01

The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods

Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods.

Chrysobrycon eliasi, new species of stevardiine fish (Characiformes: Characidae from the río Madre de Dios and upper río Manuripe basins, Peru

Directory of Open Access Journals (Sweden)

James Anyelo Vanegas-Ríos

Full Text Available Chrysobrycon eliasi is described from several drainages of río Madre de Dios and upper río Manuripe basins, río Madeira basin, Peru. The new species is distinguished from its congeners by the maxillary teeth, which are tricuspidate, bicuspidate and conical, and the terminal lateral-line tube developed between caudal-fin rays 10-11. Other characters that help to recognize C. eliasi are the possession of maxilla with 6-15 (usually 11 teeth occupying more than 70% of the length of the maxilla in adults, predorsal scales 18-22, dorsal-fin to hypural complex length 36.32-41.17% of standard length (SL, gill-gland length in males 3.57-5.05% SL, maxillary length 33.76-38.75% of head length, and branched anal-fin rays 24-30. The discovery of C. eliasi allows us to extend the geographic distribution of the genus to the southeast, into the río Madeira basin.

R and D on fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

2012-01-01

Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

New digital control system for the operation of the Colombian research reactor IAN-R1

International Nuclear Information System (INIS)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

2015-09-01

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

Status of the pit disassembly and conversion facility

International Nuclear Information System (INIS)

Wood, Warren T.; Christensen, Lowell T.

2000-01-01

A planned new facility, the Pit Disassembly and Conversion Facility (PDCF), will be used to disassemble the nation's inventory of surplus nuclear weapons pits and convert the plutonium from those pits into a form suitable for storage, international inspection, and final disposition. Sized to handle 35 metric tons of plutonium from pits and other sources over its 10-year operating life, the PDCF will apply the Advanced Recovery and Integrated Extraction System (ARIES) technology. ARIES process technology has been developed at Los Alamos National Laboratory (LANL) and Lawrence Livermore National Laboratory (LLNL), and an integrated system is being demonstrated at LANL. The Los Alamos National Laboratory is the lead for technical design oversight of the PDCF. Technical data gained from the ARIES demonstrations is integral for the proper design of the PDCF

Evaporation Basin Test Reactor Area, Idaho National Engineering Laboratory: Environmental assessment

International Nuclear Information System (INIS)

1991-12-01

The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0501, on the construction and operation of the proposed Evaporation Basin at the Test Reactor Area (TRA) at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact statement (EIS) is not required, and the Department is issuing this Finding of No Significant Impact

Removing of the fuel element storage basin by explosive technique in the course of decommissioning of the Nuclear research reactor FR 2 in the research center Karlsruhe. Permission and technical execution

International Nuclear Information System (INIS)

Jehle, P.; Freund, H.U.

1999-01-01

The fuel element storage basin was removed by explosive technique in the course of the decommissioning of the nuclear research reactor FR 2. This basin consisted of 6.800 tons of heavily reinforced concrete with 25 tons of stainless steel lining. The reactor building including residual radioactive inventory was constructionally connected to the basin and had to stay undamaged. For this reason and due to the fact that the storage basin as operational part of the nuclear facility was subject to the German nuclear law the dismantling had to follow stringent licensing and control requirements. Special restrictions concerned the tolerable reactor building vibrations and the direct blast loadings which could affect the structural integrity of the building enclosing the basin. The explosive dismantling operations which also included the final removal of the building foundation were successfully performed in 90 separate blasts employing 490 kg of commercial explosives. (orig.) [de

Modifications done in the IPR-R1 reactor and their auxiliary systems

International Nuclear Information System (INIS)

Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

1986-01-01

The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

The stand prototype of minimum power NRE reactor

International Nuclear Information System (INIS)

Belogurov, A.I.; Grigorenko, L.N.; Mamontov, Yu.I.; Rachuk, V.S.; Stukalov, A.I.; Konyukhov, G.V.

1995-01-01

For ensuring of full-scale development of nuclear rocket engine (NRE) reactor was created stand prototype (reactor IRGIT?) The main differences of its are as follows: 1) Fasteners of technologies channels contents fuel assemblies in bottom are worked out the split. It is provides possibility a distance channels change without disassembly of reactor stand prototype from stand; 2) Cooling of the vessels, the moderator, the reflector and the barrel actuate is carried out by hydrogen; 3) The lower bottom modified for organization the hydrogen efflux in the form a reactor jet; 4) Radiation defence is introduced as part of stand prototype for ensuring of serviceability of stand accessories and tests routine service; 5) Each technology channels is provided of critical nozzle; 6) Control, regulation and defence of reactor has being carried out on stand system

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

Design and properties of marine reactors and associated R and D

Energy Technology Data Exchange (ETDEWEB)

Gagarinski, A; Ignatiev, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

Design and properties of marine reactors and associated R and D

International Nuclear Information System (INIS)

Gagarinski, A.; Ignatiev, V.; Devell, L.

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs

Radioactive inventory in structural materials of ET-R R-1 reactor and its implication on decommissioning.

Energy Technology Data Exchange (ETDEWEB)

Elkady, A; Amin, E [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

A plan for decommissioning of ET-R R-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences decommissioning. Conservative calculations have been made to evaluate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are present in significant quantities in the reactor structural materials are aluminium, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from Co-60 and Fe-55 which are present in aluminium as trace elements in larger quantities in other construction materials. 2 figs., 4 tabs.

Phosphatase PP2A and microtubule-mediated pulling forces disassemble centrosomes during mitotic exit

Directory of Open Access Journals (Sweden)

Stephen J. Enos

2018-01-01

Full Text Available Centrosomes are microtubule-nucleating organelles that facilitate chromosome segregation and cell division in metazoans. Centrosomes comprise centrioles that organize a micron-scale mass of protein called pericentriolar material (PCM from which microtubules nucleate. During each cell cycle, PCM accumulates around centrioles through phosphorylation-mediated assembly of PCM scaffold proteins. During mitotic exit, PCM swiftly disassembles by an unknown mechanism. Here, we used Caenorhabditis elegans embryos to determine the mechanism and importance of PCM disassembly in dividing cells. We found that the phosphatase PP2A and its regulatory subunit SUR-6 (PP2ASUR-6, together with cortically directed microtubule pulling forces, actively disassemble PCM. In embryos depleted of these activities, ∼25% of PCM persisted from one cell cycle into the next. Purified PP2ASUR-6 could dephosphorylate the major PCM scaffold protein SPD-5 in vitro. Our data suggest that PCM disassembly occurs through a combination of dephosphorylation of PCM components and force-driven fragmentation of the PCM scaffold.

Direct measurement of conformational strain energy in protofilaments curling outward from disassembling microtubule tips.

Science.gov (United States)

Driver, Jonathan W; Geyer, Elisabeth A; Bailey, Megan E; Rice, Luke M; Asbury, Charles L

2017-06-19

Disassembling microtubules can generate movement independently of motor enzymes, especially at kinetochores where they drive chromosome motility. A popular explanation is the 'conformational wave' model, in which protofilaments pull on the kinetochore as they curl outward from a disassembling tip. But whether protofilaments can work efficiently via this spring-like mechanism has been unclear. By modifying a previous assay to use recombinant tubulin and feedback-controlled laser trapping, we directly demonstrate the spring-like elasticity of curling protofilaments. Measuring their mechanical work output suggests they carry ~25% of the energy of GTP hydrolysis as bending strain, enabling them to drive movement with efficiency similar to conventional motors. Surprisingly, a β-tubulin mutant that dramatically slows disassembly has no effect on work output, indicating an uncoupling of disassembly speed from protofilament strain. These results show the wave mechanism can make a major contribution to kinetochore motility and establish a direct approach for measuring tubulin mechano-chemistry.

Improved Nondestructive Disassembly Process using Augmented Reality and RFID Product/Part Tracking

Directory of Open Access Journals (Sweden)

Ile Mircheski

2017-11-01

Full Text Available The waste from electric and electronic equipment and discarded automobiles in the past grew rapidly and resulted with waste in billions of tones. The aim of this paper is to present an improved non-destructive disassembly process of electromechanical products using augmented reality based devices, such as glasses, tablets or mobile phones, and RFID technology for valuable product/part tracking. The proposed method includes tagging of components of interest in the product assembly by using a RFID tag. The valuable product is marked with specific ID number written in the RFID tag, in order to declare the product. The relevant data such as material and weight of components, guidelines for non-destructive disassembly for the valuable product and removing of component of interest will be obtained with the assistance of RFID tag and a centralized database. This modular system offers guidelines for the non-destructive disassembly process for obtaining valuable component of interest intended for easy repairs, remanufacture, reuse or recycling. The guidelines are in video presentation format using augmented reality for easy visualization of non-destructive disassembly process. The benefits of proposed modular system includes biggest percentage of reuse of the valuable components, easy maintaining, improved material recycling, environmental protection and greater total return form end of life products.

Design and analysis of sustainable computer mouse using design for disassembly methodology

Science.gov (United States)

Roni Sahroni, Taufik; Fitri Sukarman, Ahmad; Agung Mahardini, Karunia

2017-12-01

This paper presents the design and analysis of computer mouse using Design for Disassembly methodology. Basically, the existing computer mouse model consist a number of unnecessary part that cause the assembly and disassembly time in production. The objective of this project is to design a new computer mouse based on Design for Disassembly (DFD) methodology. The main methodology of this paper was proposed from sketch generation, concept selection, and concept scoring. Based on the design screening, design concept B was selected for further analysis. New design of computer mouse is proposed using fastening system. Furthermore, three materials of ABS, Polycarbonate, and PE high density were prepared to determine the environmental impact category. Sustainable analysis was conducted using software SolidWorks. As a result, PE High Density gives the lowers amount in the environmental category with great maximum stress value.

IPR-R1 TRIGA research reactor decommissioning plan

International Nuclear Information System (INIS)

Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

2008-01-01

The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

Energy Technology Data Exchange (ETDEWEB)

Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

2007-08-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

The Canadian R and D program targeted at CANDU reactors

International Nuclear Information System (INIS)

Moeck, E.O.

1988-01-01

CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

International Nuclear Information System (INIS)

Paiva, Rosemeire P.; Salvetti, Tereza C.

2005-01-01

Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

Addendum to the Closure Report for Corrective Action Unit 113: Area 25 R-MAD Facility, Nevada National Security Site, Nevada

International Nuclear Information System (INIS)

2011-01-01

This addendum to the Closure Report for Corrective Action Unit 113: Area 25, Reactor Maintenance, Assembly, and Disassembly Facility, Building 3110, Nevada Test Site, Nevada, DOE/NV--891-VOL I-Rev. 1, dated July 2003, provides details of demolition, waste disposal, and use restriction (UR) modification for Corrective Action Unit 113, Area 25 R-MAD Facility. Demolition was completed on July 15, 2010, when the last of the building debris was disposed. Final field activities were concluded on August 30, 2010, after all equipment was demobilized and UR signs were posted. This work was funded by the American Recovery and Reinvestment Act.

The assembly and disassembly of ecological networks.

Science.gov (United States)

Bascompte, Jordi; Stouffer, Daniel B

2009-06-27

Global change has created a severe biodiversity crisis. Species are driven extinct at an increasing rate, and this has the potential to cause further coextinction cascades. The rate and shape of these coextinction cascades depend very much on the structure of the networks of interactions across species. Understanding network structure and how it relates to network disassembly, therefore, is a priority for system-level conservation biology. This process of network collapse may indeed be related to the process of network build-up, although very little is known about both processes and even less about their relationship. Here we review recent work that provides some preliminary answers to these questions. First, we focus on network assembly by emphasizing temporal processes at the species level, as well as the structural building blocks of complex ecological networks. Second, we focus on network disassembly as a consequence of species extinctions or habitat loss. We conclude by emphasizing some general rules of thumb that can help in building a comprehensive framework to understand the responses of ecological networks to global change.

ARIES pit disassembly-safeguards issues for transparency

International Nuclear Information System (INIS)

Fearey, B.L.; Cremers, T.L.

1995-01-01

Historic changes are now occurring in U.S. nonproliferation and arms control policy. The quantity of nuclear weapons required to provide a deterrence has decreased (especially with the end of the Cold War). Further, various bilateral and multilateral treaties now require the removal of numerous nuclear weapons from the U.S. stockpile. Although the removal of such weapons appears straightforward, the final disposition of the surplus weapons-grade nuclear material must be carefully considered. Domestically, several plutonium disposition plans are now under consideration concerning long-term safety, materials accounting, environmental impact, accessibility, and long-term containment. The Automated Retirement and Integrated Extraction System (ARIES) currently under development at Los Alamos National Laboratory is one such disposition method for the disassembly of plutonium weapons components (pits). The ARIES system integrates and automates several features: disassembly of pits, consolidation of the plutonium material, on-line measurement of final products, waste streams, and long-term packaging. Clearly, in any plutonium disposition plan, the safeguards aspects of materials control and accounting and the security aspects must be carefully considered and evaluated

Spent-fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

International Nuclear Information System (INIS)

Townes, G.A.

1980-02-01

Doubling pool storage and tripling truck cask shipping capability appear feasible based on preliminary development work performed at the BNFP. This would be accomplished by disassembly of the fuel assembly and canning of the fuel pins. The dry, remote disassembly and canning in the headend cells offer adequate rates, easily controlled operator exposure, and visual inspection of the fuel pins and operations through shielded windows

Balancing Disassembly Line in Product Recovery to Promote the Coordinated Development of Economy and Environment

Directory of Open Access Journals (Sweden)

Jia Liu

2017-02-01

Full Text Available For environmentally conscious and sustainable manufacturing, many more manufacturers are acting to recycle and remanufacture their post-consumed products. The most critical process of remanufacturing is disassembly, since it allows for the selective extraction of the valuable components and materials from returned products to reduce the waste disposal volume. It is, therefore, important to design and balance the disassembly line to work efficiently due to its vital role in effective resource usage and environmental protection. Considering the disassembly precedence relationships and sequence-dependent parts removal time increments, this paper presents an improved discrete artificial bee colony algorithm (DABC for solving the sequence-dependent disassembly line balancing problem (SDDLBP. The performance of the proposed algorithm was tested against nine other approaches. Computational results evidently indicate the superior efficiency of the proposed algorithm for addressing the environmental and economic concerns while optimizing the multi-objective SDDLBP.

Closure Report for Corrective Action Unit 254: Area 25, R-MAD Decontamination Facility, Nevada Test Site, Nevada

Energy Technology Data Exchange (ETDEWEB)

G. N. Doyle

2002-02-01

Corrective Action Unit (CAU) 254 is located in Area 25 of the Nevada Test Site (NTS), approximately 100 kilometers (km) (62 miles) northwest of Las Vegas, Nevada. The site is located within the Reactor Maintenance, Assembly and Disassembly (R-MAD) compound and consists of Building 3126, two outdoor decontamination pads, and surrounding areas within an existing fenced area measuring approximately 50 x 37 meters (160 x 120 feet). The site was used from the early 1960s to the early 1970s as part of the Nuclear Rocket Development Station program to decontaminate test-car hardware and tooling. The site was reactivated in the early 1980s to decontaminate a radiologically contaminated military tank. This Closure Report (CR) describes the closure activities performed to allow un-restricted release of the R-MAD Decontamination Facility.

Nuclear material control at IEA-R1 nuclear research reactor

International Nuclear Information System (INIS)

1988-01-01

The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

On the optimal design of the disassembly and recovery processes.

Science.gov (United States)

Xanthopoulos, A; Iakovou, E

2009-05-01

This paper tackles the problem of the optimal design of the recovery processes of the end-of-life (EOL) electric and electronic products, with a special focus on the disassembly issues. The objective is to recover as much ecological and economic value as possible, and to reduce the overall produced quantities of waste. In this context, a medium-range tactical problem is defined and a novel two-phased algorithm is presented for a remanufacturing-driven reverse supply chain. In the first phase, we propose a multicriteria/goal-programming analysis for the identification and the optimal selection of the most 'desirable' subassemblies and components to be disassembled for recovery, from a set of different types of EOL products. In the second phase, a multi-product, multi-period mixed-integer linear programming (MILP) model is presented, which addresses the optimization of the recovery processes, while taking into account explicitly the lead times of the disassembly and recovery processes. Moreover, a simulation-based solution approach is proposed for capturing the uncertainties in reverse logistics. The overall approach leads to an easy-to-use methodology that could support effectively middle level management decisions. Finally, the applicability of the developed methodology is illustrated by its application on a specific case study.

On the optimal design of the disassembly and recovery processes

International Nuclear Information System (INIS)

Xanthopoulos, A.; Iakovou, E.

2009-01-01

This paper tackles the problem of the optimal design of the recovery processes of the end-of-life (EOL) electric and electronic products, with a special focus on the disassembly issues. The objective is to recover as much ecological and economic value as possible, and to reduce the overall produced quantities of waste. In this context, a medium-range tactical problem is defined and a novel two-phased algorithm is presented for a remanufacturing-driven reverse supply chain. In the first phase, we propose a multicriteria/goal-programming analysis for the identification and the optimal selection of the most 'desirable' subassemblies and components to be disassembled for recovery, from a set of different types of EOL products. In the second phase, a multi-product, multi-period mixed-integer linear programming (MILP) model is presented, which addresses the optimization of the recovery processes, while taking into account explicitly the lead times of the disassembly and recovery processes. Moreover, a simulation-based solution approach is proposed for capturing the uncertainties in reverse logistics. The overall approach leads to an easy-to-use methodology that could support effectively middle level management decisions. Finally, the applicability of the developed methodology is illustrated by its application on a specific case study

Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

International Nuclear Information System (INIS)

Shaw, K.W.

1993-11-01

The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

Modification of the IAN-R1 reactor

International Nuclear Information System (INIS)

Jaime, J.; Ahumada, S.; Spin, R.A.

1990-01-01

The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

ORNL R and D on advanced small and medium power reactors: selected topics

International Nuclear Information System (INIS)

White, J.D.; Trauger, D.B.

1989-01-01

From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, an assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable R and D would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current R and D efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described

Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

International Nuclear Information System (INIS)

Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

2010-01-01

In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

Science.gov (United States)

Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

2014-10-01

The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

Strengthening the R and D on fast reactor technology, and promoting its industrialization

International Nuclear Information System (INIS)

Wan Gang

2008-01-01

Based on the strategic thoughts of energy development in China expounded by Jiang Zemin in the article entitled 'Reflections on Energy Issues in China', the author points out in this paper that R and Ds on fast reactor technology shall be carried out timely in China by taking full advantage of international scientific resources, and overall planning in this regard shall be made as well. The point of view of strengthening fast reactor technology R and D and promoting its industrialization is also put forward in the paper. (authors)

Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias

2005-01-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

2009-01-01

The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

Safety approach and R and D program for future french sodium-cooled fast reactors

International Nuclear Information System (INIS)

Beils, Stephane; Carluec, Bernard; Devictor, Nicolas; Fiorini, Gian Luigi; Sauvage, Jean Francois

2011-01-01

This paper presents briefly the safety approach as well as the R and D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R and D. B) Strategy and roadmap in support of the R and D for future SFRs. This section describes the R and D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype. (author)

A Taxonomy and Comparison of Haptic Actions for Disassembly Tasks

National Research Council Canada - National Science Library

Bloomfield, Aaron; Deng, Yu; Wampler, Jeff; Rondot, Pascale; Harth, Dina; McManus, Mary; Badler, Norman

2003-01-01

.... We conducted a series of human subject experiments to compare user performance and preference on a disassembly task with and without haptic feedback using CyberGlove, Phantom, and SpaceMouse interfaces...

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Trends on R and D of the innovative nuclear reactors in Japan

International Nuclear Information System (INIS)

Kinoshita, Izumi

2002-01-01

In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

International Nuclear Information System (INIS)

Walton, L.A.

1981-01-01

A method is described of joining burnable poison rods to the spider arms of a pressurised water power reactor fuel assembly which is proof against the reactor core environment but permits these rods to be removed from the spider simply, swiftly and delicately. (U.K.)

Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

International Nuclear Information System (INIS)

Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

2013-06-01

(Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

Substantiation of the admissable concentration of radionuclides in the utilization of concrete from disassembled reactors

International Nuclear Information System (INIS)

Engovatov, I.A.; Mashkovich, V.P.; Morev, M.N.

1995-01-01

The complete or partial disassembly of the buildings and systems of a modern nuclear power plant results in the formation of hundreds of thousands of tons of wastes. More than 90% of this mass consists of reinforced concrete. Only a small fraction of these materials is contaminated and/or activated up to high level and must be treated as radioactive wastes that must be buried. For this reason, it is helpful to consider a variant of recycling of some of the wastes and secondary utilization of these wastes in different industrial production processes. In this paper, we analyze the concrete that is freed when a nuclear power plant is decommissioned, and in particular, we examine three scenarios: (1) stockpiling and long-term storage of concrete at an industrial site, (2) highway construction, and (3) industrial building construction. Admissable radionuclide concentrations of several isotopes are tabulated for each scenario, and the results provide the basic information for the development of standards for the utilization of the concrete wastes produced during the decommissioning of nuclear power plants

Optimization of Two-Level Disassembly/Remanufacturing/Assembly System with an Integrated Maintenance Strategy

Directory of Open Access Journals (Sweden)

Zouhour Guiras

2018-04-01

Full Text Available With an increase of environmental pressure on economic activities, reverse flow is increasingly important. It seeks to save resources, eliminate waste, and improve productivity. This paper investigates the optimization of the disassembly, remanufacturing and assembly system, taking into account assembly-disassembly system degradation. An analytical model is developed to consider disassembly, remanufacturing of used/end-of-life product and assembly of the finished product. The finished product is composed of remanufactured and new components. A maintenance policy is sequentially integrated to reduce the system unavailability. The aim of this study is to help decision-makers, under certain conditions, choose the most cost-effective process for them to satisfy the customer as well as to adapt to the potential risk that can perturb the disassembly-assembly system. A heuristic is developed to determine the optimal ordered date of the used end-of-life product as well as the optimum release dates of new external components. The results reveal that considering some remanufacturing and purchase components costs, the proposed model is more economical in comparison with a model without remanufactured parts. Numerical results are provided to illustrate the impact of the variation of the ordering cost and quality of the used end-of-life product on the system profitability. Finally, the risk due to system repair periods is discussed, which has an impact on managerial decision-making.

Development of a training simulator to operators of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carvalho, Ricardo Pinto de

2006-01-01

This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

Auxiliary control system of the safety parameters for IPR-R1 reactor

International Nuclear Information System (INIS)

Coura, J.G.

1986-01-01

This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

Energy Technology Data Exchange (ETDEWEB)

Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

International Nuclear Information System (INIS)

Kang, S. H.; Ha, K. S.

2013-01-01

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity feedback (Doppler and

Decontamination and decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

High-temperature reactor in modular construction

International Nuclear Information System (INIS)

Mueller, F.U.; Reutler, H.; Ullrich, M.

1981-01-01

Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

A Parameter Study of Large Fast Reactor Nuclear Explosion Accidents

Energy Technology Data Exchange (ETDEWEB)

Wiesel, J R

1969-02-15

An IBM-code EEM (Explosive Excursion Model) has been developed for calculating the energy releases associated with the explosive disassembly of a large fast reactor following a superprompt critical condition. The assumed failure chain of events and the possible core collapse following a fuel meltdown give the input data and initial conditions, the most important of which is the reactivity insertion rate at the moment of the explosive core disassembly. The dependence of the energy releases on the reactivity insertion rate, the Doppler reactivity feedback, the power form factor and the core size have been studied. The model enables a quick estimation of conservative values of the destructive mechanical energy releases following a nuclear explosion and gives suggestions as to how to reduce or even avoid such excursions.

A Parameter Study of Large Fast Reactor Nuclear Explosion Accidents

International Nuclear Information System (INIS)

Wiesel, J.R.

1969-02-01

An IBM-code EEM (Explosive Excursion Model) has been developed for calculating the energy releases associated with the explosive disassembly of a large fast reactor following a superprompt critical condition. The assumed failure chain of events and the possible core collapse following a fuel meltdown give the input data and initial conditions, the most important of which is the reactivity insertion rate at the moment of the explosive core disassembly. The dependence of the energy releases on the reactivity insertion rate, the Doppler reactivity feedback, the power form factor and the core size have been studied. The model enables a quick estimation of conservative values of the destructive mechanical energy releases following a nuclear explosion and gives suggestions as to how to reduce or even avoid such excursions

Verification of the linearity of the IPR-R1 TRIGA reactor power channels

International Nuclear Information System (INIS)

Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

2013-01-01

The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

Comprehensive cooling water study annual report. Volume IV: radionuclide and heavy metal transport, Savannah River Plant

International Nuclear Information System (INIS)

Gladden, J.B.; Lower, M.W.; Mackey, H.E.; Specht, W.L.; Wilde, E.W.

1985-07-01

The principal sources of tritium, radiocesium, and radiocobalt in the environment at the Savannah River Plant have been reactor area effluent discharges to onsite streams. Radioactive releases began in 1955, with the period of major reactor releases occurring between 1955 and 1968. Since the early 1970s, releases, except for tritium releases, have been substantially reduced. Radioisotope liquid releases resulted specifically from leaching of reactor fuel elements with cladding failures which exposed the underlying fuel to water. The direct sources of these releases were heat exchanger cooling water, spent fuel storage and disassembly basin effluents, and process water from each of the reactor areas. Offsite radiochemical monitoring of water and sediment at upriver and downriver water treatment facilities indicates that SRP contributions of gamma-emitting radionuclide levels present at these facilities are minute. Tritium in water attributable to SRP operations is routinely detected at the downriver facilities; however, total alpha and nonvolatile beta concentrations attributable to SRP liquid releases are not detected at the downriver facilities. The historic material balance calculated for onsite releases of tritium transported to the Savannah River exhibits a high accounting of tritium released. Other radionuclides released to onsite streams have primarily remained in onsite floodplains. Radionuclide releases associated with reactor operations are derived primarily from disassembly basin water releases in the reactor areas and historically have been the major source of radioactivity released to onsite streams. The movement and interaction of these releases have been governed by cooling water discharges. Liquid releases continue to meet DOE concentration guides for the various radioisotopes in onsite streams and in the Savannah River

Spent fuel management - two alternatives at the FiR 1 reactor

International Nuclear Information System (INIS)

Salmenhaara, S.E.J.

2001-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

What if the hand piece spring disassembles during robotic radical prostatectomy?

Science.gov (United States)

Akbulut, Ziya; Canda, Abdullah Erdem; Atmaca, Ali Fuat; Asil, Erem; Isgoren, Egemen; Balbay, Mevlana Derya

2011-01-01

Robot-assisted laparoscopic radical prostatectomy (RALRP) is successfully being performed for treating prostate cancer (PCa). However, instrumentation failure associated with robotic procedures represents a unique new problem. We report the successful completion of RALRP in spite of a disassembled hand piece spring during the procedure. A PubMed/Medline search was made concerning robotic malfunction and robot-assisted laparoscopic radical prostatectomy to discuss our experience. We performed RALRP in a 60-year-old male patient with localized PCa. During the procedure, the spring of the hand piece disassembled, and we were not able to reassemble it. We completed the procedure successfully however without fixing the disassembled hand piece spring. We were able to grasp tissue and needles when we brought our fingers together. The only movement we needed to do was to move fingers apart to release tissue or needles caught by robotic instrument. Although malfunction risk related to the da Vinci Surgical System seems to be very low, it might still occur. Sometimes, simple maneuvers may compensate for the failed function as occurred in our case. However, patients should be informed before the operation about the possibility of converting their procedure to laparoscopic or open due to robotic malfunction.

Phosphorylation of p37 is important for Golgi disassembly at mitosis

International Nuclear Information System (INIS)

Kaneko, Yayoi; Tamura, Kaori; Totsukawa, Go; Kondo, Hisao

2010-01-01

Research highlights: → p37 is phosphorylated on Serine-56 and Threonine-59 by Cdc2 at mitosis. → Phosphorylated p37 does not bind to Golgi membranes. → p37 phosphorylation inhibits p97/p37-mediated Golgi membrane fusion. -- Abstract: In mammals, the Golgi apparatus is disassembled at early mitosis and reassembled at the end of mitosis. For Golgi disassembly, membrane fusion needs to be blocked. Golgi biogenesis requires two distinct p97ATPase-mediated membrane fusion, the p97/p47 and p97/p37 pathways. We previously reported that p47 phosphorylation on Serine-140 by Cdc2 results in mitotic inhibition of the p97/p47 pathway . In this study, we demonstrate that p37 is phosphorylated on Serine-56 and Threonine-59 by Cdc2 at mitosis, and this phosphorylated p37 does not bind to Golgi membranes. Using an in vitro Golgi reassembly assay, we show that mutated p37(S56D, T59D), which mimics mitotic phosphorylation, does not cause any cisternal regrowth, indicating that p37 phosphorylation inhibits the p97/p37 pathway. Our results demonstrate that p37 phosphorylation on Serine-56 and Threonine-59 is important for Golgi disassembly at mitosis.

Disassembly and Sanitization of Classified Matter

International Nuclear Information System (INIS)

Stockham, Dwight J.; Saad, Max P.

2008-01-01

The Disassembly Sanitization Operation (DSO) process was implemented to support weapon disassembly and disposition by using recycling and waste minimization measures. This process was initiated by treaty agreements and reconfigurations within both the DOD and DOE Complexes. The DOE is faced with disassembling and disposing of a huge inventory of retired weapons, components, training equipment, spare parts, weapon maintenance equipment, and associated material. In addition, regulations have caused a dramatic increase in the need for information required to support the handling and disposition of these parts and materials. In the past, huge inventories of classified weapon components were required to have long-term storage at Sandia and at many other locations throughout the DoE Complex. These materials are placed in onsite storage unit due to classification issues and they may also contain radiological and/or hazardous components. Since no disposal options exist for this material, the only choice was long-term storage. Long-term storage is costly and somewhat problematic, requiring a secured storage area, monitoring, auditing, and presenting the potential for loss or theft of the material. Overall recycling rates for materials sent through the DSO process have enabled 70 to 80% of these components to be recycled. These components are made of high quality materials and once this material has been sanitized, the demand for the component metals for recycling efforts is very high. The DSO process for NGPF, classified components established the credibility of this technique for addressing the long-term storage requirements of the classified weapons component inventory. The success of this application has generated interest from other Sandia organizations and other locations throughout the complex. Other organizations are requesting the help of the DSO team and the DSO is responding to these requests by expanding its scope to include Work-for- Other projects. For example

AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

Energy Technology Data Exchange (ETDEWEB)

Ploger, Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowciz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts�

Core calculations for the upgrading of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

1998-01-01

The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

International Nuclear Information System (INIS)

D'Utra Bitelli, U.

1993-01-01

This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

New burnup calculation of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

2015-01-01

The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

Energy Technology Data Exchange (ETDEWEB)

Palmer, E.R. [Westinghouse Savannah River Company, AIKEN, SC (United States)

1997-07-01

Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA.

Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

International Nuclear Information System (INIS)

Palmer, E.R.

1997-07-01

Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

2013-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2016-01-01

The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

Development of the spent fuel disassembling process by utilizing the 3D graphic design technology

International Nuclear Information System (INIS)

Song, T. K.; Lee, J. Y.; Kim, S. H.; Yun, J. S.

2001-01-01

For developing the spent fuel disassembling process, the 3D graphic simulation has been established by utilizing the 3D graphic design technology which is widely used in the industry. The spent fuel disassembling process consists of a downender, a rod extraction device, a rod cutting device, a pellet extracting device and a skeleton compaction device. In this study, the 3D graphical design model of these devices is implemented by conceptual design and established the virtual workcell within kinematics to motion of each device. By implementing this graphic simulation, all the unit process involved in the spent fuel disassembling processes are analyzed and optimized. The 3D graphical model and the 3D graphic simulation can be effectively used for designing the process equipment, as well as the optimized process and maintenance process

Spent fuel management plans for the FiR 1 Reactor

International Nuclear Information System (INIS)

Salmenhaara, S. E. J.

2002-01-01

The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-07-01

This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

A Study on Dismantling of Westinghouse Type Nuclear Reactor

International Nuclear Information System (INIS)

Jeong, Woo-Tae; Lee, Sang-Guk

2014-01-01

KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

Design study of blanket structure for tokamak experimental fusion reactor

International Nuclear Information System (INIS)

1979-11-01

Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

Design of sodium cooled reactor systems and components for maintainability

International Nuclear Information System (INIS)

Carr, R.W.; Charnock, H.O.; McBride, J.P.

1978-09-01

Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

Cleaning, disassembly, and requalification of the FFTF in vessel handling machine

International Nuclear Information System (INIS)

Coops, W.J.

1977-10-01

The Engineering Model In Vessel Handling Machine (IVHM) was successfully removed, cleaned, disassembled, inspected, reassembled and reinstalled into the sodium test vessel at Richland, Washington. This was the first time in the United States a full size operational sodium wetted machine has been cleaned by the water vapor nitrogen process and requalified for operation. The work utilized an atmospheric control system during removal, a tank type water vapor nitrogen cleaning system and an open ''hands on'' disassembly and assembly stand. Results of the work indicate the tools, process and equipment are adequate for the non-radioactive maintenance sequence. Additionally, the work proves that a machine of this complexity can be successfully cleaned, maintained and re-used without the need to replace a large percentage of the sodium wetted parts

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

Measurements and calculations of reactivity for the IEA-R1 reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

1988-01-01

This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

International Nuclear Information System (INIS)

Duque O, J.; Velez A, L.H.

1987-01-01

Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

Hseih, S.Y.; Danby, G.; Powell, J.R.

1979-01-01

The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

2009-01-01

The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

PIK reactor construction status

International Nuclear Information System (INIS)

Konoplev, K.A.; Smolsky, S.L.

2001-01-01

The 100MW reactor PIK for fundamental researches has a thermal neutron flux of more than 10 15 n/cm 2 sec. This presentation outlines the construction state as of 2001, its prospects and completion tactics in the conditions of unstable finance. Construction of the reactor started in 1976. In 1986 construction of the building was completed and significant part of the installation work fulfilled. Construction of cooling systems was finished, the control panel assembled, and adjustment of the pump and gate valve control circuits started. After Chernobyl catastrophe, the USSR nuclear reactor safety requirements were revised. The PIK design did not meet these requirements and underwent considerable revision. The reconstruction design resulted in double the initial cost. Creation of the containment was the bulkiest part of the reconstruction. It brought about the need to disassemble the roofing of the building, dismantle all the equipment of the two upper floors, and lay up the equipment of the lower floors. As of 2001, construction in accordance with the revised design is at the stage of assemblage of the most important units, i.e. reactor itself, cooling system, heavy water system, and a number of auxiliary systems, such as depleted fuel storage, emergency cooling system etc. (orig.)

Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

Calculation of the main neutron parameters of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Ojima, Mario Katsuhiko

1977-01-01

The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

The PEC reactor. Safety analysis: Detailed reports

Energy Technology Data Exchange (ETDEWEB)

1988-01-01

In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

Neutron radiography in the IEA-R1 reactor

International Nuclear Information System (INIS)

Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

1988-08-01

Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

Reactor coolant purification system circulation pumps (CUW pumps)

International Nuclear Information System (INIS)

Tsutsui, Toshiaki

1979-01-01

Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

Disposal of sediments from the 1300-N Emergency Dump Basin

International Nuclear Information System (INIS)

Keen, R.; Duncan, G.M.

1996-01-01

This report describes the characterization of the 1300-N Emergency Dump Basin (EDB) sediments, summarizes the data obtained, the resultant waste categorization, and the preferred disposal method. The EDB is an outdoor, concrete storage pond with a 3/16-in. carbon steel liner. The basin (completed in 1963) originally served as a quenching pool for reactor blowdown in the event of a primary coolant leak. Later, the basin received blowdown from the N Reactor steam generators. The steam generator blowdowns and leading isolation valves allowed radioactively contaminated water (from primary and secondary reactor coolant leaks) to enter the basin. Windblown dust and sand have settled in the basin over the years (because of its outdoor location), causing the present layer of sediments. To minimize potential airborne contamination, the water level was kept constant by adding water. However, the addition of water was stopped to minimize the amount of contaminated water needing disposal. To ensure that the surfaces exposed as a result of evaporation pose no immediate airborne contaminant problem, the contamination levels are monitored by Radiation Control Technicians. As part of the deactivation of N Reactor facilities, the EDB will be stabilized for long-term surveillance and maintenance prior to final decontamination and demolition

Transport of fallout and reactor radionuclides in the drainage basin of the Hudson River estuary

International Nuclear Information System (INIS)

Simpson, H.J.; Linsalata, P.; Olsen, C.R.; Cohen, N.; Trier, R.M.

1987-01-01

Fallout plutonium and radiocesium derived from both weapons testing and local reactor releases are found in the water column and sediments of the Hudson River in readily measurable amounts. The history of fallout delivery and dissolved phase runoff from the drainage basin of 80 Sr, 137 Cs, and /sup 239,240/Pu have been extensively documented since the mid-1950s. Sediment and water column concentrations of 134 Cs, 137 Cs, and /sup 239,240/Pu in the Hudson have also been documented since the mid-1960's and are summarized. Since the peak fallout years, substantial portions of the fallout radionuclides in the drainage basin have become unavailable to normal weathering processes as reflected by a measured decrease in the fallout nuclide transport to the waters of the tidal Hudson. Budget calculations indicate that plutonium may be transported into the estuary from the coastal ocean, and that desorption of radiocesium from particles has allowed a substantial fraction of radiocesium to be exported from the Hudson to marine waters. 29 references, 6 figures, 8 tables

Status of the R and D activities on fast reactors and ADS in Brazil

International Nuclear Information System (INIS)

Maiorino, Jose Rubens

2001-01-01

Research and Development in Nuclear Science and Technology is conducted by Research Institutes of the Brazilian Nuclear Energy Commission. In Fast Reactor, R and D activities started in the sixties, and in 1972 a small Na loop (100 kW) was constructed. At the same time, during the seventies at IPEN, research in cooperation with GA for Gas Cooled Fast Breeder Reactor was conducted. The motivation of such research was Thorium Fuel Cycle. As a result of this research a Helium Loop was constructed and a Split Table Critical Assembly (ZPR) was designed. During the eighties, an agreement with ANSALDO-NIRA resulted in an acquisition of a Sodium Loop for Thermohydraulics studies, however it never had been assembled. At the same time, a concept of a Binary Breeder Reactor using two cycles, Th and U, was developed. During the nineties, a National Program to conduct R and D (pyroprocess; U-Zr Metallic Fuel; HT-9; Electromagnetic Pump; and a conceptual design of a Experimental Reactor (60/20 MWth/MWe)) was proposed, however it was closed at the end of the decade. Now, only academic research is being conducted, and it is summarized in this report. Basically, they are: an integral lead fast reactor concept for developing countries, and an alternative concept for a fast energy amplifier accelerator driven system. The first is an combination of best characteristics of the American Integral Fast Reactor and the Russian Lead Cooled Reactor. The second is a conceptual design of ADS helium cooled imbedded in a solid lead subcritical array of fuel, using more than one point of spallation trying to reduce the requirement for energy and current of the accelerator

Method for inspecting nuclear reactor fuel elements

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter

2015-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Current activities at the FiR 1 TRIGA reactor

International Nuclear Information System (INIS)

Salmenhaara, Seppo

2002-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

Modernization of the CDTN IPR-R1 TRIGA reactor instrumentation and control

International Nuclear Information System (INIS)

Mesquita, A.Z.; Costa, A.C.L.; Souza, R.M.G.P.

2009-01-01

The control system of the IPR-R1 was changed in 1995. Although since the year's 80 was generalized the use of microprocessor technology and video monitors for visual interface, in the IPR-R1 control room it was used analogical system by relay-based logic, and were maintained the mechanical strip chart recorders (ink-pen drive) to measure, monitor and store the operational parameters. It was maintained the measure and the control of, practically, the same variables of the original system, although the reactor power already have been upgraded to 100 kW and began the studies to increase it to 250 kW, which is the current core configuration. For 250 kW operations the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art instrumentation and control system using microprocessor technology is proposed to replace the present analogical systems. The new system can eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards as NRC and IEE specifications. This paper describes the research project in process in CDTN that has as objective the modernization of the IPR-R1 TRIGA reactor instrumentation and control of the operational variables. The project also will improve the accomplishment of neutronic and thermal-hydraulic experiments, foreseen in the CDTN research program. (author)

Improvement of the in vivo cellular repopulation of decellularized cardiovascular tissues by a detergent-free, non-proteolytic, actin-disassembling regimen.

Science.gov (United States)

Assmann, Alexander; Struß, Marc; Schiffer, Franziska; Heidelberg, Friederike; Munakata, Hiroshi; Timchenko, Elena V; Timchenko, Pavel E; Kaufmann, Tim; Huynh, Khon; Sugimura, Yukiharu; Leidl, Quentin; Pinto, Antonio; Stoldt, Volker R; Lichtenberg, Artur; Akhyari, Payam

2017-12-01

Low immunogenicity and high repopulation capacity are crucial determinants for the functional and structural performance of acellular cardiovascular implants. The present study evaluates a detergent-free, non-proteolytic, actin-disassembling regimen (BIO) for decellularization of heart valve and vessel grafts, particularly focusing on their bio-functionality. Rat aortic conduits (rAoC; n = 89) and porcine aortic valve samples (n = 106) are decellularized using detergents (group DET) or the BIO regimen. BIO decellularization results in effective elimination of cellular proteins and significantly improves removal of DNA as compared with group DET, while the extracellular matrix (ECM) structure as well as mechanical properties are preserved. The architecture of rAoC in group BIO allows for improved bio-functionalization with fibronectin (FN) in a standardized rat implantation model: BIO treatment significantly increases speed and amount of autologous medial cellular repopulation in vivo (p < 0.001) and decreases the formation of hyperplastic intima (p < 0.001) as compared with FN-coated DET-decellularized grafts. Moreover, there are no signs of infiltration with inflammatory cells. The present biological, detergent-free, non-proteolytic regimen balances effective decellularization and ECM preservation in cardiovascular grafts, and provides optimized bio-functionality. Additionally, this study implies that the actin-disassembling regimen may be a promising approach for bioengineering of acellular scaffolds from other muscular tissues, as for example myocardium or intestine. Copyright © 2017 John Wiley & Sons, Ltd. Copyright © 2017 John Wiley & Sons, Ltd.

Phosphorylation and disassembly of intermediate filaments in mitotic cells

International Nuclear Information System (INIS)

Chou, Yinghao; Rosevear, E.; Goldman, R.D.

1989-01-01

As baby hamster kidney (BHK-21) cells enter mitosis, networks of intermediate filaments (IFs) are transformed into cytoplasmic aggregates of protofilaments. Coincident with this morphological change, the phosphate content of vimentin increases from 0.3 mol of P i per mol of protein in interphase to 1.9 mol of P i per mol of protein in mitosis. A similar increase in phosphate content is observed with desmin, from 0.5 mol of P i per mol of protein to 1.5 mol of P i per mol of protein. Fractionation of mitotic cell lysates by hydroxylapatite column chromatography reveals the presence of two IF protein kinase activities, designated as IF protein kinase I and IF protein kinase II. Comparison of two-dimensional 32 P-labeled phosphopeptide maps of vimentin and desmin phosphorylated in vivo in mitosis, and in vitro using partially purified kinase fractions, reveals extensive similarity in the two sets of phosphorylation sites. Phosphorylation of in vitro polymerized IFs by IF protein kinase II induces complete disassembly as determined by negative-stain electron microscopy. The results support the idea that the disassembly of IFs in mitosis is regulated by the phosphorylation of its subunit proteins

Aspects of the Iea-R1 research reactor seismic evaluation

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1996-01-01

Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

Science.gov (United States)

2012-05-03

... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

Decommissioning of the 105-F and 105-H fuel storage basin in the 100 Area at the Hanford Site

International Nuclear Information System (INIS)

Griffin, P.W.

1991-09-01

The US Department of Energy (DOE) owns the eight surplus production reactors at the Hanford Site north of Richland, Washington. The fuel storage basins at the 105-F and 105-H reactors were filled with equipment, associated with the operation of the basins and clean fill in 1970. This was done to stabilize the residual sediment and a few feet of water in the reactors' fuel storage basins. This project investigates the subject basins to locate and remove overlooked fuel elements left in the basins

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

International Nuclear Information System (INIS)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

2011-01-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Energy Technology Data Exchange (ETDEWEB)

Wiersma, B.

2010-05-24

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill

Design-only conceptual design report for pit disassembly and conversion facility. Rev 0

International Nuclear Information System (INIS)

Zygmunt, S.; Christensen, L.; Richardson, C.

1997-01-01

This design-only conceptual design report (DOCDR) was prepared to support a funding request by the Department of Energy (DOE)-Office of Fissile Material Disposition (OFMD) for engineering design of the Pit Disassembly and Conversion Facility (PDCF) Project No. 99-D-141. The PDCF will be used to disassemble the nation's inventory of surplus nuclear weapons pits and convert the plutonium recovered from those pits into a form suitable for storage, international inspection, and final disposition. The PDCF is a complex consisting of a hardened building that will contain the plutonium processes in a safe and secure manner, and conventional buildings and structures that will house support personnel, systems, and equipment. The PDCF uses the Advanced Recovery and Integrated Extraction System (ARIES), a low waste, modular pyroprocessing system to convert pits to plutonium oxide. The PDCF project consists of engineering and design, and construction of the buildings and structures, and engineering and design, procurement, installation, testing and start-up of equipment to disassemble pits and convert plutonium in pits to oxide form. The facility is planned to operate for 10 years, averaging 3.5 metric tons (3.86 tons) of plutonium metal per year. On conclusion of operations, the PDCF will be decontaminated and decommissioned

Design-only conceptual design report for pit disassembly and conversion facility. Rev 0

Energy Technology Data Exchange (ETDEWEB)

Zygmunt, S.; Christensen, L.; Richardson, C.

1997-12-12

This design-only conceptual design report (DOCDR) was prepared to support a funding request by the Department of Energy (DOE)-Office of Fissile Material Disposition (OFMD) for engineering design of the Pit Disassembly and Conversion Facility (PDCF) Project No. 99-D-141. The PDCF will be used to disassemble the nation`s inventory of surplus nuclear weapons pits and convert the plutonium recovered from those pits into a form suitable for storage, international inspection, and final disposition. The PDCF is a complex consisting of a hardened building that will contain the plutonium processes in a safe and secure manner, and conventional buildings and structures that will house support personnel, systems, and equipment. The PDCF uses the Advanced Recovery and Integrated Extraction System (ARIES), a low waste, modular pyroprocessing system to convert pits to plutonium oxide. The PDCF project consists of engineering and design, and construction of the buildings and structures, and engineering and design, procurement, installation, testing and start-up of equipment to disassemble pits and convert plutonium in pits to oxide form. The facility is planned to operate for 10 years, averaging 3.5 metric tons (3.86 tons) of plutonium metal per year. On conclusion of operations, the PDCF will be decontaminated and decommissioned.

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Montmorillonite-induced Bacteriophage φ6 Disassembly

Science.gov (United States)

Trusiak, A.; Gottlieb, P.; Katz, A.; Alimova, A.; Steiner, J. C.; Block, K. A.

2012-12-01

It is estimated that there are 1031 virus particles on Earth making viruses an order of magnitude more prevalent in number than prokaryotes with the vast majority of viruses being bacteriophages. Clays are a major component of soils and aquatic sediments and can react with RNA, proteins and bacterial biofilms. The clays in soils serve as an important moderator between phage and their host bacteria, helping to preserve the evolutionary balance. Studies on the effects of clays on viral infectivity have given somewhat contradictory results; possibly a consequence of clay-virus interactions being dependent on the unique structure of particular viruses. In this work, the interaction between montmorillonite and the bacteriophage φ6 is investigated. φ6 is a member of the cystovirus family that infects Pseudomonas syringe, a common plant pathogen. As a member of the cystovirus family with an enveloped structure, φ6 serves as a model for reoviruses, a human pathogen. Experiments were conducted with φ6 suspended in dilute, purified homoionic commercial-grade montmorillonite over a range of virus:clay ratios. At a 1:100000 virus:clay ratio, the clay reduced viral infectivity by 99%. The minimum clay to virus ratio which results in a measurable reduction of P. syringae infection is 1:1. Electron microscopy demonstrates that mixed suspensions of smectite and virus co-aggregate to form flocs encompassing virions within the smectite. Both free viral particles as well as those imbedded in the flocs are seen in the micrographs to be missing the envelope- leaving only the nucleocapsid (NC) intact; indicating that smectite inactivates the virus by envelope disassembly. These results have strong implications in the evolution of both the φ6 virus and its P. syringae host cells. TEM of aggregate showing several disassembled NCs.

Hazards classification determination for PUREX fuel transfer to K-Basins

International Nuclear Information System (INIS)

Dodd, E.N. III.

1995-01-01

The PUREX Plant presently contains 2.9 metric tons of an aluminum clad Single Pass Reactor (SPR) fuel which is stored under water in four open top buckets in the PUREX slug storage basin. The PUREX dissolver cells contain approximately 0.5 metric tons of zirconium clad N Reactor fuel which was inadvertently placed into the process cell during charging operations. The dissolver N reactor elements will be recovered from the process floors using new crane operated tools. When the fuel shipment(s) is scheduled, the cask cars will be positioned into the PUREX rail tunnel and the overhead door will be opened. All the SPR fuel will be loaded into two cask rail cars inside four casks. The N Reactor fuel will be loaded into a separate rail car inside two or three casks. The car loading is initiated by opening the rail car lid and removing the cask lids. Prior to loading the canisters of N Reactor fuel, the canisters will be refilled with water (as needed) and a lid will be installed. The baskets of SPR fuel or canisters of N Reactor fuel will then be loaded into the casks. The lids to the casks will then be reinstalled and the car lids closed. The rail cars will then be decontaminated as necessary. The cask cars will be shipped either in two shipments or a combined single shipment using the rail route between PUREX and the K Basins. At the basin, the cask car will be positioned in the loadout area. The cask car lid will be opened and a single cask moved into the loadout pit, which is a lowered section of the basin. The cask lid is removed while the cask is lower into the pit. The fuel is then removed from the cask and stored in the basin. The cask is then removed, the lid reinstalled during removal, and the cask replaced into the cask car. This document identifies the hazard classification of the Fuel Transfer from the PUREX facility to K-Basins

Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2011-07-01

The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

International Nuclear Information System (INIS)

Auterinen, Iiro; Salmenhaara, Seppo

2008-01-01

FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

Heat transfer study of a submerged reactor channel under boil-off condition

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

2012-12-15

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

Studies and research concerning BNFP design and construction of a spent-fuel disassembly/encapsulation system

International Nuclear Information System (INIS)

Dabolt, R.J.

1981-04-01

Commercial light water power reactor operation in the United States is developing a cumbersome inventory of spent fuel. Systems for interim storage and handling of this fuel are being developed by the Federal Government and industry. Disassembly and canning of the spent fuel elements is one of these systems. It has the potential to double the storage capacity of a prereprocessing storage facility or to triple the capacity of conventional shipping casks. Prototype equipment and controls required to perform this operation in a dry environment have been primarily designed and fabricated at the Barnwell Nuclear Fuel Plant (BNFP). Ridihalgh, Eggers, and Associates have provided design support and fabrication of the control system. This system is capable of extracting and canning the fuel pins and compacting the nonfuel-bearing components of spent fuel assemblies at processing rates of 10 to 12 assemblies per day. The process also provides the potential for enhanced inspection and assay of spent fuel by reducing the interference encountered from the high gamma fields of the nonfuel bearing hardware

Asymmetric ring structure of Vps4 required for ESCRT-III disassembly

Science.gov (United States)

Caillat, Christophe; Macheboeuf, Pauline; Wu, Yuanfei; McCarthy, Andrew A.; Boeri-Erba, Elisabetta; Effantin, Gregory; Göttlinger, Heinrich G.; Weissenhorn, Winfried; Renesto, Patricia

2015-12-01

The vacuolar protein sorting 4 AAA-ATPase (Vps4) recycles endosomal sorting complexes required for transport (ESCRT-III) polymers from cellular membranes. Here we present a 3.6-Å X-ray structure of ring-shaped Vps4 from Metallosphera sedula (MsVps4), seen as an asymmetric pseudohexamer. Conserved key interface residues are shown to be important for MsVps4 assembly, ATPase activity in vitro, ESCRT-III disassembly in vitro and HIV-1 budding. ADP binding leads to conformational changes within the protomer, which might propagate within the ring structure. All ATP-binding sites are accessible and the pseudohexamer binds six ATP with micromolar affinity in vitro. In contrast, ADP occupies one high-affinity and five low-affinity binding sites in vitro, consistent with conformational asymmetry induced on ATP hydrolysis. The structure represents a snapshot of an assembled Vps4 conformation and provides insight into the molecular motions the ring structure undergoes in a concerted action to couple ATP hydrolysis to ESCRT-III substrate disassembly.

A Trio-Rac1-PAK1 signaling axis drives invadopodia disassembly

Science.gov (United States)

Moshfegh, Yasmin; Bravo-Cordero, Jose Javier; Miskolci, Veronika; Condeelis, John; Hodgson, Louis

2014-01-01

Rho family GTPases control cell migration and participate in the regulation of cancer metastasis. Invadopodia, associated with invasive tumor cells, are crucial for cellular invasion and metastasis. To study Rac1 GTPase in invadopodia dynamics, we developed a genetically-encoded, single-chain Rac1 Fluorescence Resonance Energy Transfer (FRET) biosensor. The biosensor shows Rac1 activity exclusion from the core of invadopodia, and higher activity when invadopodia disappear, suggesting that reduced Rac1 activity is necessary for their stability, and Rac1 activation is involved in disassembly. Photoactivating Rac1 at invadopodia confirmed this previously-unknown Rac1 function. We built an invadopodia disassembly model, where a signaling axis involving TrioGEF, Rac1, PAK1, and phosphorylation of cortactin, causing invadopodia dissolution. This mechanism is critical for the proper turnover of invasive structures during tumor cell invasion, where a balance of proteolytic activity and locomotory protrusions must be carefully coordinated to achieve a maximally invasive phenotype. PMID:24859002

77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

Science.gov (United States)

2012-11-15

... Research Institute TRIGA Reactor: Facility Operating License No. R-84 AGENCY: Nuclear Regulatory Commission... considering an application for the renewal of Facility Operating License No. R-84 (Application), which... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

Assembly, operation and disassembly manual for the Battelle Large Volume Water Sampler (BLVWS)

International Nuclear Information System (INIS)

Thomas, V.W.; Campbell, R.M.

1984-12-01

Assembly, operation and disassembly of the Battelle Large Volume Water Sampler (BLVWS) are described in detail. Step by step instructions of assembly, general operation and disassembly are provided to allow an operator completely unfamiliar with the sampler to successfully apply the BLVWS to his research sampling needs. The sampler permits concentration of both particulate and dissolved radionuclides from large volumes of ocean and fresh water. The water sample passes through a filtration section for particle removal then through sorption or ion exchange beds where species of interest are removed. The sampler components which contact the water being sampled are constructed of polyvinylchloride (PVC). The sampler has been successfully applied to many sampling needs over the past fifteen years. 9 references, 8 figures

Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

Energy Technology Data Exchange (ETDEWEB)

Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

2014-08-15

IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

Ufd1-Npl4 Recruit Cdc48 for Disassembly of Ubiquitylated CMG Helicase at the End of Chromosome Replication

Directory of Open Access Journals (Sweden)

Marija Maric

2017-03-01

Full Text Available Disassembly of the Cdc45-MCM-GINS (CMG DNA helicase is the key regulated step during DNA replication termination in eukaryotes, involving ubiquitylation of the Mcm7 helicase subunit, leading to a disassembly process that requires the Cdc48 “segregase”. Here, we employ a screen to identify partners of budding yeast Cdc48 that are important for disassembly of ubiquitylated CMG helicase at the end of chromosome replication. We demonstrate that the ubiquitin-binding Ufd1-Npl4 complex recruits Cdc48 to ubiquitylated CMG. Ubiquitylation of CMG in yeast cell extracts is dependent upon lysine 29 of Mcm7, which is the only detectable site of ubiquitylation both in vitro and in vivo (though in vivo other sites can be modified when K29 is mutated. Mutation of K29 abrogates in vitro recruitment of Ufd1-Npl4-Cdc48 to the CMG helicase, supporting a model whereby Ufd1-Npl4 recruits Cdc48 to ubiquitylated CMG at the end of chromosome replication, thereby driving the disassembly reaction.

Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

2015-07-01

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

High-field, high-density tokamak power reactor

International Nuclear Information System (INIS)

Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

1977-11-01

A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a open-quotes critical eventclose quotes in a geologic system

International Nuclear Information System (INIS)

Gauthier-Lafaye, F.; Holliger, P.; Blanc, P.L.

1996-01-01

Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500 degrees C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO 2 matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO 2 matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs

Microstructure in Zircaloy Creep Tested in the R2 Reactor

International Nuclear Information System (INIS)

Pettersson, Kjell

2004-12-01

Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

Digital mock-up for the spent fuel disassembly processes

International Nuclear Information System (INIS)

Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

2000-12-01

In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

R and D status of an integral type small reactor MRX in JAERI

International Nuclear Information System (INIS)

Hoshi, Tsutao; Ochiai, Masaaki; Iida, Hiromasa; Yamaji, Akio; Shimazaki, Junya

1995-01-01

JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety study of the thermal hydraulic phenomena as well as the feasibility study of the applicability to merchant ships. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the results of feasibility study. (author)

Nuclear reactor fuel element with a cluster of parallel fuel pins

International Nuclear Information System (INIS)

Macfall, D.; Butterfield, C.E.; Butterfield, R.S.

1977-01-01

An improvement of the design of nuclear reactor fuel elements is described and illustrated by the example of a gas-cooled, graphite-moderated nuclear reactor. The fuel element has a cluster of parallel fuel pins with an outer can of structure material and an inner sleeve, as well as tie bars and spacing devices for all of these parts. The fuel element designed according to the invention allows lasy assembling and disassembling before and after use. During use, no relative axial motions are possible; nevertheless, the graphite sleeve is at no time subject to tensile stress: the individual parts are held in position from below by a single holding device. (UWI) [de

An alternative proposal for the disposal of potentially contaminated railroad crossties at the Savannah River Site. Revision 1

International Nuclear Information System (INIS)

Hochel, R.C.

1996-06-01

The Savannah River Site (SRS) has accumulated approximately 300,000 crossties that have been replaced over a 40-year period of maintaining the site's 62 miles of railroad track. The ties reside in a pile on an open area of 2.3 acres near the site's F-Area facilities. A small fraction of the ties are potentially contaminated with radioactivity as a result of past site practices and service. Contamination was possible from occasional leaks in transport-casks moved by rail from the site's five nuclear materials production reactors to its two reprocessing facilities. Casks typically were filled with spent fuel, targets, and water from the reactor disassembly basins, which contained small amounts of fission and neutron activation product radioactivity normal to the operation of nuclear reactors

Applications of neutron activation analysis technique in the IPR-R1 research reactor

International Nuclear Information System (INIS)

Sabino, C.V.S.; Mansur, N.

1986-01-01

A review is made of the neutron activation analysis technique used in the IPR-R1 reactor of the Centro de Desenvolvimento da Tecnologia Nuclear - NUCLEBRAS. Some characteristics of the method are described, types of samples and elements analyzed are also mentioned. (Author) [pt

Present status and future perspective of R and D on lead heavy metal-cooled fast reactors

International Nuclear Information System (INIS)

Takahashi, Minoru

2007-01-01

Since a lead heavy metal (lead-bismuth eutectic) is chemically inert and has higher boiling point compared to a sodium, a lead heavy metal-cooled fast reactor can be inherently safe and has good nuclear characteristics and is so suitable to a medium-small size of the reactor. R and D on corrosion of a lead heavy metal has been carried out in the world and this issue might be solved to choose specific corrosion resistant alloys for structural materials and fuel cans of a lead heavy metal-cooled reactor. This article reviews present status and future perspective on lead heavy metal-cooled fast reactors. (T. Tanaka)

Commissioning of the new heat exchanger for the research nuclear reactor IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo Jose Alvim de; Cassiano, Douglas Alves; Umbehaun, Pedro Ernesto; Carvalho, Marcos Rodrigues de; Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ajcastro@ipen.br; docass@gmail.com.br; umbehaun@ipen.br; carvalho@ipen.br; frajndli@ipen.br

2008-07-01

The Research Reactor IEA-R1 placed at IPEN/CNEN-SP is of the swimming pool type, light water moderated and with graphite reflectors, and was build and designed by Babcock and Wilcox Co. Start up operation was in September the 16{sup th}, 1957, being the first criticality for South Hemisphere. Although designed to operate at 5 MW, the IEA-R1 was operated until 2001 with 2 MW and was suitable for use in basic and applied research as well as the production of medical radioisotopes, industry and natural sciences applications. Due to a recent demand increase on radioisotopes in Brazil for medical diagnoses and therapies applications, IPEN /CNEN updated the IEA-R1 power to 5 MW and to work at continuous operation regime. Studies on the Ageing Management for the Research Reactor IEA-R1 were conducted according to IAEA procedures. As result of these studies critical components within the Ageing Management Program were identified. Also were made recommendations on the implementation of test scheduling and standardization procedures to organize data and documents. One of the main results was the need of monitoring the two heat exchangers, the two primary circuit pumps and the data acquisition system. During monitoring procedures, issues were observed on the IEA-R1 operation at 5 MW mainly due to the ageing of the Babcox and Wilcox TCA heat exchanger, and excessive vibrations at high flow rates on CBC's TCB heat exchanger. So, from 2005 on, it was decided to work with 3,5 MW and provide a new IESA heat exchanger with 5 MW capacity, to substitute the TCA heat exchanger. This work presents results on the commissioning of the new heat exchanger and compares against the values calculated in the IESA project. The results show that the IEA-R1 Reactor can be operated more safety and continuously at 5 MW with the new IESA heat exchanger. (author)

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

9th Pacific Basin Nuclear Conference. Nuclear energy, science and technology - Pacific partnership. Proceedings Volume 1

International Nuclear Information System (INIS)

1994-04-01

The theme of the 9th Pacific Basin Nuclear conference held in Sydney from 1-6 May 1994, embraced the use of the atom in energy production and in science and technology. The focus was on selected topics of current and ongoing interest to countries around the Pacific Basin. The two-volume proceedings include both invited and contributed papers. They have been indexed separately. This document, Volume 1 covers the following topics: Pacific partnership; perspectives on nuclear energy, science and technology in Pacific Basin countries; nuclear energy and sustainable development; economics of the power reactors; new power reactor projects; power reactor technology; advanced reactors; radioisotope and radiation technology; biomedical applications

Artificial biofilms establish the role of matrix interactions in staphylococcal biofilm assembly and disassembly

Science.gov (United States)

Stewart, Elizabeth J.; Ganesan, Mahesh; Younger, John G.; Solomon, Michael J.

2015-01-01

We demonstrate that the microstructural and mechanical properties of bacterial biofilms can be created through colloidal self-assembly of cells and polymers, and thereby link the complex material properties of biofilms to well understood colloidal and polymeric behaviors. This finding is applied to soften and disassemble staphylococcal biofilms through pH changes. Bacterial biofilms are viscoelastic, structured communities of cells encapsulated in an extracellular polymeric substance (EPS) comprised of polysaccharides, proteins, and DNA. Although the identity and abundance of EPS macromolecules are known, how these matrix materials interact with themselves and bacterial cells to generate biofilm morphology and mechanics is not understood. Here, we find that the colloidal self-assembly of Staphylococcus epidermidis RP62A cells and polysaccharides into viscoelastic biofilms is driven by thermodynamic phase instability of EPS. pH conditions that induce phase instability of chitosan produce artificial S. epidermidis biofilms whose mechanics match natural S. epidermidis biofilms. Furthermore, pH-induced solubilization of the matrix triggers disassembly in both artificial and natural S. epidermidis biofilms. This pH-induced disassembly occurs in biofilms formed by five additional staphylococcal strains, including three clinical isolates. Our findings suggest that colloidal self-assembly of cells and matrix polymers produces biofilm viscoelasticity and that biofilm control strategies can exploit this mechanism. PMID:26272750

Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Zacarias

2005-07-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

Méthodologie de l'extrapolation des réacteurs chimiques Methodology for Scaling Up Chemical Reactors

Directory of Open Access Journals (Sweden)

Trambouze P.

2006-11-01

Full Text Available Après un exposé général relatif à la méthodologie du développement des procédés, applicable à l'extrapolation des réacteurs, est présenté un rapide examen critique des deux principales techniques mises en oeuvre, à savoir : - la théorie de la similitude ; - l'élaboration de modèles mathématiques. Deux exemples pratiques, relatifs aux réacteurs homogènes et aux réacteurs catalytiques à lit fixe et deux phases fluides, sont ensuite examinés à la lumière des considérations générales précédentes. After giving a general description of process-development methodology applicable to scaling up reactors, this article makes a quick critical examination of the two main techniques involved, i. e. : (a the theory of similarity, and (b the compiling of mathematical models. Two practical examples relating to homogeneous reactors and trickle-bed catalytic reactors are then examined in the light of the preceding general considerations.

Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

International Nuclear Information System (INIS)

Moreira, Joao Manoel Losada; Frajndlich, Roberto

1995-01-01

This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

Nano-functionalized filamentous fungus hyphae with fast reversible macroscopic assembly & disassembly features.

Science.gov (United States)

Wang, Haiying; Li, Xiaorui; Chai, Liyuan; Zhang, Liyuan

2015-05-18

A uniform decoration of hyphae by polyaniline nanoparticles (PANI NPs) was achieved here. This novel hybrid structure can be effectively assembled into a film by filtration and disassembled in water by shaking. This reversible process is very fast, which promises applications in nanomaterials including adsorption.

25th birthday of the first criticality of IPR-R1 nuclear research reactor

International Nuclear Information System (INIS)

Tofani, P.C.; Stasiulevicius, R.; Roedel, G.

1988-01-01

The historical evolution of IPR-R1 research reactor of Instituto de Pesquisas Radioativas-Nuclebras, since the data of its first criticality, is presented. The modifications and the main activities carried out, are presented. (M.C.K.) [pt

Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

International Nuclear Information System (INIS)

Castiblanco, L.A.; Sarta, J.A.

1997-01-01

The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

International Nuclear Information System (INIS)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

2014-10-01

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

International Nuclear Information System (INIS)

Ricardo, Carla Pereira; Mesquita, Amir Zacarias

2007-01-01

During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

Application of safety checklist to the analysis of the IEA-R1 reactor water retreatment system

International Nuclear Information System (INIS)

Sauer, Maria Eugenia Lago Jacques; Sara Neto, Antonio Jorge; Lima, Toni Carlos Caboclo de; Ribeiro, Maria Alice Morato

2005-01-01

In 1999, the management of the IEA-R1 Research Reactor (pool type - 5 MWth), located at IPEN/CNEN-SP, started the evaluation of the Reactor Pool Water Retreatment System to identify operational aspects, which could compromise the operators safety. The purpose was to identify and propose enhancements to the system which would be installed to substitute for the existing one. This process was conducted through a qualitative study of the system in operation. This study was carried out by a team composed of specialists in reactor operation, systems maintenance and radiological protection, and one safety analyst. The study consisted, basically, in local inspections to verify the physical and operational conditions of each equipment / component as well as aspects related to maintenance activities of the system. The process control and the operator procedures associated with the retreatment of the reactor pool water were also reviewed. The methodology adopted to develop the study was based in process hazard analysis technique named Safety Checklist. This paper presents a summary of this study and the main results obtained. Some operational and safety problems identified, the prevention and/or correction means to avoid them, and the recommendations and suggestions that have been implemented to the new design of the IEA-R1 Reactor Water Retreatment System, whose installation was concluded in 2003, are also presented. (author)

Actin-interacting Protein 1 Promotes Disassembly of Actin-depolymerizing Factor/Cofilin-bound Actin Filaments in a pH-dependent Manner.

Science.gov (United States)

Nomura, Kazumi; Hayakawa, Kimihide; Tatsumi, Hitoshi; Ono, Shoichiro

2016-03-04

Actin-interacting protein 1 (AIP1) is a conserved WD repeat protein that promotes disassembly of actin filaments when actin-depolymerizing factor (ADF)/cofilin is present. Although AIP1 is known to be essential for a number of cellular events involving dynamic rearrangement of the actin cytoskeleton, the regulatory mechanism of the function of AIP1 is unknown. In this study, we report that two AIP1 isoforms from the nematode Caenorhabditis elegans, known as UNC-78 and AIPL-1, are pH-sensitive in enhancement of actin filament disassembly. Both AIP1 isoforms only weakly enhance disassembly of ADF/cofilin-bound actin filaments at an acidic pH but show stronger disassembly activity at neutral and basic pH values. However, a severing-defective mutant of UNC-78 shows pH-insensitive binding to ADF/cofilin-decorated actin filaments, suggesting that the process of filament severing or disassembly, but not filament binding, is pH-dependent. His-60 of AIP1 is located near the predicted binding surface for the ADF/cofilin-actin complex, and an H60K mutation of AIP1 partially impairs its pH sensitivity, suggesting that His-60 is involved in the pH sensor for AIP1. These biochemical results suggest that pH-dependent changes in AIP1 activity might be a novel regulatory mechanism of actin filament dynamics. © 2016 by The American Society for Biochemistry and Molecular Biology, Inc.

The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

1999-01-01

The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

The Santos Basin Ocean Observing System: From R&D to Operational Regional Forecasts

Science.gov (United States)

Da Rocha Fragoso, M.; Moore, A. M.; dos Santos, F. A.; Marques Da Cruz, L.; Carvalho, G. V.; Soares, F.

2016-02-01

Santos Basin is located on the Southwestern Brazilian Ocean Basin and comprises the main offshore oil reserves of Brazil. The exploration and production activities on its ocean are growing in accelerated pace, which means that oil spill contingency and search & rescue operations are likely to be more frequent. Therefore, ocean current reliable nowcasts and forecasts has become even more important for this region. The Santos Basin Ocean Observing System was designed as an R&D project and its main objective was to establish and maintain a systematic oceanographic data collection for this region in order to study its ocean dynamics and improve regional ocean forecast through data assimilation. In the first three years of the project surface drifters, profiling floats and gliders were deployed to measure and monitor mainly the Brazil Current Western Boundary System, a highly unstable baroclinic current system, that present several meanders and mesoscale eddies activities. Throughout the development of the project, the team involved was able to learn how to operate the equipment, treat the collected data and use it to assimilate on the Regional Ocean Modeling System (ROMS). After performing a one-year 4DVAR assimilation cycle (Fragoso et al., 2015) in which the forecasting skill was assessed, the system was considered mature enough to start producing ocean circulation forecasts for Santos Basin. It is the first time in Brazil that a regional ocean model using a 4DVAR data assimilation scheme was used to produce high resolution operational ocean current forecasts. This paper describes all the components of this forecasting system, its main results and discoveries with special focus on the Brazil Current System Transport and mesocale eddies dynamics and statistics.

77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

Science.gov (United States)

2012-02-13

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-264; NRC-2012-0026] Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108 AGENCY: Nuclear Regulatory Commission... Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow Chemical Company...

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

2009-01-01

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

Analysis of fuel sodium interaction in a fast breeder reactor

International Nuclear Information System (INIS)

Tezuka, M.; Suzuki, K.; Sasanuma, K.; Nagasima, K.; Kawaguchi, O.

A code ''SUGAR'' has been developed to evaluate molten Fuel Sodium Interaction (FSI) in a fast breeder reactor. This code computes thermohydrodynamic behavior by heat transfer from fuel to sodium and dynamic deformation of reactor structures simultaneously. It was applied to evaluate FSI in local fuel melting accident in a fuel assembly and in core disassembly accident for the 300MWe fast breeder reactor under development in Japan. The analytical methods of the SUGAR code are mainly shown in the following: 1) the thermal and dynamic model of FSI is mainly based on Cho-Wright's model; 2) the axial and radial expansions of surroundings of FSI region are calculated with one-dimensional and compressive hydrodynamics equation; 3) the structure response is calculated with one-dimensional and dynamic stress equation. Our studies show that mass of fuel interacted with sodium, ratio of fuel mass to sodium mass, fuel particle size, heat transfer coefficient from fuel to sodium, and structure's force have great effect on pressure amplitude and deformation of reactor structures

Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.

1989-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Economics of water basin storage of spent light water reactor fuel

International Nuclear Information System (INIS)

Driggers, F.E.

1978-01-01

As part of the International Spent Fuel Storage program, a preliminary Venture Guidance Assessment of the cost was made. The escalated cost of a reference facility with a capacity to receive 2000 MT/y of spent LWR fuel and to store 5000 MT in water-filled pools was converted to $180 million in 1978 dollars for a stand-alone facility. It was estimated that the receiving rate could be increased to 3000 MT/y for an additional $15 million and that increments could be added to the storage capacity for $13 million per 1000 MT. If a receipt rate of more than 3000 MT/y is required, a new facility in another part of the country might be built to reduce total costs including transportation. Operating costs are determined by the number of people employed and by the costs of stainless steel baskets. An operating crew of 150 is required for the reference facility; the associated cost, including overhead and supplies, is $6 million. During an extended storage-only period, this cost is assumed to drop to $4 million. Fuel baskets are estimated to cost $6.20/kg of spent fuel averaged over a reactor mix of two-thirds PWRs and one-third BWRs. The nominal basket requirements of $10 million for the first year are capitalized. If the facility is financed by the government and a one-time fee is charged to recover all of the away-from-reactor (AFR) basin costs, the fee is about $60/kg of spent fuel plus any government surcharge to cover research and development, overhead, and additional contingencies. If the facility is financed by industry with an annual charge that includes a fixed charge on capital of 25%, the annual fee is about $16/kg-y. In calculating both fees, it is assumed that each storage position is occupied for ten years. 8 tables

78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

Science.gov (United States)

2013-01-28

... University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating License No. R-115, for the University of Illinois... Operating License No. R-115 is terminated. The above referenced documents may be examined, and/or copied for...

Remote handling requirements and considerations for D-T fusion reactors

International Nuclear Information System (INIS)

Spampinato, P.T.

1984-01-01

This paper presents an overview of the maintenance considerations for next-generation fusion reactors. It draws upon the work done at the Fusion Engineering Design Center over the past several years in the conceptual development of tokamaks and tandem mirrors. It specifically addresses the maintenance philosophy adopted for these devices, the configuration development using a modular design approach, scheduled and unscheduled maintenance operations, assembly and disassembly scenarios for component replacements, maintenance equipment requirements, and the operating availability of these devices. In addition, recent work on the development of a totally remote tokamak configuration is presented

Remote handling requirements and considerations for D-T fusion reactors

International Nuclear Information System (INIS)

Spampinato, P.T.

1984-01-01

This paper an overview of the maintenance considerations of next-generation fusion reactors. It draws upon the work done at the Fusion Engineering Design Center over the past several years in the conceptual development of tokamaks and tandem mirrors. It specifically addresses the maintenance philosophy adopted for these devices, the configuration development using a modular design approach, scheduled and unscheduled maintenance operations, assembly and disassembly scenarios for component replacements, maintenance equipment requirements, and the operating availability of these devices. In addition, recent work on the development of a totally remote tokamak configuration is presented

International R and D project on development of coated particle fuel for innovative reactors

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

The paper presents an outline for an international collaborative project of coated particle fuel development for innovative reactors. Specific issues include identification of R and D needs and the Member State facilities for meeting the needs followed by development and demonstration of technology. (author)

Fast reactor fuel reprocessing plant D1206: disassembly cave window 4 replacement

International Nuclear Information System (INIS)

Sutherland, H.G.; Beckitt, S.; Potts, A.B.

1996-01-01

At UKAEA's fast reactor reprocessing plant at Dounreay, the containment glass on the zinc bromide cave viewing window tank failed after 13 years active use. External shielding was fitted and the window tank subsequently drained to make it safe. Fuel cropping operations carried out behind the window were resited to enable cave work to continue whilst a project team made arrangements and plans to replace the damaged window. Because of the complexity of the task and high (alpha, beta, gamma and neutron) radiation levels in excess of 500 Sv/hr a rehearsal facility was built to develop the remote handling techniques to be employed in the task. (UK)

Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

International Nuclear Information System (INIS)

Sculthorpe, B.R.; Johnson, K.M.

1986-01-01

With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

Early cysteine-dependent inactivation of 26S proteasomes does not involve particle disassembly

Directory of Open Access Journals (Sweden)

Martín Hugo

2018-06-01

Full Text Available Under oxidative stress 26S proteasomes suffer reversible disassembly into its 20S and 19S subunits, a process mediated by HSP70. This inhibits the degradation of polyubiquitinated proteins by the 26S proteasome and allows the degradation of oxidized proteins by a free 20S proteasome. Low fluxes of antimycin A-stimulated ROS production caused dimerization of mitochondrial peroxiredoxin 3 and cytosolic peroxiredoxin 2, but not peroxiredoxin overoxidation and overall oxidation of cellular protein thiols. This moderate redox imbalance was sufficient to inhibit the ATP stimulation of 26S proteasome activity. This process was dependent on reversible cysteine oxidation. Moreover, our results show that this early inhibition of ATP stimulation occurs previous to particle disassembly, indicating an intermediate step during the redox regulation of the 26S proteasome with special relevance under redox signaling rather than oxidative stress conditions.

Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

International Nuclear Information System (INIS)

Bittelli, U.D.

1988-01-01

This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

The United States pit disassembly and conversion project -- Meeting the MOX fuel specification

International Nuclear Information System (INIS)

Nelson, T.O.; James, C.A.; Kolman, D.G.

1998-01-01

The US is actively involved in demonstrating the disassembly of nuclear weapons pits to an unclassified form readied for disposition. The MOX option is the most likely path forward for plutonium that originated from nuclear weapon pits. The US demonstration line for pit disassembly and conversion is known as ARIES, the advanced recovery and integrated extraction system. The ARIES demonstration line is being used to gather data in an integrated fashion of the technologies needed for pit disassembly and conversion. These activities include the following modules: pit bisection, hydride-dehydride, oxide conversion, canning, electrolytic decontamination, and nondestructive assay (NDA). Pit bisection swages in a pit in half. Hydride-dehydride converts the pit plutonium metal to an unclassified metal button. To convert the plutonium metal to an oxide the US is investigating a number of options. The primary oxide conversion approach involves variations of combining plutonium hydriding and subsequent oxidation. Another approach is to simply oxidize the metal under controlled conditions-direct metal oxidation (DMO). To remove the gallium from the plutonium oxide, a thermal distillation approach is being used. These pyrochemical approaches will substantially reduce the wastes produced for oxide conversion of weapon plutonium, compared to traditional aqueous processing. The packaging of either the plutonium metal or oxide to long term storage criteria involves the canning and electrolytic decontamination modules. The NDA suite of instruments is then used to assay the material in the containers, which enables international verification without the need to open the containers and repackage them. All of these processes are described

Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

International Nuclear Information System (INIS)

Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

1987-01-01

The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

Pilot scale, alpha disassembly and decontamination facility at the Savannah River Laboratory

International Nuclear Information System (INIS)

Cadieux, J.R.; Becker, G.W. Jr.; Richardson, G.W.; Coogler, A.L.

1982-01-01

An alpha-contained pilot facility is being built at the Savannah River Laboratory (SRL) for research into the disassembly and dcontamination of noncombustible, Transuranic (TRU) waste. The design and program objectives for the facility are presented along with the initial test results from laboratory scale decontamination experiments with Pu-238 and Cm-244

M-area basin closure-Savannah River Site

International Nuclear Information System (INIS)

McMullin, S.R.; Horvath, J.G.

1991-01-01

M-Area, on the Savannah River Site, processes raw materials and manufactures fuel and target rods for reactor use. Effluent from these processes were discharged into the M-Area settling basin and Lost Lake, a natural wetland. The closure of this basin began in 1988 and included the removal and stabilization of basin fluids, excavation of all contaminated soils from affected areas and Lost Lake, and placement of all materials in the bottom of the emptied basin. These materials were covered with a RCRA style cap, employing redundant barriers of kaolin clay and geosynthetic material. Restoration of excavated uplands and wetlands is currently underway

Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

International Nuclear Information System (INIS)

Von Lensa, W.

1998-01-01

The nuclear industries and the nuclear research and development (R and D) programmes world-wide have undergone considerable changes over recent years which have resulted in the formation of international industrial consortiums on the one hand and the need for synergistic collaboration in the R and D area due to the reductions of national R and D activities in the nuclear field on the other hand. International networking starting from precompetitive medium- or long-term oriented R and D could be an efficient mean to overcome the problems nuclear energy is facing today with respect to the lack of public acceptance and economic attractivity in a joint effort. Additional motivation is provided by the fact that there is not only a globalisation of markets but also a 'globalisation of problems' to be addressed internationally like reductions of environmental impacts and long-term availability of economic energy supply. The tools for telecommunication and telecollaboration are evolving in parallel and offer better conditions for closer collaboration of different R and D teams at distant locations than ever before. It is obvious that these trends and boundary conditions will drastically influence the structures of collaboration not only in the industries, but for R and D on an international level, too. The chances emerging from the creation of a European Union and from the globalisation trends have to be converted into strategic benefits by active response on these 'historic changes'. New initiatives have been undertaken in Europe to push for innovations of nuclear reactor technologies via international R and D Networks under the European R and D Framework Programmes (FWP). Innovative approaches are already addressed with limited funding under the actual 4th FWP and should be extended for complementing the commercial efforts on evolutionary LWR concepts by medium- and long-term oriented innovations and R and D. The MICHELANGELO initiative as well as the EU-funded Concerted

Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

International Nuclear Information System (INIS)

Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

2014-01-01

This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

Experience with the RE fuel transition at the Studsvik R2 reactor

International Nuclear Information System (INIS)

Pazsit, I.; Saltvedt, K.

1991-01-01

Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

Spent LWR fuel-storage costs

International Nuclear Information System (INIS)

Clark, H.J.

1981-01-01

Expanded use of existing storage basins is clearly the most economic solution to the spent fuel storage problem. The use of high-density racks followed by fuel disassembly and rod storage is an order of magnitude cheaper than building new facilities adjacent to the reactor. The choice of a new storage facility is not as obvious; however, if the timing of expenditures and risk allowance are to be considered, then modular concepts such as silos, drywells, and storage casks may cost less than water basins and air-cooled vaults. A comparison of the costs of the various storage techniques without allowances for timing or risk is shown. The impact of allowances for discounting and early resumption of reprocessing is also shown. Economics is not the only issue to be considered in selecting a storage facility. The licensing, environmental impact, timing, and social responses must also be considered. Each utility must assess all of these issues for their particular reactors before the best storage solution can be selected

Department of Energy HEU ES and H vulnerability assessment, Savannah River Site, Site Assessment Team report. Revision 2

International Nuclear Information System (INIS)

Geddes, R.L.; Barone, A.; Shook, H.E. Varner, C.E.; Rollins, R.

1996-01-01

This report fulfills the directive issued by the Secretary of Energy on February 22, 1996 to complete a comprehensive assessment of potential vulnerabilities associated with the management of highly enriched uranium (HEU) throughout the DOE complex. In a subsequent letter instruction, the DOE-SR Field Office formally directed WSRC to conduct an assessment of the HEU materials at SRS. The term ''ES and H vulnerabilities'' is defined for the purpose of this assessment to mean conditions or weaknesses that could lead to unnecessary or increased exposure of workers or the public to radiation or to HEU-associated chemical hazards, or to the release of radioactive materials to the environment. The assessment will identify and prioritize ES and H vulnerabilities, and will serve as an information base for identifying corrective actions for the safe management of HEU. Primary facilities that hold HEU at SRS are H-Canyon, K-Reactor assembly area, K, L, and P-Reactor disassembly basins, and the Receiving Basin for Offsite Fuels (RBOF)

In situ characterization of Hanford K Basins fuel

Energy Technology Data Exchange (ETDEWEB)

Pitner, A.L.

1998-01-06

Irradiated N Reactor uranium metal fuel is stored underwater in the Hanford K East and K West Basins. In K East Basin, fuel is stored in open canisters and defected fuel is free to react with the basin water. In K West Basin, the fuel is stored in sealed canisters filled with water containing a corrosion inhibitor (potassium nitrite). To gain a better understanding of the physical condition of the fuel in these basins, visual surveys using high resolution underwater cameras were conducted. The inspections included detailed lift and look examinations of a number of fuel assemblies from selected canisters in each basin. These examinations formed the bases for selecting specific fuel elements for laboratory testing and analyses as prescribed in the characterization plan for Hanford K Basin Spent Nuclear Fuel.

Disassembled DJ-1 high molecular weight complex in cortex mitochondria from Parkinson's disease patients

Directory of Open Access Journals (Sweden)

Adler Charles

2009-07-01

Full Text Available Abstract Correction to Nural H, He P, Beach T, Sue L, Xia W, Shen Y. Disassembled DJ-1 high molecular weight complex in cortex mitochondria from Parkinson's disease patients Molecular Neurodegeneration 2009, 4:23.

Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

International Nuclear Information System (INIS)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A.

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U 3 O 8 and U 3 Si 2 dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U 3 O 8 -Al and five containing U 3 Si 2 -Al), with densities of 3.2 gU/cm 3 and 4.8 gU/cm 3 respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

International Nuclear Information System (INIS)

Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

1994-01-01

ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

Review of the 6th Pacific Basin Nuclear Conference

International Nuclear Information System (INIS)

Zhao Renkai; Jiang Yue; Yao Minzhi; Tong Yunxian; Ruan Keqiang; Luo Cheng; Liu Dingqin; Yang Chuande; Tu Zhuguo.

1987-01-01

This paper reviews the 6th Pacific Basin Nuclear Conference on eight monographs. These monographs are present status and prospects of nucliar energy, advancement of commercial power reactor, advanced nuclear reactor and research and test reactor, perspectives on nuclear power safety in 1990's, application of radioisotope and irradiation technique, advancement in nuclear fuel, improvement of nuclear power plant management and regional cooperation and personnel training

Safety analysis for K reactor and impact of cooling tower installation

International Nuclear Information System (INIS)

Fields, C.C.; Wooten, L.A.; Geeting, M.W.; Morgan, C.E.; Buczek, J.A.; Smith, D.C.

1993-01-01

This paper describes the safety analysis of the Savannah River site K-reactor loss-of-cooling-water-supply (LOCWS) event and the impact on the analysis of a natural-draft cooling tower, which was installed in 1992. Historically (1954 to 1992), the K-reactor secondary cooling system [called the cooling water system (CWS)] used water from the Savannah River pumped to a 25-million-gal basin adjacent to the reactor. Approximately 170 000 gal/min were pumped from the basin through heat exchangers to remove heat from the primary cooling system. This water then entered a smaller basin, where it flowed over a weir and eventually returned to the Savannah River. The 25-million-gal basin is at a higher elevation than the heat exchangers and the smaller basin to supply cooling by gravity flow (which is sufficient to remove decay heat) if power to the CWS pumps is interrupted. Small amounts of cooling water are also used for other essential equipment such as diesels, motors, and oil coolers. With the cooling tower installed, ∼85% of the cooling water flows from the small basin by gravity to the cooling tower instead of returning to the Savannah River. After being cooled, it is pumped back to the 25-million-gal basin. River water is supplied only to make up for evaporation and the blowdown stream

Development of disassembly and pin chopping technology for FBR spent fuels

International Nuclear Information System (INIS)

Kobayashi, Tsuguyuki; Namba, Takashi; Kawabe, Yukinari; Washiya, Tadahiro

2008-01-01

Japan Atomic Power Company (JAPC) and Japan Atomic Energy Agency (JAEA) have been developing fuel disassembly and fuel pin chopping systems for a future Japanese commercial FBR. At first, the wrapper tube is cut by the slit-cut to pull it out, then the fuel pins are cut by the crop-cut at their end-plugs to separate them from the entrance nozzle. The pins are transferred to the magazine of the chopping machine. A series of tests were performed to develop this procedure. As the result of mechanical cutting tests, the CBN wheel was selected. The slit-cut tests were carried out to evaluated the cutting performance of the wheel. The wrapper tube is normally slit-cut in the circumferential direction. One CBN wheel could cut more than 5 fuel assemblies in this direction. The slit-cut in the axial direction is prepared as provision when the tube is difficult to put out. More work is needed to cut 5mm thick PNC-FMS plate in this direction without damaging the pins beneath it. As the result of the crop-cut tests of end-plugs made of ODS steel, the CBN wheel could cut the 61 pin bundle by two strokes. More work is needed to cut the 217 pin bundle. Fuel pin handling tests were performed to transfer them from the disassembly machine to the chopping machine. The Saucer tray was selected to receive the disassembled pins. All the pins were transferred and loaded into a magazine of the chopping machine. Fuel pin loading tests were conducted to optimize the magazine configuration to make the chopping length within 1.0±0.5 cm. In order to decrease the disturbance during chopping, the width of the magazine was adjusted to be 12 cm and installation of a height adjuster is favourable to control the free space above the pins. (author)

Subproject plan for demonstration of 3M technology for treatment of N Basin water

International Nuclear Information System (INIS)

Plastino, J.C.

1996-02-01

A dissolved radionuclides removal demonstration is being conducted at the 105-N Basin as part of the 100-N Area Projects' policy of aggressively integrating innovative technologies to achieve more cost effective, faster, and/or safer deactivation operations. This subproject plan demonstrates new technology (marketed by the 3M trademark Company) that absorbs specific ions from water. The demonstration will take place at the spent fuel basin at the N Reactor facility. The 105-N Basin contains 1 million gal of water consisting of approximately 32 Ci of dissolved 90 Sr at a concentration of 8.4 uCi/L and 7.3 Ci of dissolved 137 Cs at a concentration of 1.92 uCi/L. The Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement [Ecology et al. 1990]) Milestone M-16-01E-T2 requires the initiation of pretreatment and removal of all N Reactor fuel storage basin waters by September 30, 1996, pursuant to the N Reactor Deactivation Program Plan (WHC 1993). 105-N Basin dewatering is on the critical path for overall deactivation of N Reactor by March 1997. The 105-N Basin Deactivation Program Plan (BHI 1995) includes removing debris, hardware, algae and sediment from the basin, followed by pretreatment (filtration) and removal of the 1005-N Basin water. Final water removal is currently scheduled for September 30, 1996. The recommended method of the 105-N Basin water is the treatment of the water at the Effluent Treatment Facility (ETF) in the 200 East Area. The demonstration of the 3M technology could be a feasible treatment alternative to the ETF if the ETF is not available to meet the project schedule or if additional pretreatment is needed to reduce the inventory of radioactive species to be handled at the ETF. Demonstration of this technology could be of value for other fuel basins at the Hanford Site and possibly other US Department of Energy (DOE) sites and non- DOE nuclear power plants

Current status of the Thai Research Reactor (TRR-1/M1)

International Nuclear Information System (INIS)

Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

2006-01-01

The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

The Effect of Uncertain End-of-Life Product Quality and Consumer Incentives on Partial Disassembly Sequencing in Value Recovery Operations

OpenAIRE

Rickli, Jeremy Lewis

2013-01-01

This dissertation addresses gaps in the interaction between End-of-Life (EoL) product acquisition systems and disassembly sequencing. The research focuses on two remanufacturing research problems; 1) modeling uncertain EoL product quality, quantity, and timing in regards to EoL product acquisition and disassembly sequencing and 2) designing EoL product acquisition schemes considering EoL product uncertainty. The main research objectives within these areas are; analyzing, predicting, and contr...

Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

2007-03-30

The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

International Nuclear Information System (INIS)

Harvey, R.S.

1983-01-01

An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

Development of an advanced, continuous mild-gasification process for the production of coproducts. Report for Task 4.8, Decontamination and disassembly of the mild-gasification and char-to-carbon PRUs and disposal of products from testing

Energy Technology Data Exchange (ETDEWEB)

Merriam, N.W. [Western Research Inst., Laramie, WY (United States); Jha, Mahesh C. [AMAX Research and Development Center, Golden, CO (United States)

1991-11-01

This report contains descriptions of mild-gasification and char-to-carbon process research units (PRUS) used by WRI and AMAX R&D Center to conduct tests under contract AC21-87MC24268. Descriptions of materials produced during those tests are also contained herein. Western Research Institute proposes to dispose of remaining fines and dried coal by combustion and remaining coal liquids by incineration during mid-1992. The mild-gasification PRU will be used for additional tests until 1993, at which time WRI proposes to decontaminate and disassemble the PRU. AMAX R&D Center intends to return the spent char, any remaining feed char, and unusable product carbon to the Eagle Butte Mine near Gillette, Wyoming, from where the coal originally came. The solid products will be added to the mine`s coal product stream. Coal liquids collected from condensers will be concentrated and sent to a local oil and solvent recycling company where the liquids will be burned as fuel. The char-to-carbon PRU will be operated periodically until 1993 when the plant will be decontaminated and disassembled.

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Rodents from the Upper Miocene Tuğlu Formation (Çankırı Basin, Central Anatolia, Turkey)

NARCIS (Netherlands)

Joniak, Peter; de Bruijn, Hans

2015-01-01

The upper Miocene assemblages of rodents collected from two layers of the type section of the Tuğlu Formation (Çankırı Basin, Central Anatolia, Turkey) are described. The assemblage from the lower level is considerably less diverse than that from the upper level. It contains Progonomys together with

Replacement of thermal column elastomeric gasket in pool type research reactors based on ageing and radiation degradation

International Nuclear Information System (INIS)

Garai, S.K.

2006-01-01

Pool type research reactors are designed with Thermal column facilities to irradiate samples at different flux levels of thermal neutrons. The sealing of demineralised pool water between stainless steel lined pool wall and the Aluminium Thermal column plate is achieved by an elastomeric gasket. The gasket joint is subjected to pool water temperature ranging from 25degC to 45degC and radiation field of the order of 104 -106 R/hr. The gasket loses its sealing properties due to ageing and radiation degradation after a few years, leading to the leakage and loss of the pool water. Though degradation of the gasket is, generally, predictable, some amount of uncertainty always remains in the leakage rate. The paper describes the study of a few elastomers in radiation environment and replacement of the Thermal column gasket of a swimming pool type research reactor. It includes the details of features like planning and scheduling, the actual sequential execution of the job, various problems encountered and corrective measures applied, engineering and radiological safety measures adopted, development of remote tools, disassembly and reassembly procedure and finally satisfactory completion of the site job in high radiation environment with minimum time and man rem consumption. (author)

Severe Accident R and D for Enhanced CANDU-6 Reactors

International Nuclear Information System (INIS)

Nitheanandan, Thambiayah

2012-01-01

CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

Prey and mound disassembly, manipulation and transport by fire ant collectives

Science.gov (United States)

Dutta, Bahnisikha; Monaenkova, Daria; Goodisman, Michael A.; Goldman, Daniel

Fire ants inhabit subterranean nests covered by a hemispherical mound of soil permeated by narrow ( 1 body length diameter) tunnels. Fire ants can use their mound for long-term food storage [Gayahan &Tschinkel, J. Insect Sci.,2008]. Since mound tunnels are narrow, we expect that in addition to prey manipulation, mound reconfiguration could also be an important aspect of the food storage strategy. Ant colonies collected from wild were allowed to build nests in containers filled with clay soil in the laboratory. These colonies were offered diverse prey embedded with lead markers, including mealworms, crickets and shrimp. Ant-prey-soil interactions on the nest surface were recorded using overhead video and subsurface using x-ray imaging. Individual ants involved in prey storage exhibited three distinct behaviors: prey maneuvering, prey dissection and mound reconfiguration. Small prey (e.g. mealworms) were collectively carried intact into the mound through a tunnel, and then disassembled within the mound. Larger prey (e.g. shrimp) were dismantled into small pieces above the surface and carried to mound tunnels. The bodies of hard medium-sized prey (e.g. crickets) were buried after limb removal and then disassembled and moved into tunnels. Soil reconfiguration occurred in all cases.

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

Energy Technology Data Exchange (ETDEWEB)

Klann, R. T.; Perret, G.; Nuclear Engineering Division

2007-10-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

Phylogeography of the Neotropical catfish Pimelodus albicans (Siluriformes: Pimelodidae from río de la Plata basin, South America, and conservation remarks

Directory of Open Access Journals (Sweden)

Julia Vergara

Full Text Available Pimelodus albicans Valenciennes, 1840 (common name "moncholo" or "bagre blanco" is an endemic species of the family Pimelodidae in the río de la Plata basin. Phylogenetic approach based on cytochrome b sequences was performed to test the existence of a unique evolutionary lineage in P. albicans and to discriminate populations units or subpopulations related to a migration behavior of this taxon in the río de la Plata basin. This study included 34 samples of P. albicans of different collecting sites in the río de la Plata estuary and in the río Arrecifes belonging to the río Paraná basin. Among 614 base pairs in the cytochrome b sequence data set, 203 were variable and 120 were phylogenetically informative sites in P. albicans. A total of twenty haplotypes, nucleotide diversity (p = 0.032 and haplotype diversity = 0.941 were found. Tajima's test showed significant value D= -1.88 (p<0.05 rejecting the neutral mutation hypothesis for the P. albicans data set. All phylogenetic approaches showed that P. albicans included four monophyletic assemblages that were supported by high bootstrap and Bayesian posterior probability values. Minimum spanning network corroborated these groups for P. albicans haplotypes. High genetic structure was found in P. albicans by means of AMOVA analysis showing that the río Arrecifes samples constitute an isolated lineage. Moreover, the high value of genetic divergence (10% between the río de la Plata and the río Arrecifes populations could suggest that P. albicans may be conformed by a sibling species complex. On the other hand, a degree of genetic structuring was detected among different sites of the río de la Plata. A partial isolation of the 760 site may suggest that P. albicans could migrates to different tributaries for reproduction, generating different schools of haplotypes which could mix in the río de la Plata estuary. The high nucleotide diversity found in the 765 site and the existence of gene flow

Calculation of neutronic parameters of IEA-R1 reactor and purpose of a new configuration

International Nuclear Information System (INIS)

Kosaka, N.; Fanaro, L.C.C.B.; Yamaguchi, M.

1989-01-01

The program for reducing the fuel enrichment of the IEA-R1 reactor considers fuel plates containing U308-AL with 19,9% of U-235. The geometry of the new 18 fuel plate fuel elements has been kept the same. This work describes the calculation methods utilized at IPEN-CNEN/SP and some neutronic parameters of the present configuration of IEA-R1 as well as for a new configuration porposed with a new LEU fuel element are shown. (author) [pt

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Process modeling of a reversible solid oxide cell (r-SOC) energy storage system utilizing commercially available SOC reactor

International Nuclear Information System (INIS)

Mottaghizadeh, Pegah; Santhanam, Srikanth; Heddrich, Marc P.; Friedrich, K. Andreas; Rinaldi, Fabio

2017-01-01

Highlights: • An electric energy storage system was developed based on a commercially available SOC reactor. • Heat generated in SOFC mode of r-SOC is utilized in SOEC operation of r-SOC using latent heat storage. • A round trip efficiency of 54.3% was reached for the reference system at atmospheric pressure. • An improved process system design achieved a round-trip efficiency of 60.4% at 25 bar. - Abstract: The increase of intermittent renewable energy contribution in power grids has urged us to seek means for temporal decoupling of electricity production and consumption. A reversible solid oxide cell (r-SOC) enables storage of surplus electricity through electrochemical reactions when it is in electrolysis mode. The reserved energy in form of chemical compounds is then converted to electricity when the cell operates as a fuel cell. A process system model was implemented using Aspen Plus® V8.8 based on a commercially available r-SOC reactor experimentally characterized at DLR. In this study a complete self-sustaining system configuration is designed by optimal thermal integration and balance of plant. Under reference conditions a round trip efficiency of 54.3% was achieved. Generated heat in fuel cell mode is exploited by latent heat storage tanks to enable endothermic operation of reactor in its electrolysis mode. In total, out of 100 units of thermal energy stored in heat storage tanks during fuel cell mode, 90% was utilized to offset heat demand of system in its electrolysis mode. Parametric analysis revealed the significance of heat storage tanks in thermal management even when reactor entered its exothermic mode of electrolysis. An improved process system design demonstrates a system round-trip efficiency of 60.4% at 25 bar.

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

Energy Technology Data Exchange (ETDEWEB)

Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

Energy Technology Data Exchange (ETDEWEB)

Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

International Nuclear Information System (INIS)

Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

2008-01-01

The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

Device for neutron flux monitoring in IEA-R1 reactor using rhodium self powered neutron detector; Dispositivo de mapeamento de fluxo de neutron atraves do SPN/Rodio no IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Ricci Filho, Walter; Fernando, Alberto de Jesus; Jerez, Rogerio; Tondin, Julio B.M.; Pasqualetto, Hertz [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

2000-07-01

The IEA-R1 reactor has undergone a modernization tio increase its operating power to 5 MW, in order to allow a more efficient production of radioisotopes. The objective of this work is to provide the reactor with flux monitoring device using a rhodium self powered neutron detector. Self powered detectors are rugged miniature devices with are increasingly being used for fixed in core reactor monitoring both for safety purposes and flux mapping. The work presents the results obtained with Rhodium-SPND in several irradiation position inside the reactor core. (author)

Cofilin phosphorylation is elevated after F-actin disassembly induced by Rac1 depletion.

Science.gov (United States)

Liu, Linna; Li, Jing; Zhang, Liwang; Zhang, Feng; Zhang, Rong; Chen, Xiang; Brakebusch, Cord; Wang, Zhipeng; Liu, Xinyou

2015-01-01

Cytoskeletal reorganization is essential to keratinocyte function. Rac1 regulates cytoskeletal reorganization through signaling pathways such as the cofilin cascade. Cofilin severs actin filaments after activation by dephosphorylation. Rac1 was knocked out in mouse keratinocytes and it was found that actin filaments disassembled. In the epidermis of mice in which Rac1 was knocked out only in keratinocytes, cofilin phosphorylation was aberrantly elevated, corresponding to repression of the phosphatase slingshot1 (SSH1). These effects were independent of the signaling pathways for p21-activated kinase/LIM kinase (Pak/LIMK), protein kinase C, or protein kinase D or generation of reactive oxygen species. Similarly, when actin polymerization was specifically inhibited or Rac1 was knocked down, cofilin phosphorylation was enhanced and SSH1 was repressed. Repression of SSH1 partially blocked actin depolymerization induced by Rac1 depletion. Therefore, aberrant cofilin phosphorylation that induces actin polymerization might be a consequence of actin disassembly induced by the absence of Rac1. © 2015 International Union of Biochemistry and Molecular Biology.

Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Baptista Filho, Benedito Dias

2000-01-01

This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

Engineering multifunctional protein nanoparticles by in vitro disassembling and reassembling of heterologous building blocks

Science.gov (United States)

Unzueta, Ugutz; Serna, Naroa; Sánchez-García, Laura; Roldán, Mónica; Sánchez-Chardi, Alejandro; Mangues, Ramón; Villaverde, Antonio; Vázquez, Esther

2017-12-01

The engineering of protein self-assembling at the nanoscale allows the generation of functional and biocompatible materials, which can be produced by easy biological fabrication. The combination of cationic and histidine-rich stretches in fusion proteins promotes oligomerization as stable protein-only regular nanoparticles that are composed by a moderate number of building blocks. Among other applications, these materials are highly appealing as tools in targeted drug delivery once empowered with peptidic ligands of cell surface receptors. In this context, we have dissected here this simple technological platform regarding the controlled disassembling and reassembling of the composing building blocks. By applying high salt and imidazole in combination, nanoparticles are disassembled in a process that is fully reversible upon removal of the disrupting agents. By taking this approach, we accomplish here the in vitro generation of hybrid nanoparticles formed by heterologous building blocks. This fact demonstrates the capability to generate multifunctional and/or multiparatopic or multispecific materials usable in nanomedical applications.

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

Test reports for K Basins vertical fuel handling tools

Energy Technology Data Exchange (ETDEWEB)

Meling, T.A.

1995-02-01

The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Analysis of the IEA-R1 reactor start-up procedures - an application of the HazOp method

International Nuclear Information System (INIS)

Sauer, Maria Eugenia Lago Jacques

2000-01-01

An analysis of technological catastrophic events that took place in this century shows that human failure and vulnerability of risk management programs are the main causes for the occurrence of accidents. As an example, plants and complex systems where the interface man-machine is close, the frequency of failures tends to be higher. Thus, a comprehensive knowledge of how a specific process can be potentially hazardous is a sine qua non condition to the operators training, as well as to define and implement more efficient plans for loss prevention and risk management. A study of the IEA-R1 research reactor start-up procedures was carried out, based upon the methodology Hazard and Operability Study (HazOp). The analytical and qualitative multidisciplinary HazOp approach provided means to a comprehensive review of the reactor start-up procedures, contributing to improve the understanding of the potential hazards associated to deviations on performing this routine. The present work includes a historical summary and a detailed description of the HazOp technique, as well as case studies in the process industries and the use of expert systems in the application of the method. An analysis of 53 activities of the IEA-R1 reactor start-up procedures was made, resulting in 25 recommendations of changes covering aspects of the project, operation and safety of the reactor. Eleven recommendations have been implemented. (author)

Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

Energy Technology Data Exchange (ETDEWEB)

Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

1986-10-01

There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

Energy Technology Data Exchange (ETDEWEB)

2009-01-15

This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

International Nuclear Information System (INIS)

2009-01-01

This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

Evaluation of water chemistry on the pitting susceptibility of aluminum

International Nuclear Information System (INIS)

Chandler, G.T.; Sindelar, R.L.; Lam, P.S.

1997-01-01

Aluminum-clad spent nuclear fuels are being stored in water in the Receiving Basin for Off-site Fuels (RBOF) and the reactor disassembly (cooling) basins at the Savannah River Site (SRS). Experience shows that fuels stored in water are subject to rapid pitting corrosion if the water quality is poor. Upgrade projects and actions, including those to improve water quality, were recently undertaken to upgrade the disassembly basins for extended storage. A technical strategy was developed for continued basin storage of aluminum-clad fuel assemblies. The strategy includes development and implementation of basin technical standards for water quality to minimize attack due to pitting corrosion over a desired storage period. In the absence of localized corrosion, only slow, general corrosion of the cladding would be expected. A laboratory corrosion program is being performed to provide the bases for technical standards by identifying the region of aggressive water qualities where existing oxide films would tend to break down and pits would initiate and remain active. Initial results from corrosion potential and cyclic polarization testing of aluminum alloys in various water chemistries have shown that low conductivity water (< 50 μS/cm) should not be aggressive to cause self-pitting corrosion. Initial results from tests of 8001 and 5052 aluminum and aluminium-10% uranium alloy indicate that a strong galvanic couple should not exist between the aluminum cladding materials and the aluminum-uranium fuel. Additional laboratory testing will include immersion testing to allow characterization of the growth rate of active pits to benchmark a kinetic model. This model will form the basis for a water quality technical standard and enable prediction of the life of aluminum-clad spent nuclear fuels in basin storage

Sampling and analysis plan for the 116-C-5 retention basins characteristic dangerous waste determination

International Nuclear Information System (INIS)

Bauer, R.G.; Dunks, K.L.

1996-03-01

Cooling water flow from the rear face of the 100-B and 100-C reactors was diverted to large retention basins prior to discharge to the Columbia River. These retention basins delayed the release of the reactor coolant for decay of the short-lived activation products and for thermal cooling. Some of the activation products were deposited in sludge that settled in the basins and discharge lines. In addition, some contamination was deposited in soil around the basins and associated piping. The sampling objective of this project is to determine if regulated levels of leachable lead are present in the abrasive materials used to decontaminate the retention basin tank walls, in the material between the tank base plate and the concrete foundation, and in the soils immediately surrounding the perimeter of the retention basins. Sampling details, including sampling locations, frequencies, and analytical requirements, are discussed. Also described is the quality assurance plan for this project

Sequential phosphorylation of GRASP65 during mitotic Golgi disassembly

Directory of Open Access Journals (Sweden)

Danming Tang

2012-09-01

GRASP65 phosphorylation during mitosis and dephosphorylation after mitosis are required for Golgi disassembly and reassembly during the cell cycle. At least eight phosphorylation sites on GRASP65 have been identified, but whether they are modified in a coordinated fashion during mitosis is so far unknown. In this study, we raised phospho-specific antibodies that recognize phosphorylated T220/T224, S277 and S376 residues of GRASP65, respectively. Biochemical analysis showed that cdc2 phosphorylates all three sites, while plk1 enhances the phosphorylation. Microscopic studies using these antibodies for double and triple labeling demonstrate sequential phosphorylation and dephosphorylation during the cell cycle. S277 and S376 are phosphorylated from late G2 phase through metaphase until telophase when the new Golgi is reassembled. T220/224 is not modified until prophase, but is highly modified from prometaphase to anaphase. In metaphase, phospho-T220/224 signal localizes on both Golgi haze and mitotic Golgi clusters that represent dispersed Golgi vesicles and Golgi remnants, respectively, while phospho-S277 and S376 labeling is more concentrated on mitotic Golgi clusters. Expression of a phosphorylation-resistant GRASP65 mutant T220A/T224A inhibited mitotic Golgi fragmentation to a much larger extent than the expression of the S277A and S376A mutants. In cytokinesis, T220/224 dephosphorylation occurs prior to that of S277, but after S376. This study provides evidence that GRASP65 is sequentially phosphorylated and dephosphorylated during mitosis at different sites to orchestrate Golgi disassembly and reassembly during cell division, with phosphorylation of the T220/224 site being most critical in the process.

Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

Project Management Plan 105-KE Basin sludge retrieval and packaging

International Nuclear Information System (INIS)

McWethy, L.M.

1994-01-01

The KE Basin contains over 1,100 metric tons of spent nuclear fuel (SNF). The bulk of this inventory consists of over 50,000 zircaloy clad, uranium metal N-Reactor fuel element assemblies, along with less than half a metric ton of single-pass reactor fuel elements, stored in over 3,600 open top canister assemblies. In addition, sludge containing fissile and fission product material from damaged/degraded fuel has accumulated in the basin. The sludge, particularly the fines, impacts basin operations by clouding the water and making activities requiring a clear view impossible to complete until after sludge settles. Packaging would get the sludge out of the operator's way and allow it to be moved within the basin in a more manageable state. The primary project objective is to develop, procure, and quality the equipment needed to remove all sludge from the KE Basin with minimal dose commitment, minimal cost, and on schedule. The project will provide: (1) the development, testing, and installation of equipment for sludge retrieval and packaging; (2) understanding of and experience with actual sludge through near-term sludge packaging feature tests in the KE Basin; (3) sludge removal and handling equipment required to support debris removal, fuel handling, and other activities involving sludge within the KE Basin; and (4) enlist industry expertise in all phases of the project. This Project Management Plant establishes the organizational responsibilities, control systems, and procedures for the execution of project activities for KE Basin sludge retrieval packaging, to meet programmatic requirements within authorized funding and approved schedules

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

2001-01-01

The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

Considerations about decommissioning of the IEA-R1 research reactor and the future of its installations after shutdown

International Nuclear Information System (INIS)

Frajndlich, Roberto

2014-01-01

The IEA-R1 Nuclear Research Reactor, in operation since 1957, in the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), is one of the oldest research reactors in the world. However at some point in time in the future, as example of the other reactors, it will be shutdown definitively. Before that time actually arrives, the operational organization needs to plan the future of its installations and define the final destination of equipment and radioactive as well as non-radioactive material contained inside the installations. These and other questions should be addressed in the so called Preliminary decommissioning plan of the installation, which is the subject of this work. The work initially presents an over view about the theme and defines the general and specific objectives describing, in succession, the directions that the operating organization should consider for the formulation of a decommissioning plan. The present structure of the Brazilian nuclear sector emphasizing principally the norms utilized in the management of radioactive waste is also presented. A description of principle equipment of the IEA-R1 reactor which constitutes its inventory of radioactive and non-radioactive material is given. The work emphasizes the experience of the reactor technicians, acquired during several reforms and modifications of the reactor installations realized during its useful life time. This experience may be of great help for the decommissioning in the future. An experiment using the high resolution gamma spectrometric method and computer calculation using Monte Carlo theory were performed with the objective of obtaining an estimate of the radioactive waste produced from dismantling of the reactor pool walls. The cost of reactor decommissioning for different choices of strategies was determined using the CERREX code. Finally, a discussion about different strategies is presented. On the basis of these discussions it is concluded that the most advantageous

Effect of fluid-to-structure heat transfer on the structural damage potential to a liquid-metal fast breeder reactor

International Nuclear Information System (INIS)

Hakim, S.J.; Abramson, P.B.

1979-01-01

Deterministic calculations simulating a hypothetical accident in a liquid-metal fast breeder reactor that leads to a hydrodynamic disassembly of the core have been carried out to estimate the system's damage potential due to the vapor-pressure-driven expansion of molten core material and its dependency on the heat transfer to the remaining structure. These calculations ignored the effect on the work potential of sodium left in the core during the disassembly. Results indicate that steel cladding in the upper axial blankets and fission gas plenum acts as a thermodynamic energy sink that could reduce the total thermodynamic work energy by between one and two orders of magnitude, provided little or no sodium is present in the core at the time of interaction. These results have been found to be insensitive to the rate of heat transferred from the molten fuel to the molten steel that comprises the molten core material

Test plan for K-Basin fuel handling tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

2013-01-01

The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

Energy Technology Data Exchange (ETDEWEB)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.br, E-mail: amir@cdtn.br, E-mail: fsl@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

International Nuclear Information System (INIS)

Edwards, G.T.; Schluderberg, D.C.

1981-01-01

A technique is provided for engaging and disengaging burnable poison rods from a spider in a nuclear reactor fuel assembly. A cap on the end of each of the burnable poison rods is provided with a shank or stem that is received in a respective bore formed in the spider. A frangible flange secures the shank and rod to the spider. Pressing the shank in the direction of the bore axis by means, e.g., of a plate ruptures the frangible flange to release the rod from the spider. (author)

Improved and verification of fast reactor safety analysis techniques. Annual summary, March 1, 1975--February 29, 1976

International Nuclear Information System (INIS)

Jackson, J.F.; Bott, T.F.

1976-01-01

Analyses of the Kiwi-TNT and SNAPTRAN-2 experiments have been performed with the VENUS-II fast-reactor disassembly code. The results show that VENUS-II provides an adequate characterization of these experiments. As is the case for LMFBRs, the excursions were initially turned over by temperature feedback effects, with ultimate shutdown coming from core disassembly. The calculated fission energies agree with the experimental values to within about 50 percent for the Kiwi excursion and 10 percent for the SNAPTRAN-2 experiment. The results of the analyses are being evaluated to determine the reasons for the remaining differences. It appears that part of the difference observed in the Kiwi-TNT analysis could relate to not explicitly treating the heat-transfer from the beaded fuel (a problem not present in LMFBR calculations). Both analyses also have uncertainties associated with the new equation-of-state that had to be added to VENUS-II to allow treatment of the core materials not used in fast reactors. Finally, there are uncertainties in the temperature feedback coefficients being used. In general, the uncertainties associated with applying VENUS-II to LMFBR excursions should be even smaller than those encountered in these experimental comparisons. This is because the temperature (Doppler) coefficients and core material equations-of-state are better known, and the complications associated with heat transfer from the beaded fuel are not present

[Study on the liver-protective and choleretic effect of zhizi baipi soup and its disassembled prescription].

Science.gov (United States)

Xiao, Xu; Zhu, Ji-Xiao; Luo, Guang-Ming; Li, Lei; Zhu, Yu-Ye; Zeng, Jin-Xiang; Wang, Xiao-Yun; Wu, Bo

2013-07-01

To investigate the effect of Zhizi Baipi soup and its disassembled prescription on protecting liver and improving choleresis and explore the regularity of Zhizi Baipi soup composition. The model of mouse liver injury induced by carbon tetraehlofide (CCl4) was used to observe the effects of Zhizi Baipi soup and its disassembled prescription by oral adminstration, the bile volume was determinied by common bile duct drainage. Zhizi Baipi soup and each treatment group with gardenia could significantly inhibit the increased serum ATL and AST activities, reduce liver MDA level, and significantly promote the bile flow and bilirubin in bile in normal rats. Zhizi Baipi soup has effects on protecting liver and increasing bile secretion, its monarch drug, gardenia plays an important role in the decoction, the effect of eliminating dampness and heat are mainly ascribed to the synergic effect of gardenia and phellodendron.

Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

Remote disassembly of radioactively contaminated vessels by means of an arc saw

International Nuclear Information System (INIS)

Beitel, G.A.

1977-01-01

The arc saw, a newly developed tool, is a toothless circular saw which cuts by means of an electric arc. Cutting speeds between 20 to 30 cm 2 /S and depths up to 45 cm are possible. There is no mechanical contact between blade and work piece, no binding, and no blade breakage. The arc saw will be applied to the rapid and remote disassembly of multiple ton, contaminated stainless steel vessels

Commercial nuclear reactors and waste: the current status

International Nuclear Information System (INIS)

Platt, A.M.; Robinson, J.V.

1980-04-01

During the last five years, the declared size of the commercial light water reactor (LWR) nuclear power industry in the US has steadily decreased. As of January 1980, the total number of power plants had dropped to 191 from the 226 in December 31, 1974. At least another nine were cancelled in the last few months. This report was developed as the first of a series to track implications to waste management due to such changes in the declared size of the industry. For the presently declared size, key conclusions are: the declared reactors will peak at a capacity of 162 GWe and consume about 10 6 MTU as enrichment feed. As few as two repositories of about 100,000 MTHM capacity each would hold the waste. Predisposal storage (reactor basins and AFRs) would peak at less than 100,000 MTHM (in the year 2020) with one repository opening in the year 1997 and the other in the year 2020. Most of the 100,000 MTHM would have to be in AFR storage unless current practice regarding reactor basin size was radically changed

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

International Nuclear Information System (INIS)

Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2004-01-01

The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2011-07-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

2011-01-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

International Nuclear Information System (INIS)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

2017-01-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

2017-11-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

Dynamic loading of the structural wall in a lithium fall fusion reactor

International Nuclear Information System (INIS)

Glenn, L.A.

1979-01-01

In one version of an inertial confinement fusion (ICF) power reactor, the laser-imploded pellet is surrounded by a thick, annular 'waterfall' of liquid lithium. The fall has three functions: to breed tritium for pellet resupply, to act as an energy sink and heat exchange mdeium with an external power loop, and to protect the first wall of the reactor from excessive neutronic and hydrodynamic loading. Our primary concern here is with this last function. We formulated a simple model of a lithium-fall ICF reactor and calculated the fall disassembly and the subsequent fluid-wall interaction resulting from the energy deposition by the imploded pellet. Two potential mechanisms for wall damage were identified: surface erosion and hoop failure. For single fall designs, the erosion problem appears to be serious. Concentric annuli (multiple fall) or packed jet configurations may be feasible but experiments are needed to clarify the physical model, especially with reg (orig.)ard to /orig.the characteristics of the cavitated liquid lithium and of the two-phase liquid-vapor region.

Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

International Nuclear Information System (INIS)

Lord, D.L.; Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L.

1993-04-01

A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed

Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

Energy Technology Data Exchange (ETDEWEB)

Lord, D.L. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1993-04-01

A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed.

2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

International Nuclear Information System (INIS)

Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

1969-01-01

1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

R[ionuclide transport after the Chernobyl reactor accident and derivation of r[ioecological parameters

International Nuclear Information System (INIS)

Bonka, H.

1998-01-01

Since due to the nuclear reactor accident in Chernobyl r[ionuclides arrived in the vicinity of Aachen, the enhancement of the local dose rate, the deposition of the different r[ionuclides on ground and vegetation and the transport of the r[ionuclides into the environment were measured. Partly the measurements were continued until today. Very informative time sequences of the specific activity in grass, food, cow's milk, beef, in the different plants, trees, ploughed soil and undisturbed soil, mushrooms, game, in humans etc. resulted. During different private and official journeys in the old Laender of the Federal Republic of Germany surface covering measurements of the 134 Cs and 137 Cs activity deposited on grass land at different places were carried out. These data were implemented into a map on ground contamination in 1986 in Germany, published in 1991 by the Institute for Water, Soil and Air Hygiene of the Federal Public Health Department in Berlin. Transfer factors soil-grass were measured in the whole Federal Republic of Germany analyzing grass samples which were partly taken at the same time. A large amount of r[ioecological parameters could be derived from the different time sequences. These are in particular: The deposition velocity for iodine and particle bound r[ioanuclides on grass and in forests, the rainout coefficient in dependence of the precipitation intensity, the retention factors on grass, the biological half-life time on grass, the transfer factor soil-grass in dependence of time, the transfer factor food-milk during the pasture period and during stable stay, the transfer factor food-beef, the transfer factors in eatable mushrooms, the translocation factor of cesium in cereals etc. A multi-compartment model was developed to calculate the specific Cs activity in cow's milk and beef. The specific activity in milk can be calculated sufficiently exact using a simple single compartment model. The correlation of the specific Cs activity in spruce

Instrumental Neutron Activation Analysis- INAA: environmental studies in Das Velhas Basin, Minas Gerais, Brazil

International Nuclear Information System (INIS)

Rabelo V, M.A.; Andrade Q, M.T.; Araujo M, R.; Albernaz A, I.; Oliveira, A.H. de

2006-01-01

The Instrumental Neutron Activation Analysis - INAA was applied to determine concentrations of several elements in unpolluted areas and in the mining and farming region of the Das Velhas Basin, Minas Gerais State, Brazil. INAA was applied using the TRIGA Mark I IPR - R1 reactor at the Nuclear Technology Development Center of the National Committee of Nuclear Energy (CDTN/CNEN), in Belo Horizonte city, Minas Gerais State. At 100 kW of potency the flux of neutrons is 6.6 10 11 n.cm -2 .s -1 . The samples analyzed were: water; sediment; gravel of gold mine and forage. The obtained results for the Das Velhas Basin in water and sediment samples - mining companies region - show a high level (μg/g) of contamination with the analyzed elements, mainly in the sediment samples. During the period of floods, in farming region hundreds of kilometers away, contamination is found in fish and forage, reaching and harming both people and animals that live in the marginal region. (Author)

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

International Nuclear Information System (INIS)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-01

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-15

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

The Hack's law applied to young volcanic basin: the Tahiti case

Science.gov (United States)

Ye, F.; Sichoix, L.; Barriot, J.; Serafini, J.

2010-12-01

We study the channel morphology over the Tahiti island from the Hack’s law perspective. The Hack’s law is an empirical power relationship between basin drainage area and the length of its main channel. It had also been shown that drainage area becomes more elongate with increasing basin size. For typical continental basins, the exponent value lies between 0.47 for basins larger than 260,000 km2 and 0.7 for those spanning less than 20,720 km2 (Muller, 1973). In Tahiti, we extracted 27 principal basins ranging from 7 km2 to 90 km2 from a Digital Terrain Model of the island with a 5 m-resolution. We demonstrate that the Hack’s law still apply for such small basins (correlation coefficient R2=0.7) with an exponent value being approximately 0.5. It appears that the exponent value is influenced by the local geomorphic condition, and does not follow the previous study results (the exponent value decreases with increasing drainage area.) Our exponent value matches the result found w.r.t. debris-flow basins of China for drainage areas less than 100 km2 (Li et al., 2008). Otherwise, the young volcanic basins of Tahiti do not become longer and narrower with increasing basin size (R2=0.1). Besides, there is no correlation between the basin area and the basin convexity (R2=0). This means that there is no statistical change in basin shape with basin size. We present also the drainage area-slope relationship with respect to sediment or transport-limited processes. Key words: Hack’s law, channel morphology, DTM

[PHAHs levels in soil samples from the E-waste disassembly sites and their sources allocation].

Science.gov (United States)

Zhao, Gao-Feng; Wang, Zi-Jian

2009-06-15

Soil samples (each with 3 replicates of - 1 kg, at the top 0-5 cm layer) were collected from each of the e-waste disassembly sites and the control site. Also obtained from each disassembly site were samples (each weighing - 0.2 kg) of cable coating,stuffing powder, and circuit boards chipping. The contents of 23 PBB congeners, 12 PBDE congeners, and 27 PCB congeners in soil and in their potential sources, including e-waste residues, were measured using the GC-MS5975B technique. The highest level of PBBs was found in the cable coating among the three e-waste residues, with a concentration of 35.25 ng x g(-1). The contents of low-brominated PBBs (including monobromobiphenyls and dibromobiphenyls) accounted for 38% of the total PBBs concentration observed in cable coating sample. The highest levels of PBDEs and PBDE209 were found in the stuffing powder for electronic component among the collected e-waste residues, with a concentration of 29.71 and 4.19 x 10(3) ng x g(-1). PBDE153 and PBDE183 were the most predominant PBDE congeners, with their concentration accounting for 43% and 24% of the total PBDEs concentration observed in the stuffing powder sample, respectively. Levels of PCBs in cable coating were the highest in these e-waste residues, with a concentration of 680.02 ngx g(-1). The observed values of the three PHAHs in soils from the disassembly site were considerably higher than their corresponding values observed in the control site (p < 0.05), which indicates that these PHAHs from e-waste is the pollution source of local environment.

Decommissioning of the AVR reactor, concept for the total dismantling

International Nuclear Information System (INIS)

Marnet, C.; Wimmers, M.; Birkhold, U.

1998-01-01

After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

1997 N-Basin Administrative Control Level Dose Extension

International Nuclear Information System (INIS)

Nellesen, A.L.

1997-04-01

This document provides justification for extending the Administrative Control Level of 500 mrem per year to 1,000 mrem per year Total Effective Dose Equivalent for workers involved with N-Reactor Basin Deactivation in accordance with established procedures

Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor; Estudo da viabilidade de aplicação do método Prompt Gamma Neutron Activation Analysis (PGNAA) no reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Guerra, Bruno Teixeira

2016-07-01

The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

Tectonic-sedimentary evolution of foreland basins: U-Pb dating of the discharge that would have originated the piggy-back basin of Rodeo-Iglesias, San Juan-Argentina

International Nuclear Information System (INIS)

Santos, Romulo Duarte Moreira dos; Hauser, Natalia; Matteini, Massimo; Pimentel, Marcio Martins

2015-01-01

Between the 28 ° and 31 ° LS parallels of the Argentinean west, in the province of San Juan, foreland basins originated by the subhorizontal subduction of oceanic crust as a result of the Andean orogeny in the late Oligocene emerges. The Bermejo basin and Rodeo-Iglesias piggy-back basin would be associated with the progressive development of landslides, backscatter and minor faults, and basin fragmentation. Two samples of volcanic rocks, R-1 (rhyolitic dome) and R-3 (fall deposit) of the Rodeo-Iglesias basin, had ages of 8.2 ± 0.11 Ma and 8.7 ± 0.24 Ma. At the same time, the age of the (R-1) made it possible to infer quantitatively the age of the first cavalcade that occurred approximately 8.2 ± 0.11 Ma. From the data obtained in the Rodeo-Iglesias basin both volcanism and the first cavalcade could have been synchronous

The R and D issues necessary to achieve the safety design of commercialized liquid-metal cooled fast reactors

International Nuclear Information System (INIS)

Shoji, Kotake; Koji, Dozaki; Shigenobu, Kubo; Yoshio, Shimakawa; Hajime, Niwa; Masakazu, Ichimiya

2002-01-01

Within the framework of the feasibility study on commercialized fast reactor cycle systems (hereafter described as F/S), the safety design principle is investigated and several kinds of design studies are now in progress. Among the designs for liquid-metal cooled fast reactor (LMR), the advanced loop type sodium cooled fast reactor (FR) is one of the promising candidate as future commercialized LMR. In this paper, the safety related research and development (R and D) issues necessary to achieve the safety design are described along the defence-in-depth principle, taking account of not only the system characteristics of the advanced loop concepts but also design studies and R and D experiences so far. Safety issues related to the hypothetical core disruptive accidents (CDA) are emphasized both from the prevention and mitigation. A re-criticality free core concept with a special fuel assembly is pursued by performing both analytical and experimental efforts, in order to realize the rational design and to establish easy-to-understand safety logic. Sodium related issues are also given to ensure plant availability and to enhance the acceptability to the public. (authors)

Nuclear research reactor IEA-R1 heat exchanger inlet nozzle flow - a preliminary study

International Nuclear Information System (INIS)

Angelo, Gabriel; Andrade, Delvonei Alves de; Fainer, Gerson; Angelo, Edvaldo

2009-01-01

As a computational fluid mechanics training task, a preliminary model was developed. ANSYS-CFX R code was used in order to study the flow at the inlet nozzle of the heat exchanger of the primary circuit of the nuclear research reactor IEA-R1. The geometry of the inlet nozzle is basically compounded by a cylinder and two radial rings which are welded on the shell. When doing so there is an offset between the holes through the shell and the inlet nozzle. Since it is not standardized by TEMA, the inlet nozzle was chosen for a preliminary study of the flow. Results for the proposed model are presented and discussed. (author)

Connexin43 hemichannels contributes to the disassembly of cell junctions through modulation of intracellular oxidative status

Directory of Open Access Journals (Sweden)

Yuan Chi

2016-10-01

Full Text Available Connexin (Cx hemichannels regulate many cellular processes with little information available regarding their mechanisms. Given that many pathological factors that activate hemichannels also disrupts the integrity of cellular junctions, we speculated a potential participation of hemichannels in the regulation of cell junctions. Here we tested this hypothesis. Exposure of renal tubular epithelial cells to Ca2+-free medium led to disassembly of tight and adherens junctions, as indicated by the reduced level of ZO-1 and cadherin, disorganization of F-actin, and severe drop in transepithelial electric resistance. These changes were preceded by an activation of Cx43 hemichannels, as revealed by extracellular efflux of ATP and intracellular influx of Lucifer Yellow. Inhibition of hemichannels with chemical inhibitors or Cx43 siRNA greatly attenuated the disassembly of cell junctions. Further analysis using fetal fibroblasts derived from Cx43 wide-type (Cx43+/+, heterozygous (Cx43+/- and knockout (Cx43-/- littermates showed that Cx43-positive cells (Cx43+/+ exhibited more dramatic changes in cell shape, F-actin, and cadherin in response to Ca2+ depletion, as compared to Cx43-null cells (Cx43-/-. Consistently, these cells had higher level of protein carbonyl modification and phosphorylation, and much stronger activation of P38 and JNK. Hemichannel opening led to extracellular loss of the major antioxidant glutathione (GSH. Supplement of cells with exogenous GSH or inhibition of oxidative sensitive kinases largely prevented the above-mentioned changes. Taken together, our study indicates that Cx43 hemichannels promote the disassembly of cell junctions through regulation of intracellular oxidative status.

Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

International Nuclear Information System (INIS)

Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

2013-01-01

The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

Energy Technology Data Exchange (ETDEWEB)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H., E-mail: songm@dgu.edu [Department of Mechanical, Robotics, and Energy Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of)

2015-03-15

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

International Nuclear Information System (INIS)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

2015-01-01

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made

Preliminary analysis of control rod accidents in the CRCN-R1 multipurpose reactor core of Recife in Brazil

International Nuclear Information System (INIS)

Souza dos Santos, Rubens; Rubens Maiorino, Jose

1999-01-01

The paper shows some results of the neutronic accident analyses arisen by uncontrolled control rod withdrawal, based on the Conceptual Project of the CRCN-R1 MultiPurpose Reactor of Recife. In that reactor, a project of the CNEN/Brazil, under the leadership of the IPEN/Sao Paulo, is verified the thermal hydraulic limits in the reactor core during transients that simulate startup and power operation accidents. It has utilized a computer program that solved the kinetic equations based on multigroup diffusion theory, in our case we have used 4 energy groups, Two-Dimensional X-Y in the space, and 6 groups of delayed neutrons. A simple model of feedback is admitted in the capture and scattering macroscopic cross sections, in the fuel regions, temperature and coolant densities dependents. Based on those models, the results demonstrated that the reactor exhibits good degree of safety. (author)

Nuclear R and D program in Indonesia and selection of future research reactor to support it

International Nuclear Information System (INIS)

Baiquni, A.; Subki, I.

1981-01-01

The nuclear R and D program selection decision is described as a phased program, each phase having its specific objective. The elements of each phase are identified and related with the objective, from which the activities of each element are also broadly outlined. To support the nuclear R and D program and to realize the objectives in each phase, the research facilities are also developed. A new nuclear development center housing a multipurpose reactor (MPR) and various laboratories are also described. The choice of the MPR and its criteria are also described briefly

Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; the 118-H-6:6 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

International Nuclear Information System (INIS)

Appel, M.J.

2006-01-01

This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31)

Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

International Nuclear Information System (INIS)

Mesquita, A.Z.; Maretti, F. Jr.; Rezende, H.C.; Tambourgi, E.B.

2004-01-01

The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

The French R and D programme for fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Auchapt, P.; Bourgeois, M.; Calame-Longjean, A.; Miquel, P.; Sauteron, J.

1979-01-01

The process employed is the Purex process adapted to the specific case of fast breeder reactor fuels. The results achieved have demonstrated that the aqueous method can be applied to these fuels: nearly ten years of operation in the ATl workshop which reprocesses RAPSODIE fuels, and the good results obtained at the Marcoule pilot facility on large batches of fuel attest to this achievement. The CEA effort continues principally on extrapolation to industrial scale, thanks mainly to experiments conducted on industrial prototypes and to the launching of the TOR project, which will, as of 1984, allow reprocessing of FBR fuels on a significant scale, and which will provide extensive additional resources for R and D activities

Piracicaba River Basin: evaluation of chemical elements in deep sediment profile by INAA

Energy Technology Data Exchange (ETDEWEB)

França, Elvis J. de; Santos, Robson A.; Santos, Katarine M. Barbosa; Silva, Gleyce K. A. [Centro Regional de Ciencias Nucleares do Nordeste (DIAMB/CRCN-NE/CNEN-PE), Recife, PE (Brazil). Div, de Monitoração Ambiental; Fernandes, Elisabete A. de Nadai; Rodrigues, Vanessa S.; Cavalca, Isabel P.O., E-mail: ejfranca@cnen.gov.br, E-mail: lis@cena.usp.br [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil)

2017-07-01

Many hydrographic basins have been impacted by anthropogenic sources, the Piracicaba River Basin of the State of São Paulo, is one of that. The total concentrations of chemical elements in deep sediments of basin may be indicate those available in ecosystem. Therefore, in this research concentration of chemical elements on deep sediment profile sampled of Piracicaba River Basin was determined by k0-Instrumental Neutron Activation Analysis. After collecting the 60 cm depth profile, samples were obtained by sectioning the sediment profile in 5 cm layers, totalizing 12 samples. Analytical portions were transferred to polyethylene vials for neutron irradiation at the Nuclear Research Reactor IEA-R1 from the Instituto de Pesquisas Energéticas e Nucleares IPEN/CNEN. After waiting for radiation safe levels, irradiated samples were transported to the Radioisotopes Laboratory from the Centro de Energia Nuclear na Agricultura CENA/USP. High Resolution Gamma-Ray Spectrometry using HPGe detectors was applied to measure the induced radioactivity. The chemical element mass fractions and their respective expanded analytical uncertainties (95% confidence level) were determined by k0-INAA using the in-house software Quantu. Geological reference materials were analyzed with samples to evaluate the quality of the analytical procedure. Results indicated the presence of enriched surface sediments (0-10 cm depth) for As, Ba, Ca, Co, Cs, Fe, Sb, Sc, Sr, Yb and Zn, despite no alteration was observed for Eu, Ta and Tb. Therefore, the evaluation of deep sediment profile afford the chemical element dynamics for the Piracicaba Basin. (author)

Piracicaba River Basin: evaluation of chemical elements in deep sediment profile by INAA

International Nuclear Information System (INIS)

França, Elvis J. de; Santos, Robson A.; Santos, Katarine M. Barbosa; Silva, Gleyce K. A.

2017-01-01

Many hydrographic basins have been impacted by anthropogenic sources, the Piracicaba River Basin of the State of São Paulo, is one of that. The total concentrations of chemical elements in deep sediments of basin may be indicate those available in ecosystem. Therefore, in this research concentration of chemical elements on deep sediment profile sampled of Piracicaba River Basin was determined by k0-Instrumental Neutron Activation Analysis. After collecting the 60 cm depth profile, samples were obtained by sectioning the sediment profile in 5 cm layers, totalizing 12 samples. Analytical portions were transferred to polyethylene vials for neutron irradiation at the Nuclear Research Reactor IEA-R1 from the Instituto de Pesquisas Energéticas e Nucleares IPEN/CNEN. After waiting for radiation safe levels, irradiated samples were transported to the Radioisotopes Laboratory from the Centro de Energia Nuclear na Agricultura CENA/USP. High Resolution Gamma-Ray Spectrometry using HPGe detectors was applied to measure the induced radioactivity. The chemical element mass fractions and their respective expanded analytical uncertainties (95% confidence level) were determined by k0-INAA using the in-house software Quantu. Geological reference materials were analyzed with samples to evaluate the quality of the analytical procedure. Results indicated the presence of enriched surface sediments (0-10 cm depth) for As, Ba, Ca, Co, Cs, Fe, Sb, Sc, Sr, Yb and Zn, despite no alteration was observed for Eu, Ta and Tb. Therefore, the evaluation of deep sediment profile afford the chemical element dynamics for the Piracicaba Basin. (author)

Knowledge of damage identification about tensegrities via flexibility disassembly

Science.gov (United States)

Jiang, Ge; Feng, Xiaodong; Du, Shigui

2017-12-01

Tensegrity structures composing of continuous cables and discrete struts are under tension and compression, respectively. In order to determine the damage extents of tensegrity structures, a new method for tensegrity structural damage identification is presented based on flexibility disassembly. To decompose a tensegrity structural flexibility matrix into the matrix represention of the connectivity between degress-of-freedoms and the diagonal matrix comprising of magnitude informations. Step 1: Calculate perturbation flexibility; Step 2: Compute the flexibility connectivity matrix and perturbation flexibility parameters; Step 3: Calculate the perturbation stiffness parameters. The efficiency of the proposed method is demonstrated by a numeical example comprising of 12 cables and 4 struts with pretensioned. Accurate identification of local damage depends on the availability of good measured data, an accurate and reasonable algorithm.

Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.; Chang, Y.W.

1990-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

Directory of Open Access Journals (Sweden)

Yeong-il Kim

2013-01-01

Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

Research and development of super light water reactors and super fast reactors in Japan

International Nuclear Information System (INIS)

Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

2011-01-01

Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

Performance evaluation of 24 ion exchange materials for removing cesium and strontium from actual and simulated N-Reactor storage basin water

Energy Technology Data Exchange (ETDEWEB)

Brown, G.N.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.

1997-09-01

This report describes the evaluation of 24 organic and inorganic ion exchange materials for removing cesium and strontium from actual and simulated waters from the 100 Area 105 N-Reactor fuel storage basin. The data described in this report can be applied for developing and evaluating ion exchange pre-treatment process flowsheets. Cesium and strontium batch distribution ratios (K{sub d}`s), decontamination factors (DF), and material loadings (mmol g{sup -1}) are compared as a function of ion exchange material and initial cesium concentration. The actual and simulated N-Basin waters contain relatively low levels of aluminum, barium, calcium, potassium, and magnesium (ranging from 8.33E-04 to 6.40E-05 M), with slightly higher levels of boron (6.63E-03 M) and sodium (1.62E-03 M). The {sup 137}Cs level is 1.74E-06 Ci L-{sup 1} which corresponds to approximately 4.87E-10 M Cs. The initial Na/Cs ratio was 3.33E+06. The concentration of total strontium is 4.45E-06 M, while the {sup 90}Sr radioactive component was measured to be 6.13E-06 Ci L{sup -1}. Simulant tests were conducted by contacting 0.067 g or each ion exchange material with approximately 100 mL of either the actual or simulated N-Basin water. The simulants contained variable initial cesium concentrations ranging from 1.00E-04 to 2.57E- 10 M Cs while all other components were held constant. For all materials, the average cesium K{sub d} was independent of cesium concentration below approximately 1.0E-06 M. Above this level, the average cesium K{sub d} values decreased significantly. Cesium K{sub d} values exceeding 1.0E+07 mL g{sup -1} were measured in the simulated N-Basin water. However, when measured in the actual N-Basin water the values were several orders of magnitude lower, with a maximum of 1.24E+05 mL g{sup -1} observed.

Interactive domains in the molecular chaperone human alphaB crystallin modulate microtubule assembly and disassembly.

Directory of Open Access Journals (Sweden)

Joy G Ghosh

2007-06-01

Full Text Available Small heat shock proteins regulate microtubule assembly during cell proliferation and in response to stress through interactions that are poorly understood.Novel functions for five interactive sequences in the small heat shock protein and molecular chaperone, human alphaB crystallin, were investigated in the assembly/disassembly of microtubules and aggregation of tubulin using synthetic peptides and mutants of human alphaB crystallin.The interactive sequence (113FISREFHR(120 exposed on the surface of alphaB crystallin decreased microtubule assembly by approximately 45%. In contrast, the interactive sequences, (131LTITSSLSSDGV(142 and (156ERTIPITRE(164, corresponding to the beta8 strand and the C-terminal extension respectively, which are involved in complex formation, increased microtubule assembly by approximately 34-45%. The alphaB crystallin peptides, (113FISREFHR(120 and (156ERTIPITRE(164, inhibited microtubule disassembly by approximately 26-36%, and the peptides (113FISREFHR(120 and (131LTITSSLSSDGV(142 decreased the thermal aggregation of tubulin by approximately 42-44%. The (131LTITSSLSSDGV(142 and (156ERTIPITRE(164 peptides were more effective than the widely used anti-cancer drug, Paclitaxel, in modulating tubulinmicrotubule dynamics. Mutagenesis of these interactive sequences in wt human alphaB crystallin confirmed the effects of the alphaB crystallin peptides on microtubule assembly/disassembly and tubulin aggregation. The regulation of microtubule assembly by alphaB crystallin varied over a narrow range of concentrations. The assembly of microtubules was maximal at alphaB crystallin to tubulin molar ratios between 1:4 and 2:1, while molar ratios >2:1 inhibited microtubule assembly.Interactive sequences on the surface of human alphaB crystallin collectively modulate microtubule assembly through a dynamic subunit exchange mechanism that depends on the concentration and ratio of alphaB crystallin to tubulin. These are the first

Acid/Caustic Basins: Environmental information document

International Nuclear Information System (INIS)

Ward, J.W.; Johnson, W.F.; Marine, I.W.

1986-12-01

There are six Acid/Caustic Basins at SRP, all of which are located in the reactor and separations areas. These basins are unlined earthen depressions with nominal dimensions of 15.2 m in length x 15.2 m in width x 2.1 m in depth. They were used to provide mixing and partial neutralization of dilute sulfuric acid and sodium hydroxide solutions from water treatment facilities before these solutions were discharged to tributaries of local streams. Closure options considered for the Acid/Caustic Basins are waste removal and closure, no waste removal and closure, and no action. The predominant pathways for human exposure to chemical contaminants are through surface, subsurface, and atmospheric transport. Modeling calculations were made to determine the risks to human population via general pathways for the three postulated closure options. A cost estimate for each closure was also made

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Mechanical Properties of K Basin Sludge Constituents and Their Surrogates

International Nuclear Information System (INIS)

Delegard, Calvin H.; Schmidt, Andrew J.; Chenault, Jeffrey W.

2004-01-01

A survey of the technical literature was performed to summarize the mechanical properties of inorganic components in K Basins sludge. The components included gibbsite, ferrihydrite, lepidocrocite and goethite, hematite, quartz, anorthite, calcite, basalt, Zircaloy, aluminum, and, in particular, irradiated uranium metal and uranium dioxide. Review of the technical literature showed that information on the hardness of uranium metal at irradiation exposures similar to those experienced by the N Reactor fuel present in the K Basins (typically up to 3000 MWd/t) were not available. Measurements therefore were performed to determine the hardness of coupons taken from three irradiated N Reactor uranium metal fuel elements taken from K Basins. Hardness values averaged 30 ± 8 Rockwell C units, similar to values previously reported for uranium irradiated to ∼1200 MWd/t. The physical properties of candidate uranium metal and uranium dioxide surrogates were gathered and compared. Surrogates having properties closest to those of irradiated uranium metal appear to be alloys of tungsten. The surrogate for uranium dioxide, present both as particles and agglomerates in actual K Basin sludge, likely requires two materials. Cerium oxide, CeO2, was identified as a surrogate of the smaller UO2 particles while steel grit was identified for the UO2 agglomerates

The ARCHER project (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D)

Energy Technology Data Exchange (ETDEWEB)

Knol, S., E-mail: knol@nrg.eu [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Fütterer, M.A. [Joint Research Centre, Institute for Energy, Petten (Netherlands); Roelofs, F. [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Kohtz, N. [TÜV Rheinland, Köln (Germany); Laurie, M. [Joint Research Centre, Institute for Transuranium elements, Karlsruhe (Germany); Buckthorpe, D. [UMAN, University of Manchester, Manchester (United Kingdom); Scheuermann, W. [IKE, Stuttgart University, Stuttgart (Germany)

2016-09-15

The European HTR R&D project ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) builds on a solid HTR technology foundation in Europe, established through former national UK and German HTR programs and in European framework programs. ARCHER runs from 2011 to 2015 and targets selected HTR R&D subjects that would specifically support demonstration, with a focus on experimental effort. In line with the R&D and deployment strategy of the European Sustainable Nuclear Energy Technology Platform (SNETP) ARCHER contributes to maintaining, strengthening and expanding the HTR knowledge base in Europe to lay the foundations for demonstration of nuclear cogeneration with HTR systems. The project consortium encompasses conventional and nuclear industry, utilities, Technical Support Organizations, R&D organizations and academia. ARCHER shares results with international partners in the Generation IV International Forum and collaborates directly with related projects in the US, China, Japan, the Republic of Korea and South Africa. The ARCHER project has finished, and the paper comprises an overview of the achievements of the project.

LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

International Nuclear Information System (INIS)

Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

1976-04-01

The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

Polygons and adhesion plaques and the disassembly and assembly of myofibrils in cardiac myocytes.

Science.gov (United States)

Lin, Z X; Holtzer, S; Schultheiss, T; Murray, J; Masaki, T; Fischman, D A; Holtzer, H

1989-06-01

Successive stages in the disassembly of myofibrils and the subsequent assembly of new myofibrils have been studied in cultures of dissociated chick cardiac myocytes. The myofibrils in trypsinized and dispersed myocytes are sequentially disassembled during the first 3 d of culture. They split longitudinally and then assemble into transitory polygons. Multiples of single sarcomeres, the cardiac polygons, are analogous to the transitory polygonal configurations assumed by stress fibers in spreading fibroblasts. They differ from their counterparts in fibroblasts in that they consist of muscle alpha-actinin vertices and muscle myosin heavy chain struts, rather than of the nonmuscle contractile protein isoforms of stress fiber polygons. EM sections reveal the vertices and struts in cardiac polygons to be typical Z and A bands. Most cardiac polygons are eliminated by day 5 of culture. Concurrent with the disassembly and elimination of the original myofibrils new myofibrils are rapidly assembled elsewhere in the same myocyte. Without exception both distal tips of each nascent myofibril terminate in adhesion plaques. The morphology and composition of the adhesion plaques capping each end of each myofibril are similar to those of the termini of stress fibers in fibroblasts. However, whereas the adhesion complexes involving stress fibers in fibroblasts consist of vinculin/nonmuscle alpha-actinin/beta- and gamma-actins, the analogous structures in myocytes involving myofibrils consist of vinculin/muscle alpha-actinin/alpha-actin. The addition of 1.7-2.0 microns sarcomeres to the distal tips of an elongating myofibril, irrespective of whether the myofibril consists of 1, 10, or several hundred tandem sarcomeres, occurs while the myofibril appears to remain linked to its respective adhesion plaques. The adhesion plaques in vitro are the equivalent of the in vivo intercalated discs, both in terms of their molecular composition and with respect to their functioning as initiating

The past and the future in the forty years of the IPR-R1 TRIGA MARK I reactor operation

International Nuclear Information System (INIS)

Maretti Junior, Fausto

2008-01-01

Full text: The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. During these years a lot of irradiations, analysis , MSc and PhD thesis, training courses and isotopes production take place at the reactor. This paper describes the improvements made, the results obtained during the past 40 years, type of works realized, isotopes produced, the neutron activation analysis and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (authors)

GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

International Nuclear Information System (INIS)

Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

2006-01-01

A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

Doping of monocrystalline silicon with phosphorus by means of neutron irradiation at the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carbonari, A.W.; Puget, M.A.C.

1990-11-01

The first neutron irradiation experiments with monocrystal silicon in the IEA-R1 research reactor of IPEN are related. The silicon is irradiated with phosphorus producing a N type semiconductor with a very small resistivity variation throughout the crystal volume. The neutrons induce nuclear reactions in Si-30 isotope and these atoms are then transformed in to phosphorous atoms. This process is known as Neutron Transmutation Doping. In order to irradiate the silicon crystals in the reactor, a specific device has been constructed, and it permits the irradiation of up to 2.5'' diameter monocrystals. (author)

A total Ammonium Reactor (NHxR) for In Situ Mobile Measurements: A Critical Tool to Understand Aerosol Formation, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — We will develop, demonstrate, and optimize a front-end ammonium reactor (NHxR) for the fast, precise, and accurate measurement of gas-phase ammonia (NH3) and...

Conveyor system bolt failure analysis

International Nuclear Information System (INIS)

McCutcheon, S.H.; Waaser, R.E.

1981-07-01

The Savannah River Plant (SRP) is operated for the U.S. Department of Energy (DOE) by E.I. du Pont de Nemours and Co. Reactors at SRP are used to produce nuclear isotopes for national defense. As part of the routine operation of each of the reactors, irradiated fuel assemblies are remotely discharged from the reactor, transferred in air, and placed underwater in a Deposit and Exit (D and E) canal. The D and E conveyor transports the assemblies under the reactor room wall into the disassembly basin, where they are stored for several months before shipment to the separations areas. On August 13, 1980, a broken bolt was discovered on the D and E conveyor during a routine pre-operational inspection. The failed bolt was one of four which anchor a bearing on one of the conveyor drive shafts. All four of the bolts on the affected bearing were replaced as an immediate corrective measure and the discharge of irradiated fuel assemblies was subsequently completed without difficulty

Feasibility studies of producing 99 Mo by capture in the IEA-R1 research reactor

International Nuclear Information System (INIS)

Concilio, Roberta; Mendonca, Arlindo Gilson; Maiorino, Jose Rubens

1998-01-01

Everyday the production of 99 Mo for 99m Tc generators, becomes more necessary, whose properties are ideal for medical diagnosis. This works presents a description and an analysis of the production of 99 Mo by radioactive capture at 98 Mo using the research reactor IEA-R1 in 5 MW and operating 5 days a week, referring to the use of targets, separation methods, total and specific activity attained and its limitations. (author)

Commercial US nuclear reactors and waste: the current status

International Nuclear Information System (INIS)

Platt, A.M.; Robinson, J.V.

1980-09-01

Between March 1 and June 15, 1980, the declared size of the commercial light waste reactor (LWR) nuclear power industry in the US has decreased another 9 GWe. For the presently declared size: the 165 declared reactors will peak at a capacity of 153 GWe in 2001 and will consume about 870,000 MTU as enrichment feed; the theoretical rate of enrichment requirements will peak at about 19,000,000 SWUs/y in the year 2014; as few as two repositories each with capacity equivalent to 100,000 MTU would hold the waste; and predisposal storage reactor basins and AFRs (away-from-reactor basins) would peak at <85,000 MTU in the year 2020 if the two respositories were commissioned in the years 1997 and 2020. It should be noted that the number of declared LWRs has dropped from 226 on December 31, 1974 to 165 as of this writing. The oil equivalent of the energy loss, assuming a 50% efficiency in use as in cars, is 17,000 million barrels. This is about 10 years of the current rate of US consumption of OPEC oil

Commercial US nuclear reactors and waste: the current status

Energy Technology Data Exchange (ETDEWEB)

Platt, A.M.; Robinson, J.V.

1980-09-01

Between March 1 and June 15, 1980, the declared size of the commercial light waste reactor (LWR) nuclear power industry in the US has decreased another 9 GWe. For the presently declared size: the 165 declared reactors will peak at a capacity of 153 GWe in 2001 and will consume about 870,000 MTU as enrichment feed; the theoretical rate of enrichment requirements will peak at about 19,000,000 SWUs/y in the year 2014; as few as two repositories each with capacity equivalent to 100,000 MTU would hold the waste; and predisposal storage reactor basins and AFRs (away-from-reactor basins) would peak at <85,000 MTU in the year 2020 if the two respositories were commissioned in the years 1997 and 2020. It should be noted that the number of declared LWRs has dropped from 226 on December 31, 1974 to 165 as of this writing. The oil equivalent of the energy loss, assuming a 50% efficiency in use as in cars, is 17,000 million barrels. This is about 10 years of the current rate of US consumption of OPEC oil.

R&D on high-power dc reactor prototype for ITER poloidal field converter

Energy Technology Data Exchange (ETDEWEB)

Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

2015-10-15

Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

Design bases for U.S. Department of Energy storage basin

International Nuclear Information System (INIS)

King, F.D.

1978-01-01

Bases were developed for the conceptual design of a water basin to store 5000 MT of irradiated light water reactor fuel. The basin, which could be operational by 1983, is designed to receive over 2000 MT per year of spent fuel and to provide interim storage of US and some foreign fuel in support of U.S. policy of receiving and taking title to spent fuel rather than reprocessing this fuel. The bases for various design decisions are discussed

Characterization of the disassembly and reassembly of the HBV glycoprotein surface antigen, a pliable nanoparticle vaccine platform

International Nuclear Information System (INIS)