WorldWideScience

Sample records for nuclear process off-gas

  1. Monitoring off-gas O2/CO2 to predict nitrification performance in activated sludge processes.

    Science.gov (United States)

    Leu, Shao-Yuan; Libra, Judy A; Stenstrom, Michael K

    2010-06-01

    Nitrification/denitrification (NDN) processes are the most widely used technique to remove nitrogenous pollutants from municipal wastewater. The performance of nitrogen removal in the NDN process depends on the metabolism of nitrifying bacteria, and is dependent on adequate oxygen supply. Off-gas testing is a convenient and popular method for measuring oxygen transfer efficiency (OTE) under process conditions and can be performed in real-time. Since carbon dioxide is produced by carbonaceous oxidizing organism and not by nitrifiers, it should be possible to use the off-gas carbon dioxide mole fraction to estimate nitrification performance independently of the oxygen uptake rate (OUR) or OTE. This paper used off-gas data with a dynamic model to estimate nitrifying efficiency for various activated sludge process conditions. The relationship among nitrification, oxygen transfer, carbon dioxide production, and pH change was investigated. Experimental results of an online off-gas monitoring for a full-scale treatment plant were used to validate the model. The results showed measurable differences in OUR and carbon dioxide transfer rate (CTR) and the simulations successfully predicted the effluent ammonia by using the measured CO(2) and O(2) contents in off-gas as input signal. Carbon dioxide in the off-gas could be a useful technique to control aeration and to monitor nitrification rate.

  2. Selective Recovery of Radioactive Carbon Dioxide Released from Nuclear Off-gas by Adsorption

    Science.gov (United States)

    Munakata, Kenzo; Koga, Akinori

    Off gases produced in the reprocessing of spent nuclear fuel contain various radioactive gases and emission of these gases to the environment must be suppressed as low as possible. 14C with a long half-life, which is mainly released as the form of carbon dioxide, is one of such gaseous radioactive materials. One of the measures to capture radioactive gases from the off-gas is the utilization of adsorption technique. In this work, the adsorption behavior of carbon dioxide on various adsorbents was studied. It was found that a MS4A (Molecular Sieve 4A) adsorbent is more suitable for selective recovery of carbon dioxide. Thus, more detailed adsorption characteristics of carbon dioxide were studied for a MS4A adsorbent. Moreover, the authors investigated the influence of coexistent water vapor, which is also contained in the off-gas, on the adsorption behavior of carbon dioxide.

  3. Computer simulation of the off gas treatment process for the KEPCO pilot vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hey Suk; Maeng, Sung Jun; Lee, Myung Chan [Nuclear Environment Technology Institute, KEPCO, Taejon (Korea, Republic of)

    1999-07-01

    Vitrification technology for treatment of low and intermediate radioactive wastes can remarkably reduce waste volume to about one twentieth of the initial volume as they are collected and converted into a very stable form. Therefore, it can minimize environmental impact when the vitrified waste is disposed of. But an off gas treatment system is necessary to apply this technology because air pollutants and radioisotopes are generated like those of other conventional incinerators during thermal oxidation process at high temperature. KEPCO designed and installed a pilot scale vitrification plant to demonstrate the feasibility of the vitrification process and then to make a conceptual design for a commercial vitrification facility. The purpose of this study was to simulate the off gas treatment system(OGTS) in order optimize the operating conditions. Mass balance and temperature profile in the off gas treatment system were simulated for different combinations of combustible wastes by computer simulation code named OGTS code and removal efficiency of each process was also calculated with change of design parameters. The OGTS code saved efforts,time and capital because scale and configuration of the system could be easily changed. The simulation result of the pilot scale off gas process as well as pilot tests will be of great use in the future for a design of the commercial vitrification facility. (author)

  4. Avoiding Carbon Bed Hot Spots in Thermal Process Off-Gas Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Joe Enneking

    2011-05-01

    Mercury has had various uses in nuclear fuel reprocessing and other nuclear processes, and so is often present in radioactive and mixed (radioactive and hazardous) wastes. Test programs performed in recent years have shown that mercury in off-gas streams from processes that treat radioactive wastes can be controlled using fixed beds of activated sulfur-impregnated carbon, to levels low enough to comply with air emission regulations such as the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards. Carbon bed hot spots or fires have occurred several times during these tests, and also during a remediation of tanks that contained mixed waste. Hot spots occur when localized areas in a carbon bed become heated to temperatures where oxidation occurs. This heating typically occurs due to heat of absoption of gas species onto the carbon, but it can also be caused through external means such as external heaters used to heat the carbon bed vessel. Hot spots, if not promptly mitigated, can grow into bed fires. Carbon bed hot spots and fires must be avoided in processes that treat radioactive and mixed waste. Hot spots are detected by (a) monitoring in-bed and bed outlet gas temperatures, and (b) more important, monitoring of bed outlet gas CO concentrations. Hot spots are mitigated by (a) designing for appropriate in-bed gas velocity, for avoiding gas flow maldistribution, and for sufficient but not excessive bed depth, (b) appropriate monitoring and control of gas and bed temperatures and compositions, and (c) prompt implementation of corrective actions if bed hot spots are detected. Corrective actions must be implemented quickly if bed hot spots are detected, using a graded approach and sequence starting with corrective actions that are simple, quick, cause the least impact to the process, and are easiest to recover from.

  5. Avoiding Carbon Bed Hot Spots in Thermal Process Off-Gas Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Joe Enneking

    2011-05-01

    Mercury has had various uses in nuclear fuel reprocessing and other nuclear processes, and so is often present in radioactive and mixed (radioactive and hazardous) wastes. Test programs performed in recent years have shown that mercury in off-gas streams from processes that treat radioactive wastes can be controlled using fixed beds of activated sulfur-impregnated carbon, to levels low enough to comply with air emission regulations such as the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards. Carbon bed hot spots or fires have occurred several times during these tests, and also during a remediation of tanks that contained mixed waste. Hot spots occur when localized areas in a carbon bed become heated to temperatures where oxidation occurs. This heating typically occurs due to heat of absoption of gas species onto the carbon, but it can also be caused through external means such as external heaters used to heat the carbon bed vessel. Hot spots, if not promptly mitigated, can grow into bed fires. Carbon bed hot spots and fires must be avoided in processes that treat radioactive and mixed waste. Hot spots are detected by (a) monitoring in-bed and bed outlet gas temperatures, and (b) more important, monitoring of bed outlet gas CO concentrations. Hot spots are mitigated by (a) designing for appropriate in-bed gas velocity, for avoiding gas flow maldistribution, and for sufficient but not excessive bed depth, (b) appropriate monitoring and control of gas and bed temperatures and compositions, and (c) prompt implementation of corrective actions if bed hot spots are detected. Corrective actions must be implemented quickly if bed hot spots are detected, using a graded approach and sequence starting with corrective actions that are simple, quick, cause the least impact to the process, and are easiest to recover from.

  6. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    Energy Technology Data Exchange (ETDEWEB)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-10-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  7. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    Energy Technology Data Exchange (ETDEWEB)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-09-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  8. Glass melter off-gas system pluggages: Cause, significance, and remediation

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.

    1991-03-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF) where the glass will be poured into stainless steel canisters for eventual disposal in a geologic repository. Experimental glass melters used to develop the vitrification process for immobilization of the waste have experienced problems with pluggage of the off-gas line with solid deposits. Off-gas deposits from the DWPF 1/2 Scale Glass Melter (SGM) and the 1/10th scale Integrated DWPF Melter System (IDMS) were determined to be mixtures of alkali rich chlorides, sulfates, borates, and fluorides with entrained Fe{sub 2}O{sub 3}, spinel, and frit particles. The distribution and location of the alkali deposits throughout the off-gas system indicate that the deposits form by vapor-phase transport and condensation. Condensation of the alkali-rich phases cement the entrained particulates causing off-gas system pluggages. The identification of vapor phase transport as the operational mechanism causing off-gas system pluggage indicates that deposition can be effectively eliminated by increasing the off-gas velocity. Scale glass melter operating experience indicates that a velocity of >50 fps is necessary in order to transport the volatile species to the quencher to prevent having condensation occur in the off-gas line. Hotter off-gas line temperatures would retain the alkali compounds as vapors so that they would remain volatile until they reach the quencher. However, hotter off-gas temperatures can only be achieved by using less air/steam flow at the off-gas entrance, e.g. at the off-gas film cooler (OGFC). This would result in lower off-gas velocities. Maintaining a high velocity is, therefore, considered to be a more important criterion for controlling off-gas pluggage than temperature control. 40 refs., 16 figs., 5 tabs.

  9. Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.; Taylor-Pashow, Kathryn M.; Adamson, Duane J.; Crawford, Charles L.; Morse, Megan M.

    2014-01-07

    The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion

  10. Adsorption Model for Off-Gas Separation

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J. Rutledge

    2011-03-01

    The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior and feedback loops. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes will provide substantial cost savings and many technical benefits. The specific fuel cycle separation process discussed in this report is the off-gas treatment system. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed in gPROMS software. Inputs include gas stream constituents, sorbent, and column properties, equilibrium and kinetic data, and inlet conditions. It models dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions for a multiple component gas stream. The simulation outputs component concentrations along the column length as a function of time from which the breakthrough data is obtained. It also outputs temperature along the column length as a function of time and pressure drop along the column length. Experimental data will be input into the adsorption model to develop a model specific for iodine adsorption on silver mordenite as well as model(s) specific for krypton and xenon adsorption. The model will be validated with experimental breakthrough curves. Another future off-gas modeling goal is to develop a model for the unit operation absorption. The off-gas models will be made available via the server or web for evaluation by customers.

  11. Direct chlorination process for geothermal power plant off-gas - hydrogen sulfide abatement

    Energy Technology Data Exchange (ETDEWEB)

    Sims, A.V.

    1983-06-01

    The Direct Chlorination Process removes hydrogen sulfide from geothermal off-gases by reacting hydrogen sulfide with chlorine in the gas phase. Hydrogen chloride and elemental sulfur are formed by this reaction. The Direct Chlorination Process has been successfully demonstrated by an on-site operation of a pilot plant at the 3 M We HPG-A geothermal power plant in the Puna District on the island of Hawaii. Over 99.5 percent hydrogen sulfide removal was achieved in a single reaction state. Chlorine gas did not escape the pilot plant, even when 90 percent excess chlorine gas was used. A preliminary economic evaluation of the Direct Chlorination Process indicates that it is very competitive with the Stretford Process. Compared to the Stretford Process, the Direct Chlorination Process requires about one-third the initial capital investment and about one-fourth the net daily expenditure.

  12. Direct Chlorination Process for geothermal power plant off-gas - hydrogen sulfide abatement

    Energy Technology Data Exchange (ETDEWEB)

    Sims, A.V.

    1983-06-01

    The Direct Chlorination Process removes hydrogen sulfide from geothermal off-gases by reacting hydrogen sulfide with chlorine in the gas phase. Hydrogen chloride and elemental sulfur are formed by this reaction. The Direct Chlorination Process has been successfully demonstrated by an on-site operation of a pilot plant at the 3 M We HPG-A geothermal power plant in the Puna District on the island of Hawaii. Over 99.5% hydrogen sulfide removal was achieved in a single reaction stage. Chlorine gas did not escape the pilot plant, even when 90% excess chlorine gas was used. A preliminary economic evaluation of the Direct Chlorination Process indicates that it is very competitive with the Stretford Process Compared to the Stretford Process, the Direct Chlorination process requires about one-third the initial capital investment and about one-fourth the net daily expenditure. Because of the higher cost of chemicals and the restricted markets in Hawaii, the economic viability of this process in Hawaii is questionable.

  13. Direct chlorination process for geothermal power plant off-gas - hydrogen sulfide abatement

    Energy Technology Data Exchange (ETDEWEB)

    Sims, A.V.

    1983-06-01

    The Direct Chlorination Process removes hydrogen sulfide from geothermal off-gases by reacting hydrogen sulfide with chlorine in the gas phase. Hydrogen chloride and elemental sulfur are formed by this reaction. The Direct Chlorination Process has been successfully demonstrated by an on-site operation of a pilot plant at the 3 M We HPG-A geothermal power plant in the Puna District on the island of Hawaii. Over 99.5 percent hydrogen sulfide removal was achieved in a single reaction stage. Chlorine gas did not escape the pilot plant, even when 90 percent excess chlorine gas was used. Because of the higher cost of chemicals and the restricted markets in Hawaii, the economic viability of this process in Hawaii is questionable.

  14. Hydroxylamine a potential reagent for dissolution off gas scrubbing in nuclear spent fuel reprocessing: kinetics of the iodine reduction

    Energy Technology Data Exchange (ETDEWEB)

    Cau Dit Coumes, C.; Devisme, F. [CEA Centre d`Etudes de la Vallee du Rhone, 30 - Marcoule (France). Dept. d`Exploitation du Retraitement et de Demantelement; Chopin, J.; Vargas, S.

    1996-12-31

    Iodine, which can be released inside the containment buildings when accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a regent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH(1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30 deg. C) and ionic strength (0.1 mol/l). Spectrophotometry and voltametry have been coupled for analytical solved using a investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Tri-iodine has been shown non reactive towards hydroxylamine. An initial rate law have been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of hour reactions (I{sub 2} + I <=> I{sub 3}{sup -}NH{sub 3}OH{sup +} + 2 I{sub 2} + H{sub 2}O ->HNO{sub 2} + 4 I{sup -} + 5 H{sup +}; NH{sub 3}OH{sup +} + HNO{sub 2} -> N{sub 2}O + 2 H{sub 2}O + H-+ 2HNO{sub 2} + 2 I{sup -} + 2H-+ -> 2 NO + I{sub 2} + H{sub 2}O). (authors). 12 refs.

  15. Organic Iodine Adsorption by AgZ under Prototypical Vessel Off-Gas Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, J. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    U.S. regulations will require the removal of 129I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. Multiple off-gas streams within a UNF reprocessing plant combine prior to release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas and the shear off-gas. To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Preliminary studies have been completed on the adsorption of I2 onto silver mordenite (AgZ) from prototypical VOG streams. The study reported that AgZ did adsorb I2 from a prototypical VOG stream, but process upsets resulted in an uneven feed stream concentration. The experiments described in this document both improve the characterization of I2 adsorption by AgZ from dilute gas streams and further extend it to include characterization of the adsorption of organic iodides (in the form of CH3I) onto AgZ under prototypical VOG conditions. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the effect of sorbent aging on iodine removal in VOG conditions could be inferred.

  16. A state-of-the-art report on the off-gas treatment technology generated from the nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Won Zin; Lee, Hoo Geun; Park, Geun Il; Choi, Byung Seon; Lee, Kee Won; Cho, Il Hoon; Kim, Jung Kook; Park, Hyun Soo

    1997-10-01

    This state-of-the-art report describes various technologies for offgas treatment. It provides comprehensive treatment technologies of the extensive subject such as particulates, radioactive iodine, carbon dioxide, Kr/Xe and Cs/Ru. This report also incorporates the wastes generation and its characteristics as well as the historical and current management practices. A number of review articles by experts in various area of concern and some of the removal systems that have been designed for power plants and, particularly, for spent fuel reusing plants are also involved. As a result, it can be drawn that the drying processes for offgas treatment have much benefits in standpoints of simplicity, economy, disposal safety and resource reuse rather than the wet processes. (author). 226 refs., 38 tabs., 44 figs

  17. MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM - PRELIMINARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Choi, A.

    2009-03-25

    The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that come in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter off-gas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl{sub 2}, and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg{sub 2}Cl{sub 2}) to HgCl{sub 2} with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of

  18. Process system evaluation-consolidated letters. Volume 1. Alternatives for the off-gas treatment system for the low-level waste vitrification process

    Energy Technology Data Exchange (ETDEWEB)

    Peurrung, L.M.; Deforest, T.J; Richards, J.R.

    1996-03-01

    This report provides an evaluation of alternatives for treating off-gas from the low-level waste (LLW) melter. The study used expertise obtained from the commercial nonradioactive off-gas treatment industry. It was assumed that contact maintenance is possible, although the subsequent risk to maintenance personnel was qualitatively considered in selecting equipment. Some adaptations to the alternatives described may be required, depending on the extent of contact maintenance that can be achieved. This evaluation identified key issues for the off-gas system design. To provide background information, technology reviews were assembled for various classifications of off-gas treatment equipment, including off-gas cooling, particulate control, acid gas control, mist elimination, NO{sub x} reduction, and SO{sub 2} removal. An order-of-magnitude cost estimate for one of the off-gas systems considered is provided using both the off-gas characteristics associated with the Joule-heated and combustion-fired melters. The key issues identified and a description of the preferred off-gas system options are provided below. Five candidate treatment systems were evaluated. All of the systems are appropriate for the different melting/feed preparations currently being considered. The lowest technical risk is achieved using option 1, which is similar to designs for high-level waste (HLW) vitrification in the Hanford Waste Vitrification Project (HWVP) and the West Valley. Demonstration Project. Option 1 uses a film cooler, submerged bed scrubber (SBS), and high-efficiency mist eliminator (HEME) prior to NO{sub x} reduction and high-efficiency particulate air (HEPA) filtration. However, several advantages were identified for option 2, which uses high-temperature filtration. Based on the evaluation, option 2 was identified as the preferred alternative. The characteristics of this option are described below.

  19. MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM-PRELIMINARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Choi, A.

    2010-08-18

    , there are many process benefits to be gained by removing the steam-stripping step from the CPC cycle. The goal of this task was to study what adverse impact the zero-mercury-removal scenario would have on the DWPF melter off-gas system operation. It is stressed again that this study was intended to be scoping in nature, so the results presented in this report are preliminary. Any further substantiation of these results for actual implementation into the DWPF flowsheet would require an in-depth modeling study of all three reaction zones, including the aqueous-phase reactions in the quencher, OGCT, Steam Atomized Scrubber (SAS), and off-gas condenser with recirculated condensate, and the proof-of-principle experiments.

  20. Antifoam Degradation Products in Off Gas and Condensate of Sludge Batch 9 Simulant Nitric-Formic Flowsheet Testing for the Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Smith, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-14

    Ten chemical processing cell (CPC) experiments were performed using simulant to evaluate Sludge Batch 9 for sludge-only and coupled processing using the nitric-formic flowsheet in the Defense Waste Processing Facility (DWPF). Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on eight of the ten. The other two were SRAT cycles only. Samples of the condensate, sludge, and off gas were taken to monitor the chemistry of the CPC experiments. The Savannah River National Laboratory (SRNL) has previously shown antifoam decomposes to form flammable organic products, (hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and propanal), that are present in the vapor phase and condensate of the CPC vessels. To minimize antifoam degradation product formation, a new antifoam addition strategy was implemented at SRNL and DWPF to add antifoam undiluted.

  1. Development and Testing of the Advanced CHP System Utilizing the Off-Gas from the Innovative Green Coke Calcining Process in Fluidized Bed

    Energy Technology Data Exchange (ETDEWEB)

    Chudnovsky, Yaroslav; Kozlov, Aleksandr

    2013-08-15

    Green petroleum coke (GPC) is an oil refining byproduct that can be used directly as a solid fuel or as a feedstock for the production of calcined petroleum coke. GPC contains a high amount of volatiles and sulfur. During the calcination process, the GPC is heated to remove the volatiles and sulfur to produce purified calcined coke, which is used in the production of graphite, electrodes, metal carburizers, and other carbon products. Currently, more than 80% of calcined coke is produced in rotary kilns or rotary hearth furnaces. These technologies provide partial heat utilization of the calcined coke to increase efficiency of the calcination process, but they also share some operating disadvantages. However, coke calcination in an electrothermal fluidized bed (EFB) opens up a number of potential benefits for the production enhancement, while reducing the capital and operating costs. The increased usage of heavy crude oil in recent years has resulted in higher sulfur content in green coke produced by oil refinery process, which requires a significant increase in the calcinations temperature and in residence time. The calorific value of the process off-gas is quite substantial and can be effectively utilized as an “opportunity fuel” for combined heat and power (CHP) production to complement the energy demand. Heat recovered from the product cooling can also contribute to the overall economics of the calcination process. Preliminary estimates indicated the decrease in energy consumption by 35-50% as well as a proportional decrease in greenhouse gas emissions. As such, the efficiency improvement of the coke calcinations systems is attracting close attention of the researchers and engineers throughout the world. The developed technology is intended to accomplish the following objectives: - Reduce the energy and carbon intensity of the calcined coke production process. - Increase utilization of opportunity fuels such as industrial waste off-gas from the novel

  2. Adsorption modeling for off-gas treatment

    Energy Technology Data Exchange (ETDEWEB)

    Ladshaw, A.; Sharma, K.; Yiacoumi, S.; Tsouris, C. [Georgia Institute of Technology, Atlanta, GA 30332-0459 (United States); De Paoli, D.W. [Oak Ridge National Laboratory: Oak Ridge, TN 37831-6181 (United States)

    2013-07-01

    Off-gas generated from the reprocessing of used nuclear fuel contains a mixture of several radioactive gases including {sup 129}I{sub 2}, {sup 85}Kr, HTO, and {sup 14}CO{sub 2}. Over the past few decades, various separation and recovery processes have been studied for capturing these gases. Adsorption data for gaseous mixtures of species can be difficult to determine experimentally. Therefore, procedures capable of predicting the adsorption behavior of mixtures need to be developed from the individual isotherms of each of the pure species. A particular isotherm model of interest for the pure species is the Generalized Statistical Thermodynamic Adsorption isotherm. This model contains an adjustable number of parameters and will therefore describe a wide range of adsorption isotherms for a variety of components. A code has been developed in C++ to perform the non-linear regression analysis necessary for the determination of the isotherm parameters, as well as the least number of parameters needed to describe an entire set of data. (authors)

  3. Off-gas Adsorption Model and Simulation - OSPREY

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J Rutledge

    2013-10-01

    The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and REcoverY (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed.

  4. 聚丙烯装置尾气回收利用技术%Conversion of the off Gas from Polypropylene Process

    Institute of Scientific and Technical Information of China (English)

    骞伟中; 汪展文; 魏飞; 金涌; 张吉瑞

    2001-01-01

    In the process of producing polypropylene,there is 10%~20% propylene in the off gas.The technology of catalytic distillation is introduced to absorb the propyle ne.Operating at 0.6MPa,the ratio of nitrogen vs.propylene is 1∶1,the selectivi ty of cumene is more than 96%,the conversion of propylene is more than 90%.The s pace velocity of propylene is 1.5~2.5h-1.It was proved that the technol og y of catalytic distillation is hopeful for the conversion of propylene in the of f gas from the polypropylene process.%在聚丙烯的生产过程中,尾气中含有较多的丙烯。本文尝试利用催化精馏技术解决尾气中丙 烯的吸收问题,将苯与丙烯在催化剂上反应生成异丙苯。初步研究表明,使用氮气与丙烯体 积比为1∶1的混合气时,可使丙烯转化率大于90%,所得产品中异丙苯的选择性大于96%,丙 烯的质量空速可达1.5~2.5h-1。证明催化精馏技术是此方面很有前途的工程技术 之一。

  5. Methods of Off-Gas Flammability Control for DWPF Melter Off-Gas System at Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.S. [Westinghouse Savannah River Company, AIKEN, SC (United States); Iverson, D.C.

    1996-05-02

    Several key operating variables affecting off-gas flammability in a slurry-fed radioactive waste glass melter are discussed, and the methods used to prevent potential off-gas flammability are presented. Two models have played a central role in developing such methods. The first model attempts to describe the chemical events occurring during the calcining and melting steps using a multistage thermodynamic equilibrium approach, and it calculates the compositions of glass and calcine gases. Volatile feed components and calcine gases are fed to the second model which then predicts the process dynamics of the entire melter off-gas system including off-gas flammability under both steady state and various transient operating conditions. Results of recent simulation runs are also compared with available data

  6. Sorption Modeling and Verification for Off-Gas Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Tavlarides, Lawrence [Syracuse Univ., NY (United States); Yiacoumi, Sotira [Georgia Inst. of Technology, Atlanta, GA (United States); Tsouris, Costas [Georgia Inst. of Technology, Atlanta, GA (United States); Gabitto, Jorge [Prairie View Texas A& M; DePaoli, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-20

    This project was successfully executed to provide valuable adsorption data and improve a comprehensive model developed in previous work by the authors. Data obtained were used in an integrated computer program to predict the behavior of adsorption columns. The model is supported by experimental data and has been shown to predict capture of off gas similar to that evolving during the reprocessing of nuclear waste. The computer program structure contains (a) equilibrium models of off-gases with the adsorbate; (b) mass-transfer models to describe off-gas mass transfer to a particle, diffusion through the pores of the particle, and adsorption on the active sites of the particle; and (c) incorporation of these models into fixed bed adsorption modeling, which includes advection through the bed. These models are being connected with the MOOSE (Multiphysics Object-Oriented Simulation Environment) software developed at the Idaho National Laboratory through DGOSPREY (Discontinuous Galerkin Off-gas SeParation and REcoverY) computer codes developed in this project. Experiments for iodine and water adsorption have been conducted on reduced silver mordenite (Ag0Z) for single layered particles. Adsorption apparatuses have been constructed to execute these experiments over a useful range of conditions for temperatures ranging from ambient to 250°C and water dew points ranging from -69 to 19°C. Experimental results were analyzed to determine mass transfer and diffusion of these gases into the particles and to determine which models best describe the single and binary component mass transfer and diffusion processes. The experimental results were also used to demonstrate the capabilities of the comprehensive models developed to predict single-particle adsorption and transients of the adsorption-desorption processes in fixed beds. Models for adsorption and mass transfer have been developed to mathematically describe adsorption kinetics and transport via diffusion and advection

  7. HC-21C off-gas test procedure. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, L.T.

    1995-02-02

    Stabilization of plutonium bearing scrap material occurs in furnaces, FUR-21C-1 and FUR-21C-2, located in glovebox HC-21C. During previous testing and processing operations, water has been observed forming in the off-gas rotameters, FI-21C-1 and FI-21C-2. The off-gas is filtered through a 2 micron ceramic filter, F-21C-1 or F-21C-2, before discharge into the 26 inch vacuum system. The goal of this test plan is to determine the cause and location of water formation in the sludge stabilization off-gas system. The results should help determine what design improvements or processing steps will be implemented to prevent this phenomena from occurring in the future.

  8. HC-21C off-gas test procedure

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, L.T.

    1994-12-14

    The goal of this test plan is to determine the cause and location of water formation in the sludge stabilization off-gas system. The results should help determine what design improvements or processing steps will be implemented to prevent this phenomena from occurring in the future. This test procedure will include a series of tests to determine where and why liquid is condensing in the HC-21C furnace off-gas system. The tests will take a sequential, graded approach and may be concluded one the results have satisfactorily resolved the problem.

  9. CHARACTERIZATION OF DWPF MELTER OFF-GAS QUENCHER SAMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Newell, J.

    2011-11-14

    The Savannah River National Laboratory (SRNL) recently received a deposit sample from the Melter Primary Off Gas System (POG) of the Defense Waste Processing Facility (DWPF). This sample was composed of material that had been collected while the quencher was in operation January 27, 2011 through March 31, 2011. DWPF requested, through a technical assistance request, characterization of the melter off-gas deposits by x-ray diffraction (XRD), scanning electron microscopy (SEM), and chemical analysis. The purpose of the Melter Off-Gas System is to reduce the amount of radioactive particles and mercury in the gases vented to the atmosphere. Gases emitted from the melter pass through the primary film cooler, quencher, Off-Gas Condensate Tank (OGCT), Steam Atomized Scrubbers (SAS), a condenser, a high efficiency mist eliminator, and a high efficiency particulate air filter, before being vented to the Process Vessel Vent System. The film coolers cool the gases leaving the melter vapor space from {approx}750 C to {approx}375 C, by introducing air and steam to the flow. In the next step, the quencher cools the gas to about 60 C by bringing the condensate from the OGCT in contact with the effluent (Figure 1). Most of the steam in the effluent is then condensed and the melter vapor space pressure is reduced. The purpose of the OGCT is to collect and store the condensate formed during the melter operation. Condensate from the OGCT is circulated to the SAS and atomized with steam. This atomized condensate is mixed with the off-gas to wet and join the particulate which is then removed in the cyclone. The next stage incorporates a chilled water condenser which separates the vapors and elemental mercury from the off-gas steam. Primary off-gas deposit samples from the DWPF melter have previously been analyzed. In 2003, samples from just past the film cooler, from the inlet of the quencher and inside the quencher were analyzed at SRNL. It was determined that the samples were a

  10. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

  11. Sorption Modeling and Verification for Off-Gas Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Tavlarides, Lawrence L. [Syracuse Univ., NY (United States); Lin, Ronghong [Syracuse Univ., NY (United States); Nan, Yue [Syracuse Univ., NY (United States); Yiacoumi, Sotira [Georgia Inst. of Technology, Atlanta, GA (United States); Tsouris, Costas [Georgia Inst. of Technology, Atlanta, GA (United States); Ladshaw, Austin [Georgia Inst. of Technology, Atlanta, GA (United States); Sharma, Ketki [Georgia Inst. of Technology, Atlanta, GA (United States); Gabitto, Jorge [Prairie View A & M Univ., Prairie View, TX (United States); DePaoli, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-29

    The project has made progress toward developing a comprehensive modeling capability for the capture of target species in off gas evolved during the reprocessing of nuclear fuel. The effort has integrated experimentation, model development, and computer code development for adsorption and absorption processes. For adsorption, a modeling library has been initiated to include (a) equilibrium models for uptake of off-gas components by adsorbents, (b) mass transfer models to describe mass transfer to a particle, diffusion through the pores of the particle and adsorption on the active sites of the particle, and (c) interconnection of these models to fixed bed adsorption modeling which includes advection through the bed. For single-component equilibria, a Generalized Statistical Thermodynamic Adsorption (GSTA) code was developed to represent experimental data from a broad range of isotherm types; this is equivalent to a Langmuir isotherm in the two-parameter case, and was demonstrated for Kr on INL-engineered sorbent HZ PAN, water sorption on molecular sieve A sorbent material (MS3A), and Kr and Xe capture on metal-organic framework (MOF) materials. The GSTA isotherm was extended to multicomponent systems through application of a modified spreading pressure surface activity model and generalized predictive adsorbed solution theory; the result is the capability to estimate multicomponent adsorption equilibria from single-component isotherms. This advance, which enhances the capability to simulate systems related to off-gas treatment, has been demonstrated for a range of real-gas systems in the literature and is ready for testing with data currently being collected for multicomponent systems of interest, including iodine and water on MS3A. A diffusion kinetic model for sorbent pellets involving pore and surface diffusion as well as external mass transfer has been established, and a methodology was developed for determining unknown diffusivity parameters from transient

  12. Logic Interlock Setting for Off-gas Incineration in Process Water Stripper Facility%工艺水汽提装置尾气燃烧逻辑联锁的设置

    Institute of Scientific and Technical Information of China (English)

    项飞

    2013-01-01

    To dispose waste water and exhaust gas from PET plant, a process water stripper facility is put up in the original process. The process waste water flows from the top to the bottom through the stripper column. Low pressure steam is blown counter-current through the process water stripper column at the same time. The treated off-gas containing organics is sent to HTM heater for incineration. The waste water is drained to wastewater treatment facility. Detailed flowchart of process water stripper facility is introduced, and relevant logic interlock setting for ensuring plant safe production requirements are evaluated. It is expressed with the form of logic flowchart conforming to ANSI/ISA-5. 2-1976 standard.%为处理聚酯装置的生产尾气及生产废水,在原流程基础上增设工艺水汽提工段,工艺废水经汽提塔自上而下处理,同时汽提塔塔底通入低压蒸汽,处理后的含有有机物的尾气送入热媒炉燃烧,废水直接排放至污水处理.详细介绍了聚酯装置工艺水汽提工段的流程概貌,并分析为保证生产安全需要设置的相关逻辑联锁,最终以符合ANSI/ISA-5.2-1976标准的逻辑图的形式对其进行表述.

  13. Study of plasma off-gas treatment from spent ion exchange resin pyrolysis.

    Science.gov (United States)

    Castro, Hernán Ariel; Luca, Vittorio; Banchi, Hugo Luis

    2017-03-23

    Polystyrene divinylbenzene-based ion exchange resins are employed extensively within nuclear power plants (NPPs) and research reactors for purification and chemical control of the cooling water system. To maintain the highest possible water quality, the resins are regularly replaced as they become contaminated with a range of isotopes derived from compromised fuel elements as well as corrosion and activation products including (14)C, (60)Co, (90)Sr, (129)I, and (137)Cs. Such spent resins constitute a major proportion (in volume terms) of the solid radioactive waste generated by the nuclear industry. Several treatment and conditioning techniques have been developed with a view toward reducing the spent resin volume and generating a stable waste product suitable for long-term storage and disposal. Between them, pyrolysis emerges as an attractive option. Previous work of our group suggests that the pyrolysis treatment of the resins at low temperatures between 300 and 350 °C resulted in a stable waste product with a significant volume reduction (>50%) and characteristics suitable for long-term storage and/or disposal. However, another important issue to take into account is the complexity of the off-gas generated during the process and the different technical alternatives for its conditioning. Ongoing work addresses the characterization of the ion exchange resin treatment's off-gas. Additionally, the application of plasma technology for the treatment of the off-gas current was studied as an alternative to more conventional processes utilizing oil- or gas-fired post-combustion chambers operating at temperatures in excess of 1000 °C. A laboratory-scale flow reactor, using inductively coupled plasma, operating under sub-atmospheric conditions was developed. Fundamental experiments using model compounds have been performed, demonstrating a high destruction and removal ratio (>99.99%) for different reaction media, at low reactor temperatures and moderate power

  14. Behavior of technetium in nuclear waste vitrification processes.

    Science.gov (United States)

    Pegg, Ian L

    Nearly 100 tests were performed with prototypical melters and off-gas system components to investigate the extents to which technetium is incorporated into the glass melt, partitioned to the off-gas stream, and captured by the off-gas treatment system components during waste vitrification. The tests employed several simulants, spiked with (99m)Tc and Re (a potential surrogate), of the low activity waste separated from nuclear wastes in storage in the Hanford tanks, which is planned for immobilization in borosilicate glass. Single-pass technetium retention averaged about 35 % and increased significantly with recycle of the off-gas treatment fluids. The fraction escaping the recycle loop was very small.

  15. Yeast nuclear RNA processing

    Institute of Scientific and Technical Information of China (English)

    Jade; Bernstein; Eric; A; Toth

    2012-01-01

    Nuclear RNA processing requires dynamic and intricately regulated machinery composed of multiple enzymes and their cofactors.In this review,we summarize recent experiments using Saccharomyces cerevisiae as a model system that have yielded important insights regarding the conversion of pre-RNAs to functional RNAs,and the elimination of aberrant RNAs and unneeded intermediates from the nuclear RNA pool.Much progress has been made recently in describing the 3D structure of many elements of the nuclear degradation machinery and its cofactors.Similarly,the regulatory mechanisms that govern RNA processing are gradually coming into focus.Such advances invariably generate many new questions,which we highlight in this review.

  16. Improvement of melter off-gas design for commercial HALW vitrification facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, A.; Kitamura, M.; Yamanaka, T. [Ishikawajima-Harima Heavy Industries Co., Ltd., Yokohama (Japan); Yoshioka, M.; Endo, N.; Asano, N. [Japan Nuclear Cycle Development Institute, Ibaraki (Japan)

    2001-07-01

    The Japan commercial reprocessing plant is now under construction, and it will commence the operation in 2005. The High Active Liquid Waste (HALW) generated at the plant is treated into glass product at the vitrification facility using the Liquid Fed Joule-Heated Ceramic Melter (LFCM). The characteristic of the LFCM is that the HALW is fed directly onto the molten glass surface with the glass forming material. This process was developed by the Japan Nuclear Cycle Development Institute (JNC). The JNC process was first applied to the Tokai Vitrification Facility (TVF), which is a pilot scale plant having about 1/6 capacity of the commercial facility. The TVF has been in operation since 1995. During the operation, the rapid increase of the differential pressure between the melter plenum and the dust scrubber was observed. This phenomenon is harmful to the long-term continuous operation of TVF. And, it is also anticipated that the same phenomenon will occur in commercial vitrification facility. In order to solve this problem, the countermeasures were studied and developed. Through the study on the deposit growing mechanism, it was probable that the rapid increased differential pressure was attributed to the condensation of meta-boric acid at the outlet of the air-film cooler slits. And, the heating and the humidification of purge air were judged to be effective as the countermeasures to suppress the condensation. On the other hand, the water injection into melter off-gas pipe was found to be very effective to reduce the differential pressure as the results of the various tests. The deposit adhered on the inner surface of the off-gas pipe was almost washed out. And, it was also demonstrated that the system was superior to other systems by virtue of its simplicity and stability. In order to apply the system to the commercial scale plant, the scale-up tests were conducted at JNC mock-up facility using the acrylic model. (author)

  17. Development of silver impregnated alumina for iodine separation from off-gas streams

    Energy Technology Data Exchange (ETDEWEB)

    Funabashi, Kiyomi; Fukasawa, Tetsuo; Kikuchi, Makoto [Energy Research Laboratory, Hitachi (Japan)] [and others

    1995-02-01

    An inorganic iodine adsorbent, silver impregnated alumina (AgA), has been developed to separate iodine effectively from off-gas streams of nuclear facilities and to decrease the volume of waste (spent adsorbent). Iodine removal efficiency was improved at relatively high humidity by using alumina carrier with two different pore diameters. Waste volume reduction was achieved by impregnating relatively large amounts of silver into the alumina pores. The developed adsorbent was tested first with simulated off-gas streams under various experimental conditions and finally with actual off-gas streams of the Karlsruhe reprocessing plant. The decontamination factor (DF) was about 100 with the AgA bed depth of 2cm at 70% relative humidity, which was a DF one order higher than that when AgA with one pore size was used. Iodine adsorption capacity was checked by passing excess iodine into the AgA bed. Values were about 0.12 and 0.35 g-I/cm`-AgA bed for 10 and 24wt% silver impregnated AgA, respectively. The results obtained in this study demonstrated the applicability of the developed AgA to the off-gas treatment system of nuclear facilities.

  18. Biological off-gas treatment: let's make things better

    NARCIS (Netherlands)

    Groenestijn, J.W. van

    1998-01-01

    Biological off-gas treatment is the most effective cleaning method for many off-gases which contain low concentration of pollutants (<5 g/m3). The world market share in off-gas treatment is a few percent. Potential buyers are reserved because of existing biofilter quality differences and lack of exp

  19. Biological off-gas treatment: let's make things better

    NARCIS (Netherlands)

    Groenestijn, J.W. van

    1998-01-01

    Biological off-gas treatment is the most effective cleaning method for many off-gases which contain low concentration of pollutants (<5 g/m3). The world market share in off-gas treatment is a few percent. Potential buyers are reserved because of existing biofilter quality differences and lack of

  20. Radon depletion in xenon boil-off gas

    Energy Technology Data Exchange (ETDEWEB)

    Bruenner, S.; Cichon, D.; Lindemann, S.; Undagoitia, T.M.; Simgen, H. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    2017-03-15

    An important background in detectors using liquid xenon for rare event searches arises from the decays of radon and its daughters. We report for the first time a reduction of {sup 222}Rn in the gas phase above a liquid xenon reservoir. We show a reduction factor of >or similar 4 for the {sup 222}Rn concentration in boil-off xenon gas compared to the radon enriched liquid phase. A semiconductor-based α-detector and miniaturized proportional counters are used to detect the radon. As the radon depletion in the boil-off gas is understood as a single-stage distillation process, this result establishes the suitability of cryogenic distillation to separate radon from xenon down to the 10{sup -15} mol/mol level. (orig.)

  1. Radon depletion in xenon boil-off gas

    CERN Document Server

    Bruenner, S; Lindemann, S; Undagoitia, T Marrodán; Simgen, H

    2016-01-01

    An important background in detectors using liquid xenon for rare event searches arises from the decays of radon and its daughters. We report for the first time a reduction of $^{222}$Rn in the gas phase above a liquid xenon reservoir. We show a reduction factor of $\\gtrsim 4$ for the $^{222}$Rn concentration in boil-off xenon gas compared to the radon enriched liquid phase. A semiconductor-based $\\alpha$-detector and miniaturized proportional counters are used to detect the radon. As the radon depletion in the boil-off gas is understood as a single-stage distillation process, this result establishes the suitability of cryogenic distillation to separate radon from xenon down to the $10^{-15}\\,$mol/mol level.

  2. Radon depletion in xenon boil-off gas

    Science.gov (United States)

    Bruenner, S.; Cichon, D.; Lindemann, S.; Undagoitia, T. Marrodán; Simgen, H.

    2017-03-01

    An important background in detectors using liquid xenon for rare event searches arises from the decays of radon and its daughters. We report for the first time a reduction of ^{222}Rn in the gas phase above a liquid xenon reservoir. We show a reduction factor of ≳ 4 for the ^{222}Rn concentration in boil-off xenon gas compared to the radon enriched liquid phase. A semiconductor-based α -detector and miniaturized proportional counters are used to detect the radon. As the radon depletion in the boil-off gas is understood as a single-stage distillation process, this result establishes the suitability of cryogenic distillation to separate radon from xenon down to the 10^{-15} mol/mol level.

  3. Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

    Energy Technology Data Exchange (ETDEWEB)

    R. T. Jubin; D. M. Strachan; N. R. Soelberg

    2013-09-01

    Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. This report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.

  4. Iodine Adsorption by Ag-Aerogel under Prototypical Vessel Off-Gas Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    U.S. regulations will require the removal of 129I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. The required plant decontamination factor for iodine will vary based on fuel burnup, cooling time, and other factors but is very likely to be >1000 and could be as high as 8000. Multiple off-gas streams within a UNF reprocessing plant combine prior to environmental release, and each of these streams contains some amount of iodine. To achieve the decontamination factors (DFs) that are likely to be required by regulations, iodine removal from the vessel off-gas will be necessary. The vessel off-gas contains iodine at very dilute concentrations (ppb levels) and will also contain water vapor. Iodine species present are likely to include both elemental and organic iodides. There will also be solvent vapors and volatile radiolysis products. The United States has considered the use of silver-based sorbents for removal of iodine from UNF off-gas streams, but little is known about the behavior of those sorbents at very dilute iodine concentrations. The purpose of this study was to expose silver-functionalized silica aerogel (AgAerogel) to a prototypical vessel off-gas stream containing 40 ppb methyl iodide to obtain information about organic iodine capture by silver-sorbents at very low iodine concentrations. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the overall system DF could be obtained. Results show that CH3I penetrates into a AgAerogel sorbent bed to a depth of 3.9 cm under prototypical vessel off-gas conditions. An iodine loading of 22 mg I/g AgAerogel was observed in the first 0.3 cm of the bed. Of the iodine delivered to the system, 48% could not be accounted for, and future efforts will investigate this concern. Direct calculation of the decontamination factor is not

  5. Iodine Adsorption by Ag-Aerogel under Prototypical Vessel Off-Gas Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    U.S. regulations will require the removal of 129I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. The required plant decontamination factor for iodine will vary based on fuel burnup, cooling time, and other factors but is very likely to be >1000 and could be as high as 8000. Multiple off-gas streams within a UNF reprocessing plant combine prior to environmental release, and each of these streams contains some amount of iodine. To achieve the decontamination factors (DFs) that are likely to be required by regulations, iodine removal from the vessel off-gas will be necessary. The vessel off-gas contains iodine at very dilute concentrations (ppb levels), and will also contain water vapor. Iodine species present are likely to include both elemental and organic iodides. There will also be solvent vapors and volatile radiolysis products. The United States has considered the use of silver-based sorbents for removal of iodine from UNF off-gas streams, but little is known about the behavior of those sorbents at very dilute iodine concentrations. The purpose of this study was to expose silver-functionalized silica aerogel (AgAerogel) to a prototypical vessel off-gas stream containing 40 ppb methyl iodide to obtain information about organic iodine capture by silver-sorbents at very low iodine concentrations. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the overall system DF could be obtained. Results show that CH3I penetrates into a AgAerogel sorbent bed to a depth of 3.9 cm under prototypical vessel off-gas conditions. An iodine loading of 22 mg I/g AgAerogel was observed in the first 0.3 cm of the bed. Of the iodine delivered to the system, 48% could not be accounted for, and future efforts will investigate this concern. Direct calculation of the decontamination factor is not

  6. DWPF Melter Off-Gas Flammability Assessment for Sludge Batch 9

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-07-11

    The slurry feed to the Defense Waste Processing Facility (DWPF) melter contains several organic carbon species that decompose in the cold cap and produce flammable gases that could accumulate in the off-gas system and create potential flammability hazard. To mitigate such a hazard, DWPF has implemented a strategy to impose the Technical Safety Requirement (TSR) limits on all key operating variables affecting off-gas flammability and operate the melter within those limits using both hardwired/software interlocks and administrative controls. The operating variables that are currently being controlled include; (1) total organic carbon (TOC), (2) air purges for combustion and dilution, (3) melter vapor space temperature, and (4) feed rate. The safety basis limits for these operating variables are determined using two computer models, 4-stage cold cap and Melter Off-Gas (MOG) dynamics models, under the baseline upset scenario - a surge in off-gas flow due to the inherent cold cap instabilities in the slurry-fed melter.

  7. Off-Gas Adsorption Model Capabilities and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Lyon, Kevin L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Welty, Amy K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Law, Jack [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ladshaw, Austin [Georgia Inst. of Technology, Atlanta, GA (United States); Yiacoumi, Sotira [Georgia Inst. of Technology, Atlanta, GA (United States); Tsouris, Costas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-03-01

    Off-gas treatment is required to reduce emissions from aqueous fuel reprocessing. Evaluating the products of innovative gas adsorption research requires increased computational simulation capability to more effectively transition from fundamental research to operational design. Early modeling efforts produced the Off-Gas SeParation and REcoverY (OSPREY) model that, while efficient in terms of computation time, was of limited value for complex systems. However, the computational and programming lessons learned in development of the initial model were used to develop Discontinuous Galerkin OSPREY (DGOSPREY), a more effective model. Initial comparisons between OSPREY and DGOSPREY show that, while OSPREY does reasonably well to capture the initial breakthrough time, it displays far too much numerical dispersion to accurately capture the real shape of the breakthrough curves. DGOSPREY is a much better tool as it utilizes a more stable set of numerical methods. In addition, DGOSPREY has shown the capability to capture complex, multispecies adsorption behavior, while OSPREY currently only works for a single adsorbing species. This capability makes DGOSPREY ultimately a more practical tool for real world simulations involving many different gas species. While DGOSPREY has initially performed very well, there is still need for improvement. The current state of DGOSPREY does not include any micro-scale adsorption kinetics and therefore assumes instantaneous adsorption. This is a major source of error in predicting water vapor breakthrough because the kinetics of that adsorption mechanism is particularly slow. However, this deficiency can be remedied by building kinetic kernels into DGOSPREY. Another source of error in DGOSPREY stems from data gaps in single species, such as Kr and Xe, isotherms. Since isotherm data for each gas is currently available at a single temperature, the model is unable to predict adsorption at temperatures outside of the set of data currently

  8. Modeling nuclear processes by Simulink

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Nahrul Khair Alang Md, E-mail: nahrul@iium.edu.my [Faculty of Engineering, International Islamic University Malaysia, Jalan Gombak, Selangor (Malaysia)

    2015-04-29

    Modelling and simulation are essential parts in the study of dynamic systems behaviours. In nuclear engineering, modelling and simulation are important to assess the expected results of an experiment before the actual experiment is conducted or in the design of nuclear facilities. In education, modelling can give insight into the dynamic of systems and processes. Most nuclear processes can be described by ordinary or partial differential equations. Efforts expended to solve the equations using analytical or numerical solutions consume time and distract attention from the objectives of modelling itself. This paper presents the use of Simulink, a MATLAB toolbox software that is widely used in control engineering, as a modelling platform for the study of nuclear processes including nuclear reactor behaviours. Starting from the describing equations, Simulink models for heat transfer, radionuclide decay process, delayed neutrons effect, reactor point kinetic equations with delayed neutron groups, and the effect of temperature feedback are used as examples.

  9. Modeling nuclear processes by Simulink

    Science.gov (United States)

    Rashid, Nahrul Khair Alang Md

    2015-04-01

    Modelling and simulation are essential parts in the study of dynamic systems behaviours. In nuclear engineering, modelling and simulation are important to assess the expected results of an experiment before the actual experiment is conducted or in the design of nuclear facilities. In education, modelling can give insight into the dynamic of systems and processes. Most nuclear processes can be described by ordinary or partial differential equations. Efforts expended to solve the equations using analytical or numerical solutions consume time and distract attention from the objectives of modelling itself. This paper presents the use of Simulink, a MATLAB toolbox software that is widely used in control engineering, as a modelling platform for the study of nuclear processes including nuclear reactor behaviours. Starting from the describing equations, Simulink models for heat transfer, radionuclide decay process, delayed neutrons effect, reactor point kinetic equations with delayed neutron groups, and the effect of temperature feedback are used as examples.

  10. Dissolver Off-gas Hot Operations Authorization (AFCI CETE Milestone Report)

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, Robert Thomas [ORNL

    2009-06-01

    The head-end processing of the Coupled-End-to-End (CETE) Demonstration includes fuel receipt, fuel disassembly, exposure of fuel (e.g., by segmenting the fuel pins), voloxidation of the fuel to separate tritium, and fuel dissolution. All of these processing steps with the exception of the dissolution step will be accomplished in the Irradiated Fuels Examination Laboratory (IFEL) (Building 3525). The final headend step will be performed in the Radiochemical Engineering Development Center (Building 7920). The primary purpose of the fuel dissolution step is to prepare the solid fuel for subsequent liquid separations steps. This is accomplished by dissolving the fuel solids using nitric acid. During the dissolution process gases are evolved. Oxides of nitrogen are the primary off-gas components generated by the reactions of nitric acid and the fuel oxides however, during the dissolution and sparging of the resulting solution, iodine, C-14 as carbon dioxide, xenon, and krypton gasses are also released to the off-gas stream. The Dissolver Off-gas treatment rack provides a means of trapping these volatile fission products and other gases via various trapping media. Specifically the rack will recover iodine on a solid sorbent bed, scrub NOx in a water/acid column, scrub CO{sub 2} in a caustic scrubber column, remove moisture with solid sorbent drier beds and recover Xe and Kr using solid absorbent beds. The primary purpose of this experimental rack and the off-gas rack associated with the voloxidation equipment located at IFEL is to close the material balances around the volatile gases and to provide an understanding of the impacts of specific processing conditions on the fractions of the volatile components released from the various head-end processing steps.

  11. ART CCIM Phase II-A Off-Gas System Evaluation Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Jay Roach

    2009-01-01

    This test plan defines testing to be performed using the Idaho National Laboratory (INL) engineering-scale cold crucible induction melter (CCIM) test system for Phase II-A of the Advanced Remediation Technologies (ART) CCIM Project. The multi-phase ART-CCIM Project is developing a conceptual design for replacing the joule-heated melter (JHM) used to treat high level waste (HLW) in the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) with a cold crucible induction melter. The INL CCIM test system includes all feed, melter off-gas control, and process control subsystems needed for fully integrated operation and testing. Testing will include operation of the melter system while feeding a non-radioactive slurry mixture prepared to simulate the same type of waste feed presently being processed in the DWPF. Process monitoring and sample collection and analysis will be used to characterize the off-gas composition and properties, and to show the fate of feed constituents, to provide data that shows how the CCIM retrofit conceptual design can operate with the existing DWPF off-gas control system.

  12. McClellan AFB SVE Off-Gas Characterization, Literature Review, and Technology Selection

    Science.gov (United States)

    2007-11-02

    McClellan Air Force Base ( AFB ) in Sacramento , California, is part of the National Environmental Technology Test Site (NETTS) program. NETTS is a... McClellan AFB uses soil vapor extraction (SVE) systems to remove contamination from soils. The SVE systems draw air through the pore spaces between...of the NETTS program, with respect to McClellan AFB , is to develop a treatment process to remove the VOCs from the off-gas before it is discharged into

  13. Degradation of off-gas toluene in continuous pyrite Fenton system.

    Science.gov (United States)

    Choi, Kyunghoon; Bae, Sungjun; Lee, Woojin

    2014-09-15

    Degradation of off-gas toluene from a toluene reservoir and a soil vapor extraction (SVE) process was investigated in a continuous pyrite Fenton system. The removal of off-gas toluene from the toluene reservoir was >95% by 8h in the pyrite Fenton system, while it was ∼97 % by 3h in classic Fenton system and then rapidly decreased to initial level by 8h. Continuous consumption of low Fe(II) concentration dissolved from pyrite surface (0.05-0.11 mM) was observed in the pyrite Fenton system, which can lead to the effective and successful removal of the gas-phase toluene due to stable production of OH radical (OH). Inhibitor and spectroscopic test results showed that OH was a dominant radical that degraded gas-phase toluene during the reaction. Off-gas toluene from the SVE process was removed by 96% in the pyrite Fenton system, and remnant toluene from rebounding effect was treated by 99%. Main transformation products from toluene oxidation were benzoic acid (31.4%) and CO2 (38.8%) at 4h, while traces of benzyl alcohol (1.3%) and benzaldehyde (0.7%) were observed. Maximum operation time of continuous pyrite Fenton system was estimated to be 56-61 d and its optimal operation time achieving emission standard was 28.9 d. Copyright © 2014 Elsevier B.V. All rights reserved.

  14. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  15. Validation of DWPF Melter Off-Gas Combustion Model

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.S.

    2000-08-23

    The empirical melter off-gas combustion model currently used in the DWPF safety basis calculations is valid at melter vapor space temperatures above 570 degrees C, as measured in the thermowell. This lower temperature bound coincides with that of the off-gas data used as the basis of the model. In this study, the applicability of the empirical model in a wider temperature range was assessed using the off-gas data collected during two small-scale research melter runs. The first data set came from the Small Cylindrical Melter-2 run in 1985 with the sludge feed coupled with the precipitate hydrolysis product. The second data set came from the 774-A melter run in 1996 with the sludge-only feed prepared with the modified acid addition strategy during the feed pretreatment step. The results of the assessment showed that the data from these two melter runs agreed well with the existing model, and further provided the basis for extending the lower temperature bound of the model to the measured melter vapor space temperature of 445 degrees C.

  16. A Literature Survey to Identify Potentially Volatile Iodine-Bearing Species Present in Off-Gas Streams

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, S. H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Strata-G, Knoxville, TN (United States); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-30

    Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3H, 14C, 85Kr, and 129I. Of these, 129I has the longest half-life and potentially high biological impact. Accordingly, control of the release of 129I is most critical with respect to the regulations for the release of radioactive material in stack emissions. It is estimated that current EPA regulations (EPA 2010) would require any reprocessing plant in the United States to limit 129I release to less than 0.05 Ci/MTIHM for a typical fuel burnup of 55 gigawatt days per metric tonne (GWd/t) (Jubin 2011). The study of inorganic iodide in off-gas systems has been almost exclusively limited to I2 and the focus of organic iodide studies has been CH3I. In this document, we provide the results of an examination of publically available literature that is relevant to the presence and sources of both inorganic and organic iodine-bearing species in reprocessing plants. We especially focus on those that have the potential to be poorly sequestered with traditional capture methodologies. Based on the results of the literature survey and some limited thermodynamic modeling, the inorganic iodine species hypoiodous acid (HOI) and iodine monochloride (ICl) were identified as potentially low-sorbing iodine species that could present in off-gas systems. Organic species of interest included both short chain alkyl iodides such as methyl iodide (CH3I) and longer alkyl iodides up to iodododecane (C10H21I). It was found that fuel dissolution may provide conditions conducive to HOI formation and has been shown to result in volatile long-chain alkyl iodides, though these may not volatilize until later in the reprocessing sequence. Solvent extraction processes were found to be significant sources of various organic iodine-bearing species; formation of these

  17. Waste decomposition and off-gas emissions of dry, single-stage anaerobic digestion processes followed by a composting stage: The example of the KOMPOGAS process; Abbauleistung und Abluftemissionen bei trockenen, einstufigen Vergaerungsverfahren mit nachgeschalteter Rotte am Beispiel des KOMPOGAS-Verfahrens

    Energy Technology Data Exchange (ETDEWEB)

    Fricke, K. [Technische Univ. Braunschweig (Germany). Leichtweiss-Institut fuer Wasserbau; Leisner, R. [KOGAS AG, Uzwil (Switzerland); Wallmann, R. [Ingenieurgemeinschaft Witzenhausen, Fricke und Turk GmbH, Witzenhausen (Germany)

    2001-07-01

    The working group IGW/LWI (Ingenieurgemeinschaft Witzenhausen Fricke and Turk GmbH and Leichtweiss Institute of TU Brunswick University) carried out scientific and technical investigations of anaerobic digestion and subsequent composting at the mechanical-biological waste treatment plant at Kufstein/Tyrol on behalf of KOGAS GmbH. The investigations started in May 2000; they were to provide information for analysis and optimisation of the process and products with a view to the minimisation of residues for dumping or incineration. The following subjects were investigated: Mass balance; decomposition rate; gas yields and gas quality; quality of residues for dumping (biological stability and calorific value); off-gas emissions; process control. Performance data of the plant will provide a basis for further planning and will be interpreted against the background of current legal boundary conditions. [German] Die Arbeitsgemeinschaft IGW/LWI (Ingenieurgemeinschaft Witzenhausen Fricke and Turk GmbH und Leichtweiss-Institut der TU Braunschweig) wurde im Mai 2000 von der KOGAS GmbH beauftragt, die wissenschaftlich-technische Begleitung eines Versuches zur Vergaerung und Nachrotte von Restmuell auf dem Gelaende der MBA Kufstein/Tirol durchzufuehren. Ziel der Untersuchung war die Ueberpruefung und Optimierung des Vergaerungs- und Nachrotteprozesses sowie der Materialaufbereitung und Konfektionierung im Hinblick auf die Minimierung von Stofffluessen in die energetische Verwertung bzw. thermische Behandlung. Im Rahmen des Vorhabens wurden folgende Thermokomplexe bearbeitet: - Massenbilanz, - Abbauleistung, - Gasertrag und Gasqualitaet, - Qualitaet der zu deponierenden Abfaelle (biologische Stabilitaet und Heizwert), - Abluftemissionen, - Prozesssteuerung. Leistungsdaten der Verfahren sollten als Planungsgrundlagen bereitgestellt und vor dem Hintergrund der aktuellen rechtlichen Rahmenbedingungen bewertet werden. (orig.)

  18. Nuclear sensor signal processing circuit

    Science.gov (United States)

    Kallenbach, Gene A.; Noda, Frank T.; Mitchell, Dean J.; Etzkin, Joshua L.

    2007-02-20

    An apparatus and method are disclosed for a compact and temperature-insensitive nuclear sensor that can be calibrated with a non-hazardous radioactive sample. The nuclear sensor includes a gamma ray sensor that generates tail pulses from radioactive samples. An analog conditioning circuit conditions the tail-pulse signals from the gamma ray sensor, and a tail-pulse simulator circuit generates a plurality of simulated tail-pulse signals. A computer system processes the tail pulses from the gamma ray sensor and the simulated tail pulses from the tail-pulse simulator circuit. The nuclear sensor is calibrated under the control of the computer. The offset is adjusted using the simulated tail pulses. Since the offset is set to zero or near zero, the sensor gain can be adjusted with a non-hazardous radioactive source such as, for example, naturally occurring radiation and potassium chloride.

  19. Nuclear energy and process heating

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-10-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating and several industrial applications. Although only About 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven, a

  20. Nuclear energy and process heating

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating, and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating, (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating, and several industrial applications. Although only about 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven

  1. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  2. Helium extraction and nitrogen removal from LNG boil-off gas

    Science.gov (United States)

    Xiong, L.; Peng, N.; Liu, L.; Gong, L.

    2017-02-01

    The helium bearing boil off gas (BOG) from liquid natural gas (LNG) storage tank in LNG plant, which has a helium concentration of about 1%, has attracted the attention in China as a new helium source. As the BOG is usually reused by re-condensing to recover methane, it is likely to cause continuous accumulation of nitrogen in the unit, thus a nitrogen removal process must be integrated. This paper describes a conceptional cryogenic separation system aiming at recovering methane, helium and nitrogen from BOG based on cryogenic distillation and condensation process.

  3. Off-gas characteristics of defense waste vitrification using liquid-fed Joule-heated ceramic melters

    Energy Technology Data Exchange (ETDEWEB)

    Goles, R.W.; Sevigny, G.J.

    1983-09-01

    Off-gas and effluent characterization studies have been established as part of a PNL Liquid-Fed Ceramic Melter development program supporting the Savannah River Laboratory Defense Waste Processing Facility (SRL-DWPF). The objectives of these studies were to characterize the gaseous and airborne emission properties of liquid-fed joule-heated melters as a function of melter operational parameters and feed composition. All areas of off-gas interest and concern including effluent characterization, emission control, flow rate behavior and corrosion effects have been studied using alkaline and formic-acid based feed compositions. In addition, the behavioral patterns of gaseous emissions, the characteristics of melter-generated aerosols and the nature and magnitude of melter effluent losses have been established under a variety of feeding conditions with and without the use of auxiliary plenum heaters. The results of these studies have shown that particulate emissions are responsible for most radiologically important melter effluent losses. Melter-generated gases have been found to be potentially flammable as well as corrosive. Hydrogen and carbon monoxide present the greatest flammability hazard of the combustibles produced. Melter emissions of acidic volatile compounds of sulfur and the halogens have been responsible for extensive corrosion observed in melter plenums and in associated off-gas lines and processing equipment. The use of auxiliary plenum heating has had little effect upon melter off-gas characteristics other than reducing the concentrations of combustibles.

  4. FY-12 INL KR CAPTURE ACTIVITIES SUPPORTING THE OFF-GAS SIGMA TEAM

    Energy Technology Data Exchange (ETDEWEB)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D Law

    2012-08-01

    Tasks performed this year by INL Kr capture off-gas team members can be segregated into three separate task sub-sections which include: 1) The development and testing of a new engineered form sorbent, 2) An initial NDA gamma scan effort performed on the drum containing the Legacy Kr-85 sample materials, and 3) Collaborative research efforts with PNNL involving the testing of the Ni-DOBDC MOF and an initial attempt to make powdered chalcogel material into an engineered form using our binding process. This document describes the routes to success for the three task sub-sections.

  5. CHARACTERIZATION OF DWPF MELTER OFF-GAS QUENCHER AND STEAM ATOMIZED SCRUBBER DEPOSIT SAMPLES

    Energy Technology Data Exchange (ETDEWEB)

    Zeigler, K; Ned Bibler, N

    2007-06-06

    This report summarizes the results from the characterization of deposits from the inlets of the primary off-gas Quencher and Steam Atomized Scrubber (SAS) in the Defense Waste Processing Facility (DWPF), as requested by a technical assistance request. DWPF requested elemental analysis and compound identification to help determine the potential causes for the substance formation. This information will be fed into Savannah River National Laboratory modeling programs to determine if there is a way to decrease the formation of the deposits. The general approach to the characterization of these samples included x-ray diffraction (XRD), scanning electron microscopy (SEM), and chemical analysis. The following conclusions are drawn from the analytical results found in this report: (1) The deposits are not high level waste glass from the DWPF melt pool based on comparison of the compositions of deposits to the composition of a sample of glass taken from the pour stream of the melter during processing of Sludge Batch 3. (2) Chemical composition results suggest that the deposits are probably a combination of sludge and frit particles entrained in the off-gas. (3) Gamma emitters, such as Co-60, Cs-137, Eu-154, Am-241, and Am-243 were detected in both the Quencher and SAS samples with Cs-137 having the highest concentration of the gamma emitters. (4) No evidence existed for accumulation of fissile material (U-233, U-235, and Pu-239) relative to Fe in either deposit. (5) XRD results indicated both samples were primarily amorphorous and contained some crystals of the iron oxides, hematite and magnetite (Fe{sub 2}O{sub 3} and Fe(Fe{sub 2}O{sub 4})), along with sodium nitrate (NaNO{sub 3}). The other main crystalline compound in the SAS deposit was mercurous chloride. The main crystalline compound in the Quencher deposit was a uranium oxide compound. These are all sludge components. (6) SEM analysis of the Quencher deposit revealed crystalline uranium compounds within the sample

  6. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  7. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  8. Design report: An off gas trapping system for a voloxidizer in INL of US

    Energy Technology Data Exchange (ETDEWEB)

    Jung, I. H.; Shin, J. M.; Park, J. J.; Park, G. I.; Lee, H. H

    2006-09-15

    This reports on the 'Development of Voloxidation Process for Treatment of LWR Spent Fuel', and it is the second year since it has started from June 2004 as a tripartite cooperation project among KAERI(Korea Atomic Energy Research Institute), INL(Idaho National Laboratory) and ORNL(Oak Ridge National Laboratory). This report is described mainly for the Task B2 accomplished during the second project year. The Task B2 in proposal contains two sub-tasks. The first one is design of an off-gas treatment system for a voloxidizer to be used in HFEF of INL. For this, KAERI team developed the design of INL OTS (Off-gas Treatment System) for hot experiment in the HFEF. INL team modified and completed the design of the INL OTS. The second task is manufacturing and test operation of the INL OTS for a voloxidizer in the INL. Manufacturing of the OTS is accomplished by INL team with co-work of KAERI. KAERI provided four sets of trapping filters needed for conducting hot experiment in the INL HFEF.

  9. Wastewater off-gas to electricity: A look at energy recovery using the internal combustion engine

    Energy Technology Data Exchange (ETDEWEB)

    Hawkins, S.C. [Eckenfelder Inc., Greenville, SC (United States)

    1997-12-31

    Historically, publicly owned treatment works (POTWs) have not been regulated for air pollutant emissions. Federal, state, and local regulatory agencies have placed increased scrutiny on POTW volatile organic compound (VOC) and hazardous air pollutant (HAP) emissions. In fact, the USEPA is developing Maximum Achievable Control Technology (MACT) standards for publicly owned facilities that receive and treat sewage and wastewater from residential, commercial, and industrial generators. Once promulgated, this standard will require POTWs to limit and control HAP emissions. One option available for the control of HAP emissions from wastewater facilities is the use of the internal combustion (IC) engine-generator systems to convert wastewater off-gas to energy. Single or multiple IC engines can be arranged to utilize wastewater off-gas in place of natural gas and diesel fuel to produce electricity. The electricity generated can then be used to power the treatment facility or sold to the local utility. These systems utilize an otherwise wasted energy resource while significantly reducing HAP and VOC emissions from the wastewater treatment process. This paper will discuss the principles of using the internal combustion engine for energy recovery and examine the technical and economic benefits involved with their use at POTW facilities.

  10. Boil-off gas vapors are recovered by reliquefaction in LNG

    Energy Technology Data Exchange (ETDEWEB)

    Levay, M.; Petit, P.; Paradowski, H.

    1986-02-24

    Although great care is taken to prevent heat leaks into cryogenic equipment in LNG terminals, boil-off vapors evolve from LNG stored at thermodynamic equilibrium. The quantities of boil-off vapors may be quite considerable. They account for about 1% of the total gas quantity received and sent out at the monitor-de-bretagne LNG terminal of Gaz de France. A novel process has significantly cut boil-off vapor handling costs. It is free of technical problems which would arise from local utilization of the gas and makes boil-off recovery possible under optimum conditions. In addition, the process shows an excellent degree of reliability. Boil-off vapors have a lower heating value than the stored LNG. However, since they mainly consist of methane, their economic usefulness makes vapor recovery necessary. This boil-off gas, with widely fluctuating quantities and qualities, cannot be readily used locally. The vapors must be sent out into the grid.

  11. Literature search for offsite data to improve the DWPF melter off-gas model

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W.E.

    2000-05-04

    This report documents the literature search performed and any relevant data that may help relax some of the constraints on the DWPF melter off-gas model. The objective of this task was to look for outside sources of technical data to help reduce some of the conservatism built in the DWPF melter off-gas model.

  12. Processing multidimensional nuclear physics data

    Energy Technology Data Exchange (ETDEWEB)

    Becker, J. [Lawrence Livermore National Lab., CA (United States)

    1994-11-15

    Modern Ge detector arrays for gamma-ray spectroscopy are producing data sets unprecedented in size and event multiplicity. Gammasphere, the DOE sponsored array, has the following characteristics: (1) High granularity (110 detectors); (2) High efficiency (10%); and (3) Precision energy measurements (Delta EE = 0.2%). Characteristics of detector line shape, the data set, and the standard practice in the nuclear physics community to the nuclear gamma-ray cascades from the 4096 times 4096 times 4096 data cube will be discussed.

  13. Formation of the ZnFe2O4 phase in an electric arc furnace off-gas treatment system.

    Science.gov (United States)

    Suetens, T; Guo, M; Van Acker, K; Blanpain, B

    2015-04-28

    To better understand the phenomena of ZnFe2O4 spinel formation in electric arc furnace dust, the dust was characterized with particle size analysis, X-ray fluorescence (XRF), electron backscatter diffraction (EBSD), and electron probe micro-analysis (EPMA). Different ZnFe2O4 formation reaction extents were observed for iron oxide particles with different particle sizes. ZnO particles were present as both individual particles and aggregated on the surface of larger particles. Also, the slag particles found in the off-gas were shown not to react with the zinc vapor. After confirming the presence of a ZnFe2O4 formation reaction, the thermodynamic feasibility of in-process separation - a new electric arc furnace dust treatment technology - was reevaluated. The large air intake and the presence of iron oxide particles in the off-gas were included into the thermodynamic calculations. The formation of the stable ZnFe2O4 spinel phase was shown to be thermodynamically favorable in current electric arc furnace off-gas ducts conditions even before reaching the post combustion chamber.

  14. MELTER OFF-GAS FLAMMABILITY ASSESSMENT FOR DWPF ALTERNATE REDUCTANT FLOWSHEET OPTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.

    2011-07-08

    Glycolic acid and sugar are being considered as potential candidates to substitute for much of the formic acid currently being added to the Defense Waste Processing Facility (DWPF) melter feed as a reductant. A series of small-scale melter tests were conducted at the Vitreous State Laboratory (VSL) in January 2011 to collect necessary data for the assessment of the impact of these alternate reductants on the melter off-gas flammability. The DM10 melter with a 0.021 m{sup 2} melt surface area was run with three different feeds which were prepared at SRNL based on; (1) the baseline formic/nitric acid flowsheet, (2) glycolic/formic/nitric acid flowsheet, and (3) sugar/formic/nitric acid flowsheet - these feeds will be called the baseline, glycolic, and sugar flowsheet feeds, respectively, hereafter. The actual addition of sugar to the sugar flowsheet feed was made at VSL before it was fed to the melter. For each feed, the DM10 was run under both bubbled (with argon) and non-bubbled conditions at varying melter vapor space temperatures. The goal was to lower its vapor space temperature from nominal 500 C to less than 300 C at 50 C increments and maintain steady state at each temperature at least for one hour, preferentially for two hours, while collecting off-gas data including CO, CO{sub 2}, and H{sub 2} concentrations. Just a few hours into the first test with the baseline feed, it was discovered that the DM10 vapor space temperature would not readily fall below 350 C simply by ramping up the feed rate as the test plan called for. To overcome this, ambient air was introduced directly into the vapor space through a dilution air damper in addition to the natural air inleakage occurring at the operating melter pressure of -1 inch H{sub 2}O. A detailed description of the DM10 run along with all the data taken is given in the report issued by VSL. The SRNL personnel have analyzed the DM10 data and identified 25 steady state periods lasting from 32 to 92 minutes for all

  15. ART CCIM PHASE II-A OFF-GAS SYSTEM EVALUATION TEST REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg

    2009-04-01

    AREVA Federal Services (AFS) is performing a multi-year, multi-phase Advanced Remediation Technologies (ART) project, sponsored by the U.S. Department of Energy (DOE), to evaluate the feasibility and benefits of replacing the existing joule-heated melter (JHM) used to treat high level waste (HLW) in the Defense Waste Processing Facility (DWPF) at the Savannah River Site with a cold crucible induction melter (CCIM). The AFS ART CCIM project includes several collaborators from AREVA subsidiaries, French companies, and DOE national laboratories. The Savannah River National Laboratory and the Commissariat a l’Energie Atomique (CEA) have performed laboratory-scale studies and testing to determine a suitable, high-waste-loading glass matrix. The Idaho National Laboratory (INL) and CEA are performing CCIM demonstrations at two different pilot scales to assess CCIM design and operation for treating SRS sludge wastes that are currently being treated in the DWPF. SGN is performing engineering studies to validate the feasibility of retrofitting CCIM technology into the DWPF Melter Cell. The long-term project plan includes more lab-testing, pilot- and large-scale demonstrations, and engineering activities to be performed during subsequent project phases. A simulant of the DWPF SB4 feed was successfully fed and melted in a small pilot-scale CCIM system during two test series. The OGSE tests provide initial results that (a) provide melter operating conditions while feeding a DWPF SB4 simulant feed, (b) determine the fate of feed organic and metal feed constituents and metals partitioning, and (c) characterize the melter off-gas source term to a downstream off-gas system. The INL CCIM test system was operated continuously for about 30 hours during the parametric test series, and for about 58 hours during the OGSE test. As the DWPF simulant feed was continuously fed to the melter, the glass level gradually increased until a portion of the molten glass was drained from the melter

  16. Application of off-gas treatment technology to soil vapour extraction systems

    Energy Technology Data Exchange (ETDEWEB)

    Karp, G. S.; Harvey, E. M.; McKee, R. C. E. [O`Connor Associates Environmental Inc., Oakville, ON (Canada); Lucas, W. P. [Commenco Systems Inc., Concord, ON (Canada)

    1995-12-31

    Various off-gas treatment technologies, including carbon adsorption, thermal incineration, UV oxidation, bio-reactors, combustion and catalytic oxidation were investigated as means to remediate sub-surface soils contaminated with petroleum hydrocarbons or volatile organic compounds. The primary objective was to determine the most cost-effective portable off-gas treatment technology for a typical soil vapour extraction system. Advantages, disadvantages and relative costs of each technology were summarized. Catalytic oxidation was found to be the most cost-effective method for off-gas treatment for the specified soil vapour extraction systems.

  17. Airborne waste management technology applicable for use in reprocessing plants for control of iodine and other off-gas constituents

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R.T.

    1988-02-01

    Extensive work in the area of iodine removal from reprocessing plant off-gas streams using various types of solid sorbent materials has been conducted worldwide over the past two decades. This work has focused on the use of carbon filters, primarily for power plant applications. More recently, the use of silver-containing sorbents has been the subject of considerable research. The most recent work in the United States has addressed the use of silver-exchanged faujasites and mordenites. The chemical reactions of iodine with silver on the sorbent are not well defined, but it is generally believed that chemisorbed iodides and iodates are formed. The process for iodine recovery generally involves passage of the iodine-laden gas stream through a packed bed of the adsorbent material preheated to a temperature of about 150/degree/C. Most iodine removal system designs utilizing silver-containing solid sorbents assume only a 30 to 50% silver utilization. Based on laboratory tests, potentially 60 to 70% of the silver contained in the sorbents can be reacted with iodine. To overcome the high cost of silver associated with these materials, various approaches have been explored. Among these are the regeneration of the silver-containing sorbent by stripping the iodine and trapping the iodine on a sorbent that has undergone only partial silver exchange and is capable of attaining a much higher silver utilization. This summary report describes the US work in regeneration of iodine-loaded solid sorbent material. In addition, the report discusses the broader subject of plant off-gas treatment including system design. The off-gas technologies to recovery No/sub x/ and to recover and dispose of Kr, /sup 14/C, and I are described as to their impacts on the design of an integrated off-gas system. The effect of ventilation philosophy for the reprocessing plant is discussed as an integral part of the overall treatment philosophy of the plant off-gas. 103 refs., 5 figs., 8 tabs.

  18. Nuclear Processes at Solar Energy

    CERN Document Server

    Broggini, C

    2003-01-01

    LUNA, Laboratory for Underground Nuclear Astrophysics at Gran Sasso, is measuring fusion cross sections down to the energy of the nucleosynthesis inside stars. Outstanding results obtained up to now are the cross-section measurements within the Gamow peak of the Sun of $^{3}He(^{3}He,2p)^{4}He$ and the $D(p,\\gamma)^{3}He$. The former plays a big role in the proton-proton chain, largely affecting the calculated solar neutrino luminosity, whereas the latter is the reaction that rules the proto-star life during the pre-main sequence phase. The implications of such measurements will be discussed. Preliminary results obtained last year on the study of $^{14}N(p,\\gamma)^{15}O$, the slowest reaction of the CNO cycle, will also be shown.

  19. Robot development for nuclear material processing

    Energy Technology Data Exchange (ETDEWEB)

    Pedrotti, L.R.; Armantrout, G.A.; Allen, D.C.; Sievers, R.H. Sr.

    1991-07-01

    The Department of Energy is seeking to modernize its special nuclear material (SNM) production facilities and concurrently reduce radiation exposures and process and incidental radioactive waste generated. As part of this program, Lawrence Livermore National Laboratory (LLNL) lead team is developing and adapting generic and specific applications of commercial robotic technologies to SNM pyrochemical processing and other operations. A working gantry robot within a sealed processing glove box and a telerobot control test bed are manifestations of this effort. This paper describes the development challenges and progress in adapting processing, robotic, and nuclear safety technologies to the application. 3 figs.

  20. Nuclear Electronics: Superconducting Detectors and Processing Techniques

    Science.gov (United States)

    Polushkin, Vladimir

    2004-06-01

    With the commercialisation of superconducting particles and radiation detectors set to occur in the very near future, nuclear analytical instrumentation is taking a big step forward. These new detectors have a high degree of accuracy, stability and speed and are suitable for high-density multiplex integration in nuclear research laboratories and astrophysics. Furthermore, superconducting detectors can also be successfully applied to food safety, airport security systems, medical examinations, doping tests & forensic investigations. This book is the first to address a new generation of analytical tools based on new superconductor detectors demonstrating outstanding performance unsurpassed by any other conventional devices. Presenting the latest research and development in nanometer technologies and biochemistry this book: * Discusses the development of nuclear sensing techniques. * Provides guidance on the design and use of the next generation of detectors. * Describes cryogenic detectors for nuclear measurements and spectrometry. * Covers primary detectors, front-end readout electronics and digital signal processing. * Presents applications in nanotechnology and modern biochemistry including DNA sequencing, proteinomics, microorganisms. * Features examples of two applications in X-ray electron probe nanoanalysis and time-of-flight mass spectrometry. This comprehensive treatment is the ideal reference for researchers, industrial engineers and graduate students involved in the development of high precision nuclear measurements, nuclear analytical instrumentation and advanced superconductor primary sensors. This book will also appeal to physicists, electrical and electronic engineers in the nuclear industry.

  1. Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.; Nash, Charles A.; Crawford, Charles L.; McCabe, Daniel J.; Wilmarth, William R.

    2014-01-21

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also

  2. Experimental engineering section off-gas decontamination facility's fractionator column: installation and performance

    Energy Technology Data Exchange (ETDEWEB)

    Gilliam, T. M.; Fowler, V. L.; Inman, D. J.

    1978-03-01

    A detailed description of the third column recently installed in the Experimental Engineering Section Off-Gas Decontamination Facility (EES-ODF) is presented. The EES-ODF is being used to provide engineering-scale experiments (nominal gas and liquid flows of 5 scfm and 0.5 gpm, respectively) in the development of the Krypton Absorption in Liquid CO/sub 2/ (KALC) process. A detailed discussion of the column's construction is provided. This discussion includes the peripherals associated with the column, such as refrigeration, heat exchangers, instrumentation, etc. The compressibility of Goodloe packing (the packing in the other columns) and the possible reduced throughput due to this compression have revealed the desirablility of a random (i.e., noncompressible) packing. Toward this end, the third column is packed with a new random packing (PRO-PAK). A preliminary comparison between this packing and the woven wire mesh packing (Goodloe) used in the other two columns has been made. Experiments comparing the throughput capacity indicate that the PRO-PAK packing has approximately 60% the capacity of Goodloe for a CO/sub 2/ system. When used as a fractionator or stripper with the basic O/sub 2/-Kr-CO/sub 2/ KALC system, the PRO-PAK column produced HTU values less than or equal to the GOODLOE columns under similar operating conditions.

  3. Organic iodine removal from simulated dissolver off-gas streams using partially exchanged silver mordenite

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R.T.

    1982-01-01

    The removal of methyl iodide by adsorption onto silver mordenite was studied using a simulated off-gas from the fuel dissolution step of a nuclear fuel reprocessing plant. The methyl iodide adsorption of partially exchanged silver mordenite was examined for the effects of NO/sub x/, humidity, filter temperature, and degree of silver exchange. Partially exchanged silver mordenite, in general, achieved significantly higher silver utilizations than the fully exchanged material. Silver utilizations of > 95% were achieved, assuming the formation of AgI. The experimental results indicate that CH/sub 3/I loadings increase proportionally with silver loading up to 5 wt % silver and then appear to level off. Tests conducted to determine the effect of temperature on the loading showed higher loadings at 200/sup 0/C than at either 150 or 250/sup 0/C. The presence of NO, NO/sub 2/, and H/sub 2/O vapor showed negligible effects on the loading of CH/sub 3/I. In contrast to iodine loaded onto fully exchanged silver mordenite, the iodine loaded onto the partially exchanged silver mordenite could not be stripped by either 4.5% hydrogen or 100% hydrogen at temperatures up to 500/sup 0/C. A study of the regeneration characteristics of fully exchanged silver mordenite indicates a decreased adsorbent capacity after complete removal of the iodine with 4.5% hydrogen in the regeneration gas stream at 500/sup 0/C. The loss of adsorbent capacity was much higher for silver mordenite regenerated in a stainless steel filter housing than in a glass filter housing. A cost evaluation for the use of the partially exchanged silver mordenite shows that the cost of the silver mordenite on a once-through basis is < $10/h of operation for a 0.5-t/d reprocessing plant.

  4. LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K.; Nash, C.; McCabe, D.

    2014-09-29

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also

  5. Fabrication of ATALANTE Dissolver Off-Gas Sorbent-Based Capture System

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Jr., Joseph Franklin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    A small sorbent-based capture system was designed that could be placed in the off-gas line from the fuel dissolver in the ATALANTE hot cells with minimal modifications to the ATALANTE dissolver off-gas system. Discussions with personnel from the ATALANTE facility provided guidance that was used for the design. All components for this system have been specified, procured, and received on site at Oak Ridge National Laboratory (ORNL), meeting the April 30, 2015, milestone for completing the fabrication of the ATALANTE dissolver off-gas capture system. This system will be tested at ORNL to verify operation and to ensure that all design requirements for ATALANTE are met. Modifications to the system will be made, as indicated by the testing, before the system is shipped to ATALANTE for installation in the hot cell facility.

  6. Propulsion apparatus and method using boil-off gas from a cryogenic liquid

    Science.gov (United States)

    Blount, D. H. (Inventor)

    1986-01-01

    A propulsion system and method are disclosed for controlling the attitude and drag of a space vehicle. A helium dewar contains liquid helium which cools an experiment package. The helium is heated or vented to keep the temperature between 1.5 and 1.7 degrees K to maintain adequate helium boil-off gas as a propellant without adversely affecting the experiment package which is contained in the helium dewar for protection from solar heating. The apparatus includes auxiliary heater and temperature sensor for controlling the temperature of the helium. The boil-off gas propellant is delivered to thruster modules to control vehicle attutude and compensate for drag.

  7. Bench scale experiments for the remediation of Hanford Waste Treatment Plant low activity waste melter off-gas condensate

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Poirier, Michael [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-11

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  8. Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.; Nash, Charles A.; Crawford, Charles L.; McCabe, Daniel J.; Wilmarth, William R.

    2014-01-21

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also

  9. Estimation of the Waste Mass from a Pyro-Process of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo (and others)

    2008-04-15

    Pyro-Process is now developing to retrieve reusable uranium and TRU, and to reduce the volume of high level waste from a nuclear power plant. In this situation, it is strongly required for the estimation of expected masses and their physical properties of the wastes. In this report, the amount of wastes and their physical properties are presupposed through some assumptions in regard to 10MTHM of Oxide Fuel with 4.5wt% U-235, 45,000 MWD/MTU, and 5yrs cooling. The produced wastes can be divided into three categories such as metal, CWF(Ceramic Waste Form), and VWF(Vitrified Waste Form). The 42 nuclrides in a spent nuclear fuel are distributed into the waste categories on the their physical and thermodynamic properties when they exist in metal, oxide, or chloride forms. The treated atomic groups are Uranium, TRU, Noble metal, Rare earth, Alkali metal, Halogens, and others. The mass of each waste is estimated by the distribution results. The off-gas waste is included into a CWF. The heat generations by the wastes in this Pyro-Process are calculated using a ORIGEN-ARP program. It is possible to estimate the amounts of wastes and their heat generation rates in this Pyro-Process analysis. These information are very helpful to design a waste container and its quantity also can be determined. The number of container and its heat generation rate will be key factor for the construction of interim storage facilities including a underground disposal site.

  10. Methods to optimize myxobacterial fermentations using off-gas analysis

    Directory of Open Access Journals (Sweden)

    Hüttel Stephan

    2012-05-01

    Full Text Available Abstract Background The influence of carbon dioxide and oxygen on microbial secondary metabolite producers and the maintenance of these two parameters at optimal levels have been studied extensively. Nevertheless, most studies have focussed on their influence on specific product formation and condition optimization of established processes. Considerably less attention has been paid to the influence of reduced or elevated carbon dioxide and oxygen levels on the overall metabolite profiles of the investigated organisms. The synergistic action of both gases has garnered even less attention. Results We show that the composition of the gas phase is highly important for the production of different metabolites and present a simple approach that enables the maintenance of defined concentrations of both O2 and CO2 during bioprocesses over broad concentration ranges with a minimal instrumental setup by using endogenously produced CO2. The metabolite profiles of a myxobacterium belonging to the genus Chondromyces grown under various concentrations of CO2 and O2 showed considerable differences. Production of two unknown, highly cytotoxic compounds and one antimicrobial substance was found to increase depending on the gas composition. In addition, the observation of CO2 and O2 in the exhaust gas allowed optimization and control of production processes. Conclusions Myxobacteria are becoming increasingly important due to their potential for bioactive secondary metabolite production. Our studies show that the influence of different gas partial pressures should not be underestimated during screening processes for novel compounds and that our described method provides a simple tool to investigate this question.

  11. Preliminary Results from Electric Arc Furnace Off-Gas Enthalpy Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Nimbalkar, Sachin U [ORNL; Thekdi, Arvind [E3M Inc; Keiser, James R [ORNL; Storey, John Morse [ORNL

    2015-01-01

    This article describes electric arc furnace (EAF) off-gas enthalpy models developed at Oak Ridge National Laboratory (ORNL) to calculate overall heat availability (sensible and chemical enthalpy) and recoverable heat values (steam or power generation potential) for existing EAF operations and to test ORNL s new EAF waste heat recovery (WHR) concepts. ORNL s new EAF WHR concepts are: Regenerative Drop-out Box System and Fluidized Bed System. The two EAF off-gas enthalpy models described in this paper are: 1.Overall Waste Heat Recovery Model that calculates total heat availability in off-gases of existing EAF operations 2.Regenerative Drop-out Box System Model in which hot EAF off-gases alternately pass through one of two refractory heat sinks that store heat and then transfer it to another gaseous medium These models calculate the sensible and chemical enthalpy of EAF off-gases based on the off-gas chemical composition, temperature, and mass flow rate during tap to tap time, and variations in those parameters in terms of actual values over time. The models provide heat transfer analysis for the aforementioned concepts to confirm the overall system and major component sizing (preliminary) to assess the practicality of the systems. Real-time EAF off-gas composition (e.g., CO, CO2, H2, and H2O), volume flow, and temperature data from one EAF operation was used to test the validity and accuracy of the modeling work. The EAF off-gas data was used to calculate the sensible and chemical enthalpy of the EAF off-gases to generate steam and power. The article provides detailed results from the modeling work that are important to the success of ORNL s EAF WHR project. The EAF WHR project aims to develop and test new concepts and materials that allow cost-effective recovery of sensible and chemical heat from high-temperature gases discharged from EAFs.

  12. Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

    Energy Technology Data Exchange (ETDEWEB)

    First, M.W. [ed.] [Harvard Univ., Boston, MA (United States). Harvard Air Cleaning Lab.

    1991-02-01

    Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

  13. Nuclear energy an introduction to the concepts, systems, and applications of nuclear processes

    CERN Document Server

    Murray, Raymond L; Murphy, Arthur T; Rosenthal, Daniel I

    1987-01-01

    Nuclear Energy: An Introduction to the Concepts, Systems, and Applications of Nuclear Processes introduces the reader to the concepts, systems, and applications of nuclear processes. It provides a factual description of basic nuclear phenomena, as well as devices and processes that involve nuclear reactions. The problems and opportunities that are inherent in a nuclear age are also highlighted.Comprised of 27 chapters, this book begins with an overview of fundamental facts and principles, with emphasis on energy and states of matter, atoms and nuclei, and nuclear reactions. Radioactivi

  14. Analysis of copper losses throughout weak acid effluent flows generated during off-gas treatment in the New Copper Smelter RTB Bor

    Directory of Open Access Journals (Sweden)

    Dragana Ivšić-Bajčeta

    2013-09-01

    Full Text Available The previous inadequate treatment of off-gas in RTB Bor in Serbia has resulted in serious pollution of the environment and the possibly high losses of copper through the effluent flows. The project of New Copper Smelter RTB Bor, besides the new flash smelting furnace (FSF and the reconstruction of Pierce-Smith converter (PSC, includes more effective effluent treatment. Paper presents an analysis of the new FSF and PSC off-gas treatment, determination of copper losses throughout generated wastewaters and discussion of its possible valorization. Assumptions about the solubility of metals phases present in the FSF and PSC off-gas, obtained by the treatment process simulation, were compared with the leaching results of flue dusts. Determined wastewaters characteristics indicate that the PSC flow is significantly richer in copper, mostly present in insoluble metallic/sulfide form, while the FSF flow has low concentration of copper in the form of completely soluble oxide/sulfate. The possible scenario for the copper valorization, considering arsenic and lead as limiting factors, is the separation of the FSF and PSC flows, return of the metallic/sulfide solid phase to the smelting process and recovery from the sulfate/oxide liquid phase.

  15. Anomalous Nuclear Phenomena Associated with Ultrafast Processes

    Science.gov (United States)

    Jiang, Xingliu; Zhou, Xiaoping; Han, Lijun; Wang, Liyin

    2007-03-01

    Localized nuclear reactions on the tips of the surface of electrodes in electrolysis cells have been observed by using solid detectors CR-39 and autoradiography in our laboratory at the period of May, 1989. A physical model of transient vortex dynamics with torsion coherence with the zero point energy has been proposed by Xingliu Jiang based on the ultrafast processes of tripple phases area of tip effect on the electrode surface. Considering the large equivelent capacitance of electrochemical double layer, it is presumed that the double layer can exhibit nonlinear electrical response with spatial and temporal variations confined to micreoscopic areas by tip effect. Recent work reveals that nuclear reactions which usually occur at the field of high energy states, could be created in the systems of far from equilibrium with nonlinear beharvior at room tempurature.Our current understanging of science is like a puzzle with a large missing piece-zero point energy. Jiang Xingliu, Lei Jinzhi, Torsion field and tapping the zero-point energy in an electrochemical system, J. of New Energy, 4(2), 93(1999). B. Naranjo, J.K. Gimzewski & S. Putterman, Observation of nuclear fusion driven by a pyroelectric crystal, Nature, 434, 1115(2005).

  16. Quantum information processing and nuclear magnetic resonance

    CERN Document Server

    Cummins, H K

    2001-01-01

    as spectrometer pulse sequence programs. Quantum computers are information processing devices which operate by and exploit the laws of quantum mechanics, potentially allowing them to solve problems which are intractable using classical computers. This dissertation considers the practical issues involved in one of the more successful implementations to date, nuclear magnetic resonance (NMR). Techniques for dealing with systematic errors are presented, and a quantum protocol is implemented. Chapter 1 is a brief introduction to quantum computation. The physical basis of its efficiency and issues involved in its implementation are discussed. NMR quantum information processing is reviewed in more detail in Chapter 2. Chapter 3 considers some of the errors that may be introduced in the process of implementing an algorithm, and high-level ways of reducing the impact of these errors by using composite rotations. Novel general expressions for stabilising composite rotations are presented in Chapter 4 and a new class o...

  17. Effect of nuclear viscosity on fission process

    Energy Technology Data Exchange (ETDEWEB)

    Li Shidong; Kuang Huishun; Zhang Shufa; Xing Jingru; Zhuo Yizhong; Wu Xizhen; Feng Renfa

    1989-02-01

    According to the fission diffusion model, the deformation motion of fission nucleuses is regarded as a diffusion process of quasi-Brownian particles under fission potential. Through simulating such Brownian motion in two dimensional phase space by Monte-Carlo mehtod, the effect of nuclear visocity on Brownian particle diffusion is studied. Dynamical quanties, such as fission rate, kinetic energy distribution on scission, and soon are numerically calculated for various viscosity coefficients. The results are resonable in physics. This method can be easily extended to deal with multi-dimensional diffusion problems.

  18. Comparison of thermochemically calculated and measured dioxin contents in the off-gas of a sinter plant

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, P.; Eriksson, G.; Neuschuelz, D. [Lehrstuhl fuer Theoretische Huettenkunde, Aachen (Germany)

    1997-12-31

    Polychlorinated dibenzo-p-dioxins and dibenzo-furans form a family of more than 200 compounds which are relatively stable in the biosphere and tend to accumulate in the human body. The tetra- to hexa-chlorinated dioxins and furans are considered highly toxic. To facilitate the assessment of the total toxicity of dioxin and furan mixtures, the estimated toxic effects of the individual compounds relative to the 2,3,7,8-tetrachloro-dibenzo-p-dioxin (TCDD) were introduced as Toxic Equivalent Factors which yield, when multiplied with the respective concentrations, the Toxic Equivalent (TE) of the mixture. Toxic dioxins and furans are unintentionally formed in a number of industrial combustion processes such as waste incineration and iron ore sintering, in the chemical industry and in household heating. To keep the emissions as low as possible, off-gas clearing systems for the collection of dioxins and furans are increasingly prescribed by the authorities. In addition, it appears desirable to select process conditions that are unfavourable for the formation of these compounds. A simulation of the relevant processes on the basis of thermodynamic data may be helpful in defining such process conditions. To simulate dioxin formation in the sintering process, all major gas-solid reactions taking place in the sinter bed must also be simulated. A sufficiently accurate reproduction of the off-gas compositions along the length of the sinter strand requires detailed assumptions concerning the relative amounts of `active` O{sub 2} as well as the distribution of reacting carbon and water over the strand length. From this basis, an equilibrium calculation for the gas/solid reactions at the sintering temperature of 1150 deg C and an equilibrium calculation restricted to the gas phase at 700 deg C produced values for the concentrations of the major off-gas constituents in very good agreement with the measured values. The further assumption that below 700 deg C all reactions are frozen

  19. Comparison between reverse Brayton and Kapitza based LNG boil-off gas reliquefaction system using exergy analysis

    Science.gov (United States)

    Kochunni, Sarun Kumar; Chowdhury, Kanchan

    2017-02-01

    LNG boil-off gas (BOG) reliquefaction systems in LNG carrier ships uses refrigeration devices which are based on reverse Brayton, Claude, Kapitza (modified Claude) or Cascade cycles. Some of these refrigeration devices use nitrogen as the refrigerants and hence nitrogen storage vessels or nitrogen generators needs to be installed in LNG carrier ships which consume space and add weight to the carrier. In the present work, a new configuration based on Kapitza liquefaction cycle which uses BOG itself as working fluid is proposed and has been compared with Reverse Brayton Cycle (RBC) on sizes of heat exchangers and compressor operating parameters. Exergy analysis is done after simulating at steady state with Aspen Hysys 8.6® and the comparison between RBC and Kapitza may help designers to choose reliquefaction system with appropriate process parameters and sizes of equipment. With comparable exergetic efficiency as that of an RBC, a Kaptiza system needs only BOG compressor without any need of nitrogen gas.

  20. Formation of the ZnFe{sub 2}O{sub 4} phase in an electric arc furnace off-gas treatment system

    Energy Technology Data Exchange (ETDEWEB)

    Suetens, T., E-mail: thomas.suetens@mtm.kuleuven.be; Guo, M., E-mail: muxing.guo@mtm.kuleuven.be; Van Acker, K., E-mail: karel.vanacker@lrd.kuleuven.be; Blanpain, B., E-mail: bart.blanpain@mtm.kuleuven.be

    2015-04-28

    Highlights: • EAF dust was characterized with particle size analysis, XRF, and EPMA. • Slag particles showed no sign of reaction with Zn vapor. • Fe{sub 2}O{sub 3} particles showed different degrees of reaction based on their size. • The thermodynamic stability of Zn vapor in EAF off-gas ducts was reevaluated. • In presence of Fe{sub 2}O{sub 3}, Zn vapor reacts to form ZnFe{sub 2}O{sub 4} and ZnO. - Abstract: To better understand the phenomena of ZnFe{sub 2}O{sub 4} spinel formation in electric arc furnace dust, the dust was characterized with particle size analysis, X-ray fluorescence (XRF), electron backscatter diffraction (EBSD), and electron probe micro-analysis (EPMA). Different ZnFe{sub 2}O{sub 4} formation reaction extents were observed for iron oxide particles with different particle sizes. ZnO particles were present as both individual particles and aggregated on the surface of larger particles. Also, the slag particles found in the off-gas were shown not to react with the zinc vapor. After confirming the presence of a ZnFe{sub 2}O{sub 4} formation reaction, the thermodynamic feasibility of in-process separation – a new electric arc furnace dust treatment technology – was reevaluated. The large air intake and the presence of iron oxide particles in the off-gas were included into the thermodynamic calculations. The formation of the stable ZnFe{sub 2}O{sub 4} spinel phase was shown to be thermodynamically favorable in current electric arc furnace off-gas ducts conditions even before reaching the post combustion chamber.

  1. Continuous determination of bath carbon content on 150 t BOF by off-gas analyzer

    Institute of Scientific and Technical Information of China (English)

    Zhigang Hu; Ping He; Mingxiang Tan; Liu Liu

    2003-01-01

    The first imported off-gas analysis system on 150 t BOF at Benxi Plates Co. Ltd. is presented and the continuous determination of bath carbon content has been studied. The comparison between the whole-course carbon integral model and the end-point carbon prediction model has been made. The results show that the regular change of CO, CO2 and N2 content in the off-gas during blowing plays an important role in judging the smelting end-point of converter; the cubic curve fitting model has a higher hit rate over 95% for the heats whose end-point carbon content is lower than 0.10% with a precision of ±0.02% and has a large error for the heats whose end-point carbon content is more than 0.15%.

  2. Effects of headspace and oxygen level on off-gas emissions from wood pellets in storage.

    Science.gov (United States)

    Kuang, Xingya; Shankar, Tumuluru Jaya; Sokhansanj, Shahab; Lim, C Jim; Bi, Xiaotao T; Melin, Staffan

    2009-11-01

    Few papers have been published in the open literature on the emissions from biomass fuels, including wood pellets, during the storage and transportation and their potential health impacts. The purpose of this study is to provide data on the concentrations, emission factors, and emission rate factors of CO(2), CO, and CH(4) from wood pellets stored with different headspace to container volume ratios with different initial oxygen levels, in order to develop methods to reduce the toxic off-gas emissions and accumulation in storage spaces. Metal containers (45 l, 305 mm diameter by 610 mm long) were used to study the effect of headspace and oxygen levels on the off-gas emissions from wood pellets. Concentrations of CO(2), CO, and CH(4) in the headspace were measured using a gas chromatograph as a function of storage time. The results showed that the ratio of the headspace ratios and initial oxygen levels in the storage space significantly affected the off-gas emissions from wood pellets stored in a sealed container. Higher peak emission factors and higher emission rates are associated with higher headspace ratios. Lower emissions of CO(2) and CO were generated at room temperature under lower oxygen levels, whereas CH(4) emission is insensitive to the oxygen level. Replacing oxygen with inert gases in the storage space is thus a potentially effective method to reduce the biomass degradation and toxic off-gas emissions. The proper ventilation of the storage space can also be used to maintain a high oxygen level and low concentrations of toxic off-gassing compounds in the storage space, which is especially useful during the loading and unloading operations to control the hazards associated with the storage and transportation of wood pellets.

  3. Advanced Off-Gas Control System Design For Radioactive And Mixed Waste Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg

    2005-09-01

    Treatment of radioactive and mixed wastes is often required to destroy or immobilize hazardous constituents, reduce waste volume, and convert the waste to a form suitable for final disposal. These kinds of treatments usually evolve off-gas. Air emission regulations have become increasingly stringent in recent years. Mixed waste thermal treatment in the United States is now generally regulated under the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards. These standards impose unprecedented requirements for operation, monitoring and control, and emissions control. Off-gas control technologies and system designs that were satisfactorily proven in mixed waste operation prior to the implementation of new regulatory standards are in some cases no longer suitable in new mixed waste treatment system designs. Some mixed waste treatment facilities have been shut down rather than have excessively restrictive feed rate limits or facility upgrades to comply with the new standards. New mixed waste treatment facilities in the U. S. are being designed to operate in compliance with the HWC MACT standards. Activities have been underway for the past 10 years at the INL and elsewhere to identify, develop, demonstrate, and design technologies for enabling HWC MACT compliance for mixed waste treatment facilities. Some specific off-gas control technologies and system designs have been identified and tested to show that even the stringent HWC MACT standards can be met, while minimizing treatment facility size and cost.

  4. Quantum information processing through nuclear magnetic resonance

    Energy Technology Data Exchange (ETDEWEB)

    Bulnes, J.D.; Sarthour, R.S.; Oliveira, I.S. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Bonk, F.A.; Azevedo, E.R. de; Bonagamba, T.J. [Sao Paulo Univ., Sao Carlos, SP (Brazil). Inst. de Fisica; Freitas, J.C.C. [Espirito Santo Univ., Vitoria, ES (Brazil). Dept. de Fisica

    2005-09-15

    We discuss the applications of Nuclear Magnetic Resonance (NMR) to quantum information processing, focusing on the use of quadrupole nuclei for quantum computing. Various examples of experimental implementation of logic gates are given and compared to calculated NMR spectra and their respective density matrices. The technique of Quantum State Tomography for quadrupole nuclei is briefly described, and examples of measured density matrices in a two-qubit I = 3/2 spin system are shown. Experimental results of density matrices representing pseudo-Bell states are given, and an analysis of the entropy of theses states is made. Considering an NMR experiment as a depolarization quantum channel we calculate the entanglement fidelity and discuss the criteria for entanglement in liquid state NMR quantum information. A brief discussion on the perspectives for NMR quantum computing is presented at the end. (author)

  5. Nuclear parton distributions and the Drell-Yan process

    Science.gov (United States)

    Kulagin, S. A.; Petti, R.

    2014-10-01

    We study the nuclear parton distribution functions on the basis of our recently developed semimicroscopic model, which takes into account a number of nuclear effects including nuclear shadowing, Fermi motion and nuclear binding, nuclear meson-exchange currents, and off-shell corrections to bound nucleon distributions. We discuss in detail the dependencies of nuclear effects on the type of parton distribution (nuclear sea vs valence), as well as on the parton flavor (isospin). We apply the resulting nuclear parton distributions to calculate ratios of cross sections for proton-induced Drell-Yan production off different nuclear targets. We obtain a good agreement on the magnitude, target and projectile x, and the dimuon mass dependence of proton-nucleus Drell-Yan process data from the E772 and E866 experiments at Fermilab. We also provide nuclear corrections for the Drell-Yan data from the E605 experiment.

  6. The probabilities of nuclear processes; Probabilidades de los procesos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez del Rio, C.

    1972-07-01

    This report is the first draft of one of the chapters of a book being prepared under the title: Topics on Practical Nuclear Physics. It is published as a report because of its immediate educational value and in order to include in its final draft the suggestions of the readers. (Author)

  7. 费托合成过程模拟及尾气处理分析%Simulation of Fischer-Tropsch synthesis and analysis of off gas treatment

    Institute of Scientific and Technical Information of China (English)

    陆雪峰; 孙力; 贺高红

    2012-01-01

    费托(F-T)合成将煤炭转化为一种类似石油的液体燃料,是一种煤间接液化技术.其流程主要包括五部分:煤气化,水气变换,气体净化,合成液体燃料和联合循环发电.合成液体燃料过程除了生成液体燃料产品之外,同时产生尾气,其主要成分有CO、H2、C1-C4等轻烷烃,并含有少量重烷烃,尾气经过燃烧可以生成电和蒸汽,满足系统公用工程需求,而如果回收其中CO和H2等有效组分,可增加燃料产品产量.本文应用化工流程模拟软件Aspen Plus对费托合成全流程的五部分进行模拟,并着重对其尾气处理进行分析.本文研究3种尾气处理方法:尾气全部燃烧生成电和蒸汽,满足系统公用工程需求,如有剩余可输送至外围电网和蒸汽管网;尾气回收H2和CO等有效组分,增加产品产量,并通过烧煤满足系统公用工程需求;尾气部分燃烧,部分回收有效组分,如果过程的电和蒸汽供应不足需燃煤补充,本例设置不同比例的尾气回收方案.通过过程工艺流程和公用工程系统分析,综合比较各方案的系统操作费和过程收益,确定尾气20%回收经济效益最大,而尾气全部回收时经济效益最小.%Fischer-Tropsch (F-T) synthesis, which can transfer the non-fossil fuels to liquid fuels, is one of the principal coal to liquid oil technologies. The F-T process includes five parts: coal gasification, water gas shift, gas purification, liquids oil synthesis, and combined cycle power. Besides the liquid production can be generated in the process of liquid oil synthesis, off gas is also produced simultaneously. The off gas contains some CO, H2, C1-C4 light alkanes, and a small amount of heavy alkanes. Therefore, the off gas can be burned off to produce power and steam which must satisfy the requirement of the process. In addition, its components, such as CO and H2, can be recovered in order to increase the output of the product. In mis paper, the whole

  8. Nuclear Parton Distributions and the Drell-Yan Process

    CERN Document Server

    Kulagin, S A

    2014-01-01

    We study the nuclear parton distribution functions basing on our recently developed semi-microscopic model, which takes into account a number of nuclear effects including nuclear shadowing, Fermi motion and nuclear binding, nuclear meson-exchange currents and off-shell corrections to bound nucleon distributions. We discuss in details the dependencies of nuclear effects on the type of parton distribution (nuclear sea vs. valence) as well as on the parton flavour (isospin). The resulting nuclear parton distributions are applied to calculate the ratios of cross sections for proton-induced Drell-Yan production off different nuclear targets. We obtain a good agreement on the magnitude, target and projectile x and the dimuon mass dependence of proton-nucleus Drell-Yan process data from the E772 and E866 experiments at Fermilab.

  9. Intrinsic and extrinsic negative regulators of nuclear protein transport processes

    OpenAIRE

    Sekimoto, Toshihiro; Yoneda, Yoshihiro

    2012-01-01

    The nuclear–cytoplasmic protein transport is a critical process in cellular events. The identification of transport signals (nuclear localization signal and nuclear export signal) and their receptors has facilitated our understanding of this expanding field. Nuclear transport must be appropriately regulated to deliver proteins through the nuclear pore when their functions are required in the nucleus, and to export them into the cytoplasm when they are not needed in the nucleus. Altered nuclea...

  10. Nucleation Process in Asymmetric Nuclear Matter

    CERN Document Server

    Peres-Menezes, D

    1998-01-01

    An extended version of the non linear Walecka model, with rho mesons and eletromagnetic field is used to investigate the possibility of phase transitions in hot (warm) nuclear matter, giving rise to droplet formation. Surface properties of asymmetric nuclear matter as the droplet surface energy and its thickness are also examined.

  11. Integration of advanced nuclear materials separation processes

    Energy Technology Data Exchange (ETDEWEB)

    Jarvinen, G.D.; Worl, L.A.; Padilla, D.D.; Berg, J.M.; Neu, M.P.; Reilly, S.D.; Buelow, S.

    1998-12-31

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). This project has examined the fundamental chemistry of plutonium that affects the integration of hydrothermal technology into nuclear materials processing operations. Chemical reactions in high temperature water allow new avenues for waste treatment and radionuclide separation.Successful implementation of hydrothermal technology offers the potential to effective treat many types of radioactive waste, reduce the storage hazards and disposal costs, and minimize the generation of secondary waste streams. The focus has been on the chemistry of plutonium(VI) in solution with carbonate since these are expected to be important species in the effluent from hydrothermal oxidation of Pu-containing organic wastes. The authors investigated the structure, solubility, and stability of the key plutonium complexes. Installation and testing of flow and batch hydrothermal reactors in the Plutonium Facility was accomplished. Preliminary testing with Pu-contaminated organic solutions gave effluent solutions that readily met discard requirements. A new effort in FY 1998 will build on these promising initial results.

  12. Removal efficiency of silver impregnated filter materials and performance of iodie filters in the off-gas of the Karlsruhe reprocessing plant WAK

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, F.J.; Herrmann, B.; Hoeflich, V. [Wiederaufarbeitungsanlage Karlsruhe (Germany)] [and others

    1997-08-01

    An almost quantitative retention of iodine is required in reprocessing plants. For the iodine removal in the off-gas streams of a reprocessing plant various sorption materials had been tested under realistic conditions in the Karlsruhe reprocessing plant WAK in cooperation with the Karlsruhe research center FZK. The laboratory results achieved with different iodine sorption materials justified long time performance tests in the WAK Plant. Technical iodine filters and sorption materials for measurements of iodine had been tested from 1972 through 1992. This paper gives an overview over the most important results, Extended laboratory, pilot plant, hot cell and plant experiences have been performed concerning the behavior and the distribution of iodine-129 in chemical processing plants. In a conventional reprocessing plant for power reactor fuel, the bulk of iodine-129 and iodine-127 is evolved into the dissolver off-gas. The remainder is dispersed over many aqueous, organic and gaseous process and waste streams of the plant. Iodine filters with silver nitrate impregnated silica were installed in the dissolver off-gas of the Karlsruhe reprocessing plant WAK in 1975 and in two vessel vent systems in 1988. The aim of the Karlsruhe iodine research program was an almost quantitative evolution of the iodine during the dissolution process to remove as much iodine with the solid bed filters as possible. After shut down of the WAK plant in December 1990 the removal efficiency of the iodine filters at low iodine concentrations had been investigated during the following years. 12 refs., 2 figs., 2 tabs.

  13. System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Daryl R. Haefner; Jack D. Law; Troy J. Tranter

    2010-08-01

    This document provides descriptions of the off-gases evolved during spent nuclear fuel processing and the systems used to capture the gases of concern. Two reprocessing techniques are discussed, namely aqueous separations and electrochemical (pyrochemical) processing. The unit operations associated with each process are described in enough detail so that computer models to mimic their behavior can be developed. The document also lists the general requirements for the desired computer models.

  14. Considerations about the licensing process of special nuclear industrial facilities

    Energy Technology Data Exchange (ETDEWEB)

    Talarico, M.A., E-mail: talaricomarco@hotmail.com [Marinha do Brasil, Rio de Janeiro, RJ (Brazil). Coordenacao do Porgrama de Submarino com Propulsao Nuclear; Melo, P.F. Frutuoso e [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

  15. Characterization and kinetics study of off-gas emissions from stored wood pellets.

    Science.gov (United States)

    Kuang, Xingya; Shankar, Tumuluru Jaya; Bi, Xiaotao T; Sokhansanj, Shahab; Lim, C Jim; Melin, Staffan

    2008-11-01

    The full potential health impact from the emissions of biomass fuels, including wood pellets, during storage and transportation has not been documented in the open literature. The purpose of this study is to provide data on the concentration of CO(2), CO and CH(4) from wood pellets stored in sealed vessels and to develop a kinetic model for predicting the transient emission rate factors at different storage temperatures. Five 45-l metal containers (305 mm diameter by 610 mm long) equipped with heating and temperature control devices were used to study the temperature effect on the off-gas emissions from wood pellets. Concurrently, ten 2-l aluminum canisters (100 mm diameter by 250 mm long) were used to study the off-gas emissions from different types of biomass materials. Concentrations of CO(2), CO and CH(4) were measured by a gas chromatograph as a function of storage time and storage temperature. The results showed that the concentrations of CO, CO(2) and CH(4) in the sealed space of the reactor increased over time, fast at the beginning but leveling off after a few days. A first-order reaction kinetics fitted the data well. The maximum concentration and the time it takes for the buildup of gas concentrations can be predicted using kinetic equations.

  16. Design, Fabrication, and Shakeout Testing of ATALANTE Dissolver Off-Gas Sorbent-Based Capture System

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Jr, Joseph Franklin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-07-31

    A sorbent-based capture system designed for integration into the existing dissolver off-gas (DOG) treatment system at the ATelier Alpha et Laboratoires pour ANalyses, Transuraniens et Etudes de retraitement (ATALANTE) facility has been successfully designed and fabricated and has undergone shakeout testing. Discussions with personnel from the ATALANTE facility provided guidance that was used for the design. All components for this system were specified, procured, and received on site at Oak Ridge National Laboratory (ORNL). The system was then fabricated and tested at ORNL to verify operation. Shakeout testing resulted in a simplified system. This system should be easily installed into the existing facility and should be straightforward to operate during future experimental testing. All parts were selected to be compatible with ATALANTE power supplies, space requirements, and the existing DOG treatment system. Additionally, the system was demonstrated to meet all of four design requirements. These include (1) a dissolver off-gas flow rate of ≤100 L/h (1.67 L/min), (2) an external temperature of ≤50°C for all system components placed in the hot cell, (3) a sorbent bed temperature of ~150°C, and (4) a gas temperature of ~150°C upon entry into the sorbent bed. The system will be ready for shipment and installation in the existing DOG treatment system at ATALANTE in FY 2016.

  17. Design, Fabrication, and Shakeout Testing of ATALANTE Dissolver Off-Gas Sorbent-Based Capture System

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Jr, Joseph Franklin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-07-31

    A sorbent-based capture system designed for integration into the existing dissolver off-gas (DOG) treatment system at the ATelier Alpha et Laboratoires pour ANalyses, Transuraniens et Etudes de retraitement (ATALANTE) facility has been successfully designed and fabricated and has undergone shakeout testing. Discussions with personnel from the ATALANTE facility provided guidance that was used for the design. All components for this system were specified, procured, and received on site at Oak Ridge National Laboratory (ORNL). The system was then fabricated and tested at ORNL to verify operation. Shakeout testing resulted in a simplified system. This system should be easily installed into the existing facility and should be straightforward to operate during future experimental testing. All parts were selected to be compatible with ATALANTE power supplies, space requirements, and the existing DOG treatment system. Additionally, the system was demonstrated to meet all of four design requirements. These include (1) a dissolver off-gas flow rate of ≤100 L/h (1.67 L/min), (2) an external temperature of ≤50°C for all system components placed in the hot cell, (3) a sorbent bed temperature of ~150°C, and (4) a gas temperature of ~150°C upon entry into the sorbent bed. The system will be ready for shipment and installation in the existing DOG treatment system at ATALANTE in FY 2016.

  18. Steady-state and dynamic simulation study on boil-off gas minimization and recovery strategies at LNG exporting terminals

    Science.gov (United States)

    Kurle, Yogesh

    Liquefied natural gas (LNG) is becoming one of the prominent clean energy sources with its abundance, high calorific value, low emission, and price. Vapors generated from LNG due to heat leak are called boil-off gas (BOG). As world-wide LNG productions are increasing fast, BOG generation and handling problems are becoming more critical. Also, due to stringent environmental regulations, flaring of BOG is not a viable option. In this study, typical Propane-and-Mixed-Refrigerant (C3-MR) process, storage facilities, and loading facilities are modeled and simulated to study BOG generation at LNG exporting terminals, including LNG processing, storage, and berth loading areas. Factors causing BOG are presented, and quantities of BOG generated due to each factor at each location are calculated under different LNG temperatures. Various strategies to minimize, recover, and reuse BOG are also studied for their feasibility and energy requirements. Rate of BOG generation during LNG loading---Jetty BOG (JBOG)---changes significantly with loading time. In this study, LNG vessel loading is simulated using dynamic process simulation software to obtain JBOG generation profile and to study JBOG recovery strategies. Also, fuel requirements for LNG plant to run steam-turbine driven compressors and gas-turbine driven compressors are calculated. Handling of JBOG generated from multiple loadings is also considered. The study would help proper handling of BOG problems in terms of minimizing flaring at LNG exporting terminals, and thus reducing waste, saving energy, and protecting surrounding environments.

  19. Development of Curricula for Nuclear Radiation Protection, Nuclear Instrumentation, and Nuclear Materials Processing Technologies. Final Report.

    Science.gov (United States)

    Hull, Daniel M.

    A study was conducted to assist two-year postsecondary educational institutions in providing technical specialty courses for preparing nuclear technicians. As a result of project activities, curricula have been developed for five categories of nuclear technicians and operators: (1) radiation protection technician, (2) nuclear instrumentation and…

  20. Detailed off-gas measurements for improved modelling of the aeration performance at the WWTP of Eindhoven.

    Science.gov (United States)

    Amerlinck, Y; Bellandi, G; Amaral, A; Weijers, S; Nopens, I

    2016-01-01

    At wastewater treatment plants (WWTPs), the aerobic conversion processes in the bioreactor are driven by the presence of dissolved oxygen (DO). Within these conversion processes, the oxygen transfer is a rate limiting step as well as being the largest energy consumer. Despite this high importance, WWTP models often lack detail on the aeration part. An extensive measurement campaign with off-gas tests was performed at the WWTP of Eindhoven to provide more information on the performance and behaviour of the aeration system. A high spatial and temporal variability in the oxygen transfer efficiency was observed. Applying this gathered system knowledge in the aeration model resulted in an improved prediction of the DO concentrations. Moreover, an important consequence of this was that ammonium predictions could be improved by resetting the ammonium half-saturation index for autotrophs to its default value. This again proves the importance of balancing sub-models with respect to the need for model calibration as well as model predictive power.

  1. Efficiency analysis of a hydrogen-fueled solid oxide fuel cell system with anode off-gas recirculation

    Science.gov (United States)

    Peters, Roland; Deja, Robert; Engelbracht, Maximilian; Frank, Matthias; Nguyen, Van Nhu; Blum, Ludger; Stolten, Detlef

    2016-10-01

    This study analyzes different hydrogen-fueled solid oxide fuel cell (SOFC) system layouts. It begins with a simple system layout without any anode off-gas recirculation, continues with a configuration equipped with off-gas recirculation, including steam condensation and then considers a layout with a dead-end anode off-gas loop. Operational parameters such as stack fuel utilization, as well as the recirculation rate, are modified, with the aim of achieving the highest efficiency values. Drawing on experiments and the accumulated experience of the SOFC group at the Forschungszentrum Jülich, a set of operational parameters were defined and applied to the simulations. It was found that anode off-gas recirculation, including steam condensation, improves electrical efficiency by up to 11.9 percentage-points compared to a layout without recirculation of the same stack fuel utilization. A system layout with a dead-end anode off-gas loop was also found to be capable of reaching electrical efficiencies of more than 61%.

  2. Letter report: Evaluation of LFCM off-gas system technologies for the HWVP

    Energy Technology Data Exchange (ETDEWEB)

    Goles, R.W.; Mishima, J.; Schmidt, A.J.

    1996-03-01

    Radioactive high-level liquid waste (HLLW), a byproduct of defense nuclear fuel reprocessing activities, is currently being stored in underground tanks at several US sites. Because its mobility poses significant environmental risks, HLLW is not a suitable waste form for long-term storage. Thus, high-temperature processes for solidifying and isolating the radioactive components of HLLW have been developed and demonstrated by the US Department of Energy (DOE) and its contractors. Vitrification using liquidfed ceramic melters (LFCMs) is the reference process for converting US HLLW into a borosilicate glass. Two vitrification plants are currently under construction in the United States: the West Valley Demonstration Plant (WVDP) being built at the former West Valley Nuclear Fuels Services site in West Valley, New York; and the Defense Waste Processing Facility (DWPF), which is currently 85% complete at DOE`s Savannah River Plant (SRP). A third facility, the Hanford Waste Vitrification Plant (HWVP), is being designed at DOE`s Hanford Site.

  3. FY'99 final report for the expedited technology demonstration project: demonstration test results for the MSO/off-gas and salt recycle system

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M G; Hsu, P C

    1999-05-01

    Molten Salt Oxidation (MSO) is a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility in which an integrated pilot-scale MSO treatment system is being tested and demonstrated. The system consists of a MSO vessel with a dedicated off-gas treatment system, a salt recycle system, feed preparation equipment, and a ceramic final waste forms immobilization system. This integrated system was designed and engineered based on operational experience with an engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. The MSO/off-gas system has been operational since December 1997. The salt recycle system and the ceramic final forms immobilization became operational in May 1998. In FY98, we have tested the MSO facility with various organic feeds, including chlorinated solvents, tributyl phosphate/kerosene, PCB-contaminated waste oils and solvents, booties, plastic pellets, ion exchange resins, activated carbon, radioactive-spiked organics, and well-characterized low-level liquid mixed wastes. MSO is shown to be a versatile technology for hazardous waste treatment and may be a solution to many waste disposal problems in DOE sites. The results of the demonstration conducted in FY98 has been reported [1]. In FY99 (October 1998 to April 1999) we conducted further testing in the MSO/off-gas system with ion exchange resins, two real waste specimens, activated carbon, and TNT-loaded activated carbon, both at regular feed rates and higher feed rates up to a superficial gas velocity of 1.75 ft/s. We also drained the salt three times (SR7, SR8, SR9) in FY99 and sent the spent salts to the salt recycle system for further processing. This report presents the results obtained from the demonstration of the MSO/off-gas system and the salt recycle system from October 1998 to April 1999. We then shut down the operation and cleaned the

  4. Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

  5. Thorium utilization program. Quarterly progress report for the period ending May 31, 1976. [Fuel element crushing, solids handling, fluidized-bed combustion, aqueous separations, solvent extraction, off-gas studies, semiremote handling systems, alternative head-end processing, and fuel recycle design

    Energy Technology Data Exchange (ETDEWEB)

    1976-06-30

    The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel and the design and development of an integrated line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials. Work is also described on trade-off studies concerning the required design of recycle facilities for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.

  6. A Novel Boil-Off Gas Re-Liquefaction Using a Spray Recondenser for Liquefied Natural-Gas Bunkering Operations

    Directory of Open Access Journals (Sweden)

    Jiheon Ryu

    2016-11-01

    Full Text Available This study presents the design of a novel boil-off gas (BOG re-liquefaction technology using a BOG recondenser system. The BOG recondenser system targets the liquefied natural gas (LNG bunkering operation, in which the BOG phase transition occurs in a pressure vessel instead of a heat exchanger. The BOG that is generated during LNG bunkering operation is characterized as an intermittent flow with various peak loads. The system was designed to temporarily store the transient BOG inflow, condense it with subcooled LNG and store the condensed liquid. The superiority of the system was verified by comparing it with the most extensively employed conventional re-liquefaction system in terms of consumption energy and via an exergy analysis. Static simulations were conducted for three compositions; the results indicated that the proposed system provided 0 to 6.9% higher efficiencies. The exergy analysis indicates that the useful work of the conventional system is 24.9%, and the useful work of the proposed system is 26.0%. Process dynamic simulations of six cases were also performed to verify the behaviour of the BOG recondenser system. The results show that the pressure of the holdup in the recondenser vessel increased during the BOG inflow mode and decreased during the initialization mode. The maximum pressure of one of the bunkering cases was 3.45 bar. The system encountered a challenge during repetitive operations due to overpressurizing of the BOG recondenser vessel.

  7. Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-01-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter

  8. Nuclear fuels: Development, processing and disposal

    Energy Technology Data Exchange (ETDEWEB)

    Allday, C.

    1982-08-01

    The successful development of the world's energy resources has enabled industries in the more advanced countries to provide the economic basis on which improved living standards are based. As the less well-developed countries seek to improve their standards of living the pressure on existing energy resources will increase. In this context it is essential not to allow the current industrial recession in the developed countries, with its associated apparent abundancy of coal, oil and gas, to mask the longer-term energy situation. It is not here proposed to discuss the role of nuclear power in the energy scene except to say that, with the continuing need to develop energy resources, nuclear as a proven safe and economic system - will have a vital role to fulfil in meeting the world's future energy demands. This paper is concerned with the development of nuclear fuel and the industry which has grown around it during the last 30 years. It shall concentrate on its development in this country and describe the history and activities of BNFL.

  9. Studies of Fluctuation Processes in Nuclear Collisions

    Energy Technology Data Exchange (ETDEWEB)

    Ayik, Sakir [Tennessee Technological Univ., Cookeville, TN (United States). Dept. of Physics

    2016-04-14

    The standard one-body transport approaches have been extensively applied to investigate heavy-ion collision dynamics at low and intermediate energies. At low energies the approach is the mean-field description of the time-dependent Hartree-Fock (TDHF) theory. At intermediate energies the approach is extended by including a collision term, and its application has been carried out mostly in the semi-classical framework of the Boltzmann-Uhling-Uhlenbeck (BUU) model. The standard transport models provide a good understanding of the average properties of the collision dynamics in terms of the effective interactions in both low and intermediate energies. However, the standard models are inadequate for describing the fluctuation dynamics of collective motion at low energies and disassembling of the nuclear system into fragments at intermediate energies resulting from the growth of density fluctuations in the spinodal region. Our tasks have been to improve the standard transport approaches by incorporating fluctuation mechanisms into the description. There are mainly two different mechanisms for fluctuations: (i) Collisional fluctuations generated by binary nucleon collisions, which provide the dominant mechanism at intermediate energies, and (ii) One-body mechanism or mean-field fluctuations, which is the dominant mechanism at low energies. In the first part of our project, the PI extended the standard transport model at intermediate energies by incorporating collisional mechanism according to the “Generalized Langevin Description” of Mori formalism. The PI and his collaborators carried out a number of applications for describing dynamical mechanism of nuclear multi fragmentations, and nuclear collective response in the semi-classical framework of the approach, which is known as the Boltzmann-Langevin model. In the second part of the project, we considered dynamical description at low energies. Because of the effective Pauli blocking, the collisional dissipation and

  10. Innovative method for increased methane recovery from two-phase anaerobic digestion of food waste through reutilization of acidogenic off-gas in methanogenic reactor.

    Science.gov (United States)

    Yan, Bing Hua; Selvam, Ammaiyappan; Wong, Jonathan W C

    2016-10-01

    In this study, the performance of a two-phase anaerobic digestion reactor treating food waste with the reutilization of acidogenic off-gas was investigated with the objective to improve the hydrogen availability for the methanogenic reactor. As a comparison a treatment without off-gas reutilization was also set up. Results showed that acidogenic off-gas utilization in the upflow anaerobic sludge blanket (UASB) reactor increased the methane recovery up to 38.6%. In addition, a 27% increase in the production of cumulative chemical oxygen demand (COD) together with an improved soluble microbial products recovery dominated by butyrate was observed in the acidogenic leach bed reactor (LBR) with off-gas reutilization. Of the increased methane recovery, ∼8% was contributed by the utilization of acidogenic off-gas in UASB. Results indicated that utilization of acidogenic off-gas in methanogenic reactor is a viable technique for improving overall methane recovery.

  11. Laboratory optimization tests of technetium decontamination of Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-11-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable simplified operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  12. Resonance-like nuclear processes in solids: 3rd and 4th order processes

    CERN Document Server

    Kálmán, Péter

    2013-01-01

    It is recognized that in the family of heavy charged particle and electron assisted double nuclear processes resonance-like behavior can appear. The transition rates of the heavy particle assisted 3rd-order and electron assisted 4th-order resonance like double nuclear processes are determined. The power of low energy nuclear reactions in $Ni-H$ systems formed in $Ni$ placed in $H_{2}$ gas environment is treated. Nuclear power produced by quasi-resonant electron assisted double nuclear processes in these $Ni-H$ systems is calculated. The power obtained tallies with experiments and its magnitude is considerable for practical applications.

  13. Final Report on Testing of Off-Gas Treatment Technologies for Abatement of Atmospheric Emissions of Chlorinated Volatile Organic Compounds

    Energy Technology Data Exchange (ETDEWEB)

    Jarosch, T.R.; Haselow, J.S.; Rossabi, J.; Burdick, S.A.; Raymond, R.; Young, J.E.; Lombard, K.H.

    1995-01-23

    The purpose of this report is to summarize the results of the program for off-gas treatment of atmospheric emissions of chlorinated volatile organic compounds (CVOCs), in particular trichloroethylene (TCE) and perchloroethylene (PCE). This program was funded through the Department of Energy Office of Technology Development`s VOC`s in Non-Arid Soils Integrated Demonstration (VNID). The off-gas treatment program was initiated after testing of in-situ air stripping with horizontal wells was completed (Looney et al., 1991). That successful test expectedly produced atmospheric emissions of CVOCs that were unabated. It was decided after that test that an off-gas treatment is an integral portion of remediation of CVOC contamination in groundwater and soil but also because several technologies were being developed across the United States to mitigate CVOC emissions. A single platform for testing off-gas treatment technologies would facilitate cost effective evaluation of the emerging technologies. Another motivation for the program is that many CVOCs will be regulated under the Clean Air Act Amendments of 1990 and are already regulated by many state regulatory programs. Additionally, compounds such as TCE and PCE are pervasive subsurface environmental contaminants, and, as a result, a small improvement in terms of abatement efficiency or cost will significantly reduce CVOC discharges to the environment as well as costs to United States government and industry.

  14. Nuclear medium effects in Drell–Yan process

    Science.gov (United States)

    Haider, H.; Athar, M. Sajjad; Singh, S. K.; Ruiz Simo, I.

    2017-04-01

    We study the nuclear medium effects in Drell–Yan process using quark parton distribution functions calculated in a microscopic nuclear model which takes into account the effects of Fermi motion, nuclear binding and nucleon correlations through a relativistic nucleon spectral function. The contributions of π and ρ mesons as well as shadowing effects are also included. The beam energy loss is calculated using a phenomenological approach. The present theoretical results are compared with the experimental results of the E772 and E866 experiments. These results are applicable to the forthcoming experimental analysis of E906 Sea Quest experiment at the Fermi Lab.

  15. Nuclear medium effects in Drell-Yan process

    CERN Document Server

    Haider, H; Singh, S K; Simo, I Ruiz

    2016-01-01

    We study the nuclear medium effects in Drell-Yan process using quark parton distribution functions calculated in a microscopic nuclear model which takes into account the effects of Fermi motion, nuclear binding and nucleon correlations through a relativistic nucleon spectral function. The contributions of $\\pi$ and $\\rho$ mesons as well as shadowing effects are also included. The beam energy loss is calculated using a phenomenological approach. The present theoretical results are compared with the experimental results of E772 and E886 experiments. These results are applicable to the forthcoming experimental analysis of E906 Sea Quest experiment at Fermi Lab.

  16. The NJOY nuclear data processing system Version 91

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, R.E.; Muir, D.W.

    1994-10-01

    The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from elevated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code can work with neutrons, photons, and charged particles, and it can produce libraries for a wide variety of particle transport and reactor analysis codes.

  17. The NJOY Nuclear Data Processing System, Version 2016

    Energy Technology Data Exchange (ETDEWEB)

    Macfarlane, Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Muir, Douglas W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boicourt, R. M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kahler, III, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-09

    The NJOY Nuclear Data Processing System, version 2016, is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from evaluated nuclear data in the ENDF-4 through ENDF-6 legacy card-image formats. NJOY works with evaluated files for incident neutrons, photons, and charged particles, producing libraries for a wide variety of particle transport and reactor analysis codes.

  18. Impact of nuclear mass uncertainties on the $r$-process

    CERN Document Server

    Martin, Dirk; Nazarewicz, Witold; Olsen, Erik

    2015-01-01

    Nuclear masses play a fundamental role in understanding how the heaviest elements in the Universe are created in the $r$-process. We predict $r$-process nucleosynthesis yields using neutron capture and photodissociation rates that are based on nuclear density functional theory. Using six Skyrme energy density functionals based on different optimization protocols, we determine for the first time systematic uncertainty bands -- related to mass modeling -- for $r$-process abundances in realistic astrophysical scenarios. We find that features of the underlying microphysics make an imprint on abundances especially in the vicinity of neutron shell closures: abundance peaks and troughs are reflected in trends of neutron separation energy. Further advances in nuclear theory and experiments, when linked to observations, will help in the understanding of astrophysical conditions in extreme $r$-process sites.

  19. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  20. Nuclear structure studies for the astrophysical r-process

    CERN Document Server

    Pfeiffer, B; Thielemann, F K; Walters, W B

    2001-01-01

    The production of the heaviest elements in nature occurs via the r-process, i.e. a combination of rapid neutron captures, the inverse photodisintegrations, and slower beta sup - -decays, beta-delayed processes as well as fission and possibly interactions with intense neutrino fluxes. A correct understanding and modeling requires the knowledge of nuclear properties far from stability and a detailed prescription of the astrophysical environment. Experiments at radioactive ion beam facilities have played a pioneering role in exploring the characteristics of nuclear structure in terms of masses and beta-decay properties. Initial examinations paid attention to highly unstable nuclei with magic neutron numbers and their beta-decay properties, related to the location and height of r-process peaks, while recent activities focus on the evolution of shell effects at large distances from the valley of stability. We show in site-independent applications the effect of both types of nuclear properties on r-process abundanc...

  1. Nuclear medium effects in Drell-Yan process

    CERN Document Server

    Haider, H; Simo, I Ruiz; Singh, S K

    2013-01-01

    We study nuclear medium effects in Drell-Yan processes at small target x using quark parton distribution functions and nucleon structure functions for a bound nucleon calculated in a microscopic nuclear model which takes into account the effect of Fermi motion, nuclear binding and nucleon correlations through a relativistic spectral function. The contributions of $\\pi$ and $\\rho$ mesons, target mass corrections and nuclear shadowing are also included. The results are compared with the theoretical and experimental results. The model is able to successfully explain the low target x results of E772 and E866 Drell-Yan experiments and is applicable to the forthcoming experimental analysis of E906 Sea Quest experiment at Fermi Lab.

  2. Portable nuclear material detector and process

    Energy Technology Data Exchange (ETDEWEB)

    Hofstetter, Kenneth J (Aiken, SC); Fulghum, Charles K (Aiken, SC); Harpring, Lawrence J (North Augusta, SC); Huffman, Russell K (Augusta, GA); Varble, Donald L (Evans, GA)

    2008-04-01

    A portable, hand held, multi-sensor radiation detector is disclosed. The detection apparatus has a plurality of spaced sensor locations which are contained within a flexible housing. The detection apparatus, when suspended from an elevation, will readily assume a substantially straight, vertical orientation and may be used to monitor radiation levels from shipping containers. The flexible detection array can also assume a variety of other orientations to facilitate any unique container shapes or to conform to various physical requirements with respect to deployment of the detection array. The output of each sensor within the array is processed by at least one CPU which provides information in a usable form to a user interface. The user interface is used to provide the power requirements and operating instructions to the operational components within the detection array.

  3. Electrochemical fluorination for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  4. Galvanic cell for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  5. Modeling of off-gas emissions from wood pellets during marine transportation.

    Science.gov (United States)

    Pa, Ann; Bi, Xiaotao T

    2010-10-01

    After a fatal accident during the discharge of wood pellets at Helsingborg, emissions from pellets during marine transportation became a concern for the safe handling and storage of wood pellets. In this paper, a two-compartment model has been developed for the first time to predict the concentrations of CO, CO₂, CH₄, and O₂ inside the cargo ship and the time and rate of forced ventilation required before the safe entry into the stairway adjacent to the storage hatch. The hatch and stairway are treated as two perfectly mixed tanks. The gas exchange rate between these two rooms and the gas exchange rate with the atmosphere are fitted to satisfy a measured tracer final concentration of 33 p.p.m.v. in the stairway and an average final hatch to stairway CO, CO₂, and CH₄ concentration ratio of 1.62 based on measurement from five other hatch and stairway systems. The reaction kinetics obtained from a laboratory unit using a different batch of pellets, however, need to be scaled in order to bring the prediction to close agreement with onboard measured emission data at the end of voyage. Using the adjusted kinetic data, the model was able to predict the general trend of data recorded in the first 12.5 days of the voyage. Further validation, however, requires the data recorded over the whole journey. The model was applied to predict the effect of ocean temperature on the off-gas emissions and the buildup of concentrations in the hatch and stairway. For safe entry to the cargo ship, the current model predicted that a minimal ventilation rate of 4.4 hr⁻¹ is required for the stairway's CO concentration to lower to a safe concentration of 25 p.p.m.v. At 4.4 hr⁻¹, 10 min of ventilation time is required for the safe entry into the stairway studied.

  6. Maintenance of process instrumentation in nuclear power plants

    CERN Document Server

    Hashemian, H M

    2006-01-01

    Compiles 30 years of practical knowledge gained by the author and his staff in testing the I and C systems of nuclear power plants around the world. This book focuses on process temperature and pressure sensors and the verification of these sensors' calibration and response time.

  7. Identification of a Nuclear Exosome Decay Pathway for Processed Transcripts

    DEFF Research Database (Denmark)

    Meola, Nicola; Domanski, Michal; Karadoulama, Evdoxia

    2016-01-01

    , the Zn-finger protein ZCCHC8, and the RNA-binding factor RBM7. NEXT primarily targets early and unprocessed transcripts, which demands a rationale for how the nuclear exosome recognizes processed RNAs. Here, we describe the poly(A) tail exosome targeting (PAXT) connection, which comprises the ZFC3H1 Zn...

  8. The s Process: Nuclear Physics, Stellar Models, Observations

    CERN Document Server

    Kaeppeler, Franz; Bisterzo, Sara; Aoki, Wako

    2010-01-01

    Nucleosynthesis in the s process takes place in the He burning layers of low mass AGB stars and during the He and C burning phases of massive stars. The s process contributes about half of the element abundances between Cu and Bi in solar system material. Depending on stellar mass and metallicity the resulting s-abundance patterns exhibit characteristic features, which provide comprehensive information for our understanding of the stellar life cycle and for the chemical evolution of galaxies. The rapidly growing body of detailed abundance observations, in particular for AGB and post-AGB stars, for objects in binary systems, and for the very faint metal-poor population represents exciting challenges and constraints for stellar model calculations. Based on updated and improved nuclear physics data for the s-process reaction network, current models are aiming at ab initio solution for the stellar physics related to convection and mixing processes. Progress in the intimately related areas of observations, nuclear...

  9. Developing the nuclear idea: concept, technique, and process.

    Science.gov (United States)

    Billow, Richard M

    2013-10-01

    I introduce an approach to group that has remained undeveloped in the literature, but represents an essence of relationally oriented group psychotherapy. Evolving from the verbalizations and enactments through which the group symbolizes and becomes known-a nuclear idea takes shape. It emerges from the nucleus of the group process: co-created from intersubjective forces and locations that cannot be fully specified, yet may be possible to observe, name, and utilize clinically. Groups organize themselves by developing nuclear ideas, with the therapist's active participation. They are vehicles through which a group comes to think about its thinking: not only what it thinks, but also how it thinks, or chooses not to think, and when and why. Developing the nuclear idea provides a framework for how the therapist-and the group itself-goes about the task of containing. With its emphasis on meaning and the development of meaning as transformational, the concept of the nuclear idea supplements the whole group, interpersonal, and intrapsychic lenses through which the therapist comes to understand group experience and base interventions. Clinical vignettes illustrate how the therapist may develop nuclear ideas thematically, conceptualize further, and negotiate meaning with the co-participation of other group members.

  10. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  11. Towards A Unified HFE Process For The Nuclear Industry

    Energy Technology Data Exchange (ETDEWEB)

    Jacques Hugo

    2012-07-01

    As nuclear power utilities embark on projects to upgrade and modernize power plants, they are likely to discover that traditional engineering methods do not typically make provision for the integration of human considerations. In addition, human factors professionals will find that traditional human performance methods such as function allocation, task analysis, human reliability analysis and human-machine interface design do not scale well to the complexity of a large-scale nuclear power upgrade project. Up-to-date human factors engineering processes, methods, techniques and tools are required to perform these kinds of analyses. This need is recognized widely in industry and an important part of the Department of Energy’s Light Water Reactor Sustainability Program deals with identifying potential impacts of emerging technologies on human performance and the technical bases needed to address them. However, so far no formal initiative has been launched to deal with the lack of integrated processes. Although human factors integration frameworks do exist in industries such as aviation or defense, no formal integrated human factors process exists in the nuclear industry. As a first step towards creating such a process, a “unified human factors engineering process” is proposed as a framework within which engineering organizations, human factors practitioners and regulatory bodies can ensure that human factors requirements are embedded in engineering activities throughout the upgrade project life cycle.

  12. Chapter 2: selection process for nuclear sites; Capitulo 2: processo de selecao de sitios nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Atala, Drausio Lima

    2009-07-01

    The five criteria of a selection process will be discussed as follows: 1) Structure of the site selection procedure: functional stages required for discrimination from the interest region to installation license. 2) Selection criteria incorporating the nuclear regulatory, environmental and installation project requirements that should be considered in the selection process. 3) Quantification criterion: quantification of relative adequacy of a site related to selection criterion. 4) Weighing factor: weighing factor development reflecting the relative importance of individual criteria incorporating the interest composition among criterion. 5) Public involvement: development of a process integrating information and public enrollment participation.

  13. An operational approach to standard nuclear process model (SNPM) and SAP nuclear software implementation at Slovenske Elektrarne

    Energy Technology Data Exchange (ETDEWEB)

    Warren, C.C. [Nuclear Power Plants Operation Department, Slovenske Elektrarne, a.s., Mlynske nivy 47, 821 09 Bratislava (Slovakia)

    2010-07-01

    Benchmarking efforts in the fall of 2006 showed significant performance gaps in multiple measured processes between the Slovenske Elektrarne (SE) nuclear organization and the highest performing nuclear organizations in the world. While overall performance of the SE nuclear fleet was good and in the second quartile, when compared to the worldwide population of Pressurized Water Reactors (PWR), SE leadership set new goals to improve safety and operational performance to the first decile of the worldwide PWR Fleet. To meet these goals the SE nuclear team initiated a project to identify and implement the Best Practice nuclear processes in multiple areas. The benchmarking process identified the Standard Nuclear Performance Model (SNPM), used in the US nuclear fleet, as the industry best practice process model. The Slovenske Elektrarne nuclear management team used various change management techniques to clearly establish the case for organizational and process change within the nuclear organization. The project organization established by the SE nuclear management team relied heavily on functional line organization personnel to gain early acceptance of the project goals and methods thereby reducing organizational opposition to the significant organizational and process changes. The choice of a standardized process model used, all or in part, by approximately one third of the nuclear industry worldwide greatly facilitated the development and acceptance of the changes. Use of a nuclear proven templated software platform significantly reduced development and testing efforts for the resulting fully integrated solution. In the spring of 2007 SE set in motion a set of initiatives that has resulted in a significant redesign of most processes related to nuclear plant maintenance and continuous improvement. Significant organizational structure changes have been designed and implemented to align the organization to the SNPM processes and programs. The completion of the initial

  14. Potential industrial market for process heat from nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, R.W.

    1976-07-01

    A specific segment of industrial process heat use has been examined in detail to identify individual plant locations throughout the United states where nuclear generated steam may be a viable alternative. Five major industries have been studied: paper, chemicals, petroleum, rubber, and primary metals. For these industries, representing 75 percent of the total industrial steam consumption, the individual plant locations within the U.S. using steam in large quantities have been located and characterized as to fuel requirements.

  15. Analysis of suprathermal nuclear processes in the solar core plasma

    Science.gov (United States)

    Voronchev, Victor T.; Nakao, Yasuyuki; Watanabe, Yukinobu

    2017-04-01

    A consistent model for the description of suprathermal processes in the solar core plasma naturally triggered by fast particles generated in exoergic nuclear reactions is formulated. This model, based on the formalism of in-flight reaction probability, operates with different methods of treating particle slow-down in the plasma, and allows for the influence of electron degeneracy and electron screening on processes in the matter. The model is applied to examine slowing-down of 8.7 MeV α-particles produced in the {}7{Li}(p,α )α reaction of the pp chain, and to analyze suprathermal processes in the solar CNO cycle induced by them. Particular attention is paid to the suprathermal {}14{{N}}{(α ,{{p}})}17{{O}} reaction unappreciated in standard solar model simulations. It is found that an appreciable non-standard (α ,p) nuclear flow due to this reaction appears in the matter and modifies running of the CNO cycle in ∼95% of the solar core region. In this region at R> 0.1{R}ȯ , normal branching of nuclear flow {}14{{N}}≤ftarrow {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}} transforms to abnormal sequential flow {}14{{N}}\\to {}17{{O}}\\to {(}18{{F}})\\to {}18{{O}}, altering some element abundances. In particular, nuclear network calculations reveal that in the outer core the abundances of 17O and 18O isotopes can increase by a factor of 20 as compared with standard estimates. A conjecture is made that other CNO suprathermal (α ,p) reactions may also affect abundances of CNO elements, including those generating solar neutrinos.

  16. Applications of nuclear magnetic resonance imaging in process engineering

    Science.gov (United States)

    Gladden, Lynn F.; Alexander, Paul

    1996-03-01

    During the past decade, the application of nuclear magnetic resonance (NMR) imaging techniques to problems of relevance to the process industries has been identified. The particular strengths of NMR techniques are their ability to distinguish between different chemical species and to yield information simultaneously on the structure, concentration distribution and flow processes occurring within a given process unit. In this paper, examples of specific applications in the areas of materials and food processing, transport in reactors and two-phase flow are discussed. One specific study, that of the internal structure of a packed column, is considered in detail. This example is reported to illustrate the extent of new, quantitative information of generic importance to many processing operations that can be obtained using NMR imaging in combination with image analysis.

  17. Automated separation process for radioanalytical purposes at nuclear power plants.

    Science.gov (United States)

    Nagy, L G; Vajda, N; Vodicska, M; Zagyvai, P; Solymosi, J

    1987-10-01

    Chemical separation processes have been developed to remove the matrix components and thus to determine fission products, especially radioiodine nuclides, in the primary coolant of WWER-type nuclear reactors. Special procedures have been elaborated to enrich long-lived nuclides in waste waters to be released and to separate and enrich caesium isotopes in the environment. All processes are based mainly on ion-exchange separations using amorphous zirconium phosphate. Automated equipment was constructed to meet the demands of the plant personnel for serial analysis.

  18. Gap bridging enhancement of modified Urca process in nuclear matter

    CERN Document Server

    Alford, Mark G

    2016-01-01

    In nuclear matter at neutron-star densities and temperatures, Cooper pairing leads to the formation of a gap in the nucleon excitation spectra resulting in exponentially strong Boltzmann suppression of many transport coefficients. Previous calculations have shown evidence that density oscillations of sufficiently large amplitude can overcome this suppression for flavor-changing beta processes, via the mechanism of "gap bridging". We address the simplifications made in that initial work, and show that gap bridging can counteract Boltzmann suppression of neutrino emissivity for the realistic case of modified Urca processes in matter with $^3P_2$ neutron pairing.

  19. Electrosleeve process for in-situ nuclear steam generator repair

    Energy Technology Data Exchange (ETDEWEB)

    Barton, R.A. [Ontario Hydro Technologies, Toronto, ON (Canada); Moran, T.E. [Framatome Technologies Inc., Lynchburg, VA (United States); Renaud, E. [Babcock and Wilcox Industries Ltd., Cambridge, ON (Canada)

    1997-07-01

    Degradation of steam generator (SG) tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced out-ages, unit de-rating, SG replacement or even the permanent shutdown of a reactor. In response to the onset of SG tubing degradation at Ontario Hydro's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for SG tubing repair and the unique properties of the advanced sleeve material. The successful installation of Electrosleeves that have been in service for more than three years in Alloy 400 SG tubing at the Pickering-5 CANDU unit, the more recent extension of the technology to Alloy 600 and its demonstration in a U.S. pressurized water reactor (PWR), is presented. A number of PWR operators have requested plant operating technical specification changes to permit Electrosleeve SG tube repair. Licensing of the Electrosleeve by the U.S. Nuclear Regulatory Commission (NRC) is expected imminently. (author)

  20. SNL Sigma Off-Gas Team Contribution to the FY15 DOE/NE-MRWFD Campaign Accomplishments Report.

    Energy Technology Data Exchange (ETDEWEB)

    Nenoff, Tina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-21

    This program at Sandia is focused on Iodine waste form development for Fuel Cycle R&D needs. Our research has a general theme of “Capture and Storage of Iodine Fission Gas “ in which we are focused on silver loaded zeolite waste forms, evaluation of iodine loaded getter materials (eg., mordenite zeolite), and the development of low temperature glass waste forms that successfully incorporate iodine loaded getter materials from I2, organic iodide, etc. containing off-gas streams.

  1. Novel synthesis of bismuth-based adsorbents for the removal of {sup 129}I in off-gas

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Hwan, E-mail: yjh98@kaeri.re.kr [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Shin, Jin Myeong; Park, Jang Jin; Park, Geun Il [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Yim, Man Sung [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2015-02-15

    New adsorbents based on bismuth were investigated for the capture of iodine-129 ({sup 129}I) in off-gas produced from spent fuel reprocessing. Porous bulky materials were synthesized with polyvinyl alcohol (PVA) as a sacrificial template. Our findings showed that the iodine trapping capacity of as-synthesized samples could reach 1.9-fold that of commercial silver-exchanged zeolite (AgX). The thermodynamic stability of the reaction products explains the high removal efficiency of iodine. We also found that the pore volume of each sample was closely related to the ratio of the reaction products.

  2. DEVELOPMENT OF AN ANTIFOAM TRACKING SYSTEM AS AN OPTION TO SUPPORT THE MELTER OFF-GAS FLAMMABILITY CONTROL STRATEGY AT THE DWPF

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T.; Lambert, D.

    2014-08-27

    The Savannah River National Laboratory (SRNL) has been working with the Savannah River Remediation (SRR) Defense Waste Processing Facility (DWPF) in the development and implementation of an additional strategy for confidently satisfying the flammability controls for DWPF’s melter operation. An initial strategy for implementing the operational constraints associated with flammability control in DWPF was based upon an analytically determined carbon concentration from antifoam. Due to the conservative error structure associated with the analytical approach, its implementation has significantly reduced the operating window for processing and has led to recurrent Slurry Mix Evaporator (SME) and Melter Feed Tank (MFT) remediation. To address the adverse operating impact of the current implementation strategy, SRR issued a Technical Task Request (TTR) to SRNL requesting the development and documentation of an alternate strategy for evaluating the carbon contribution from antifoam. The proposed strategy presented in this report was developed under the guidance of a Task Technical and Quality Assurance Plan (TTQAP) and involves calculating the carbon concentration from antifoam based upon the actual mass of antifoam added to the process assuming 100% retention. The mass of antifoam in the Additive Mix Feed Tank (AMFT), in the Sludge Receipt and Adjustment Tank (SRAT), and in the SME is tracked by mass balance as part of this strategy. As these quantities are monitored, the random and bias uncertainties affecting their values are also maintained and accounted for. This report documents: 1) the development of an alternate implementation strategy and associated equations describing the carbon concentration from antifoam in each SME batch derived from the actual amount of antifoam introduced into the AMFT, SRAT, and SME during the processing of the batch. 2) the equations and error structure for incorporating the proposed strategy into melter off-gas flammability assessments

  3. Melt processed multiphase ceramic waste forms for nuclear waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, Jake, E-mail: jake.amoroso@srs.gov [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James C. [Savannah River National Laboratory, Aiken, SC 29808 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lin, Ye; Chen, Fanglin [University of South Carolina, Columbia, SC 29208 (United States); Su, Dong [Brookhaven National Laboratory, Upton, NY 11973 (United States); Brinkman, Kyle S. [Clemson University, Clemson, SC 29634 (United States)

    2014-11-15

    Highlights: • We explored the feasibility of melt processing multiphase titanate-based ceramics. • Melt processing produced phases obtained by alternative processing methods. • Phases incorporated multiple lanthanides and transition metals. • Processing in reducing atmosphere suppressed un-desirable Cs–Mo coupling. • Cr partitions to and stabilizes the hollandite phase, which promotes Cs retention. - Abstract: Ceramic waste forms are promising hosts for nuclear waste immobilization as they have the potential for increased durability and waste loading compared with conventional borosilicate glass waste forms. Ceramics are generally processed using hot pressing, spark plasma sintering, and conventional solid-state reaction, however such methods can be prohibitively expensive or impractical at production scales. Recently, melt processing has been investigated as an alternative to solid-state sintering methods. Given that melter technology is currently in use for High Level Waste (HLW) vitrification in several countries, the technology readiness of melt processing appears to be advantageous over sintering methods. This work reports the development of candidate multi-phase ceramic compositions processed from a melt. Cr additions, developed to promote the formation and stability of a Cs containing hollandite phase were successfully incorporated into melt processed multi-phase ceramics. Control of the reduction–oxidation (Redox) conditions suppressed undesirable Cs–Mo containing phases, and additions of Al and Fe reduced the melting temperature.

  4. Diffusion processes in tumors: A nuclear medicine approach

    Science.gov (United States)

    Amaya, Helman

    2016-07-01

    The number of counts used in nuclear medicine imaging techniques, only provides physical information about the desintegration of the nucleus present in the the radiotracer molecules that were uptaken in a particular anatomical region, but that information is not a real metabolic information. For this reason a mathematical method was used to find a correlation between number of counts and 18F-FDG mass concentration. This correlation allows a better interpretation of the results obtained in the study of diffusive processes in an agar phantom, and based on it, an image from the PETCETIX DICOM sample image set from OsiriX-viewer software was processed. PET-CT gradient magnitude and Laplacian images could show direct information on diffusive processes for radiopharmaceuticals that enter into the cells by simple diffusion. In the case of the radiopharmaceutical 18F-FDG is necessary to include pharmacokinetic models, to make a correct interpretation of the gradient magnitude and Laplacian of counts images.

  5. Scientific Opportunities to Reduce Risk in Nuclear Process Science

    Energy Technology Data Exchange (ETDEWEB)

    Bredt, Paul R.; Felmy, Andrew R.; Gauglitz, Phillip A.; Hobbs, David T.; Krahn, Steve; Machara, N.; Mcilwain, Michael; Moyer, Bruce A.; Poloski, Adam P.; Subramanian, K.; Vienna, John D.; Wilmarth, B.

    2008-07-18

    Cleaning up the nation’s nuclear weapons complex remains as one of the most technologically challenging and financially costly problems facing the U.S. Department of Energy (DOE). Safety, cost, and technological challenges have often delayed progress in retrieval, processing, and final disposition of high-level waste, spent nuclear fuel, and challenging materials. Some of the issues result from the difficulty and complexity of the technological issues; others have programmatic bases, such as contracting strategies that may provide undue focus on near-term, specific clean-up goals or difficulty in developing and maintaining stakeholder confidence in the proposed solutions. We propose that independent basic fundamental science research focused on the full cleanup life-cycle offers an opportunity to help address these challenges by providing 1) scientific insight into the fundamental mechanisms involved in currently selected processing and disposal options, 2) a rational path to the development of alternative technologies should the primary options fail, 3) confidence that models that predict long-term performance of different disposal options are based upon the best available science, 4) fundamental science discovery that enables transformational solutions to revolutionize the current baseline processes.

  6. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D.K.; Hendrickson, H.R. (comps.)

    1978-10-01

    These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

  7. Nuclear processes and neutrino production in solar flares

    Science.gov (United States)

    Lingenfelter, R. E.; Ramaty, R.; Murphy, R. J.; Kozlovsky, S.

    1985-01-01

    The determination of flare neutrino flux is approached from the standpoint of recent observations and theoretical results on the nuclear processes in solar flares. Attention is given to the energy spectra and total numbers of accelerated particles in flares, as well as their resulting production of beta(+)-emitting radionuclei and pions; these should be the primary sources of neutrinos. The observed 0.511 MeV line flux for the June 21, 1980 flare is compared with the expected from the number and spectrum of accelerated particles.

  8. Dead-ended anode polymer electrolyte fuel cell stack operation investigated using electrochemical impedance spectroscopy, off-gas analysis and thermal imaging

    Science.gov (United States)

    Meyer, Quentin; Ashton, Sean; Curnick, Oliver; Reisch, Tobias; Adcock, Paul; Ronaszegi, Krisztian; Robinson, James B.; Brett, Daniel J. L.

    2014-05-01

    Dead-ended anode operation, with intermittent purge, is increasingly being used in polymer electrolyte fuel cells as it simplifies the mass flow control of feed and improves fuel efficiency. However, performance is affected through a reduction in voltage during dead-ended operation, particularly at high current density. This study uses electrochemical impedance spectroscopy (EIS), off-gas analysis and high resolution thermal imaging to examine the source of performance decay during dead-ended operation. A novel, 'reconstructed impedance' technique is applied to acquire complete EIS spectra with a temporal resolution that allows the dynamics of cell processes to be studied. The results provide evidence that upon entering dead-ended operation, there is an initial increase in performance associated with an increase in anode compartment pressure and improved hydration of the membrane electrolyte. Subsequent reduction in performance is associated with an increase in mass transport losses due to a combination of water management issues and build-up of N2 in the anode. The purge process rapidly recovers performance. Understanding of the processes involved in the dead-end/purge cycle provides a rationale for determining the optimum cycle frequency and duration as a function of current density.

  9. Radioactive decay as a forced nuclear chemical process: Phenomenology

    Science.gov (United States)

    Timashev, S. F.

    2015-11-01

    Concepts regarding the mechanism of radioactive decay of nuclei are developed on the basis of a hypothesis that there is a dynamic relationship between the electronic and nuclear subsystems of an atom, and that fluctuating initiating effects of the electronic subsystem on a nucleus are possible. Such relationship is reflected in experimental findings that show the radioactive decay of nuclei might be determined by a positive difference between the mass of an initial nucleus and the mass of an atom's electronic subsystem, i.e., the mass of the entire atom (rather than that of its nucleus) and the total mass of the decay products. It is established that an intermediate nucleus whose charge is lower by unity than the charge of the initial radioactive nucleus is formed as a result of the above fluctuating stimuli that initiate radioactive decay, and its nuclear matter is thus in an unbalanced metastable state of inner shakeup, affecting the quark subsystem of nucleons. The intermediate nucleus thus experiences radioactive decay with the emission of α or β particles. At the same time, the high energy (with respect to the chemical scale) of electrons in plasma served as a factor initiating the processes in different nuclear chemical transformations and radioactive decays in low-temperature plasma studied earlier, particularly during the laser ablation of metals in aqueous solutions of different compositions and in near-surface cathode layers upon glow discharge. It is shown that a wide variety of nucleosynthesis processes in the Universe can be understood on the same basis, and a great many questions regarding the formation of light elements in the solar atmosphere and some heavy elements (particularly p-nuclei) in the interiors of massive stars at late stages of their evolution can also be resolved.

  10. Oxidation and Condensation of Zinc Fume From Zn-CO2-CO-H2O Streams Relevant to Steelmaking Off-Gas Systems

    Science.gov (United States)

    Bronson, Tyler M.; Ma, Naiyang; Zhu, Liang Zhu; Sohn, Hong Yong

    2017-04-01

    The objective of this research was to study the condensation of zinc vapor to metallic zinc and zinc oxide solid under varying environments to investigate the feasibility of in-process separation of zinc from steelmaking off-gas dusts. Water vapor content, temperature, degree of cooling, gas composition, and initial zinc partial pressure were varied to simulate the possible conditions that can occur within steelmaking off-gas systems, limited to Zn-CO2-CO-H2O gas compositions. The temperature of deposition and the effect of rapidly quenching the gas were specifically studied. A homogeneous nucleation model for applicable experiments was applied to the analysis of the experimental data. It was determined that under the experimental conditions, oxidation of zinc vapor by H2O or CO2 does not occur above 1108 K (835 °C) even for highly oxidizing streams (CO2/CO = 40/7). Rate expressions that correlate CO2 and H2O oxidation rates to gas composition, partial pressure of water vapor, temperature, and zinc partial pressure were determined to be as follows: Rate( mol/m2 s ) = 406 \\exp ( - 50.2 kJ/mol/RT )( p_Zn p_{CO2 - p_CO /K_{eq,CO2 ) mol/m2 × s Rate( mol/m2 s ) = 32.9 \\exp ( - 13.7 kJ/mol/RT )( p_Zn p_{H2 O - p_{H2 /K_{eq,H2 O ) mol/m2 × s It was proven that a rapid cooling rate (500 K/s) significantly increases the ratio of metallic zinc to zinc oxide as opposed to a slow cooling rate (250 K/s). SEM analysis found evidence of heterogeneous growth of ZnO as well as of homogeneous formation of metallic zinc. The homogeneous nucleation model fit well with experiments where only metallic zinc deposited. An expanded model with rates of oxidation by CO2 and H2O as shown was combined with the homogenous nucleation model and then compared with experimental data. The calculated results based on the model gave a reasonable fit to the measured data. For the conditions used in this study, the rate equations for the oxidation of zinc by carbon dioxide and water vapor as well

  11. Nuclear fuel materials processing in reactive gas plasma

    Energy Technology Data Exchange (ETDEWEB)

    Min, Jin Young; Yang, Myung Seung; Seo, Yong Dae; Kim Yong Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and have been under post irradiation examination. (Hong, J. S.)

  12. DUPIC nuclear fuel manufacturing and process technology development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Sung Paal; Lee, Jung Won; Kim, Jong Ho; Kim, Soo Sung; Kim, Woong Ki; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and has been under post irradiation examination. (Hong, J. S.)

  13. Nuclear pre-mRNA processing in plants

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, A.S.N. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Biology and Program in Molecular Plant Biology; Golovkin, M. (eds.) [Thomas Jefferson Univ., Philadelphia, PA (United States). Dept. of Microbiology

    2008-07-01

    This volume of CTMI, entitled Nuclear premRNA Processing in Plants, with 16 chapters from leading scientists in this area, summarizes recent advances in nuclear pre-mRNA processing and its role in plant growth and development. It provides researchers in the field, as well as those in related areas, with an up-to-date and comprehensive, yet concise, overview of the current status and future potential of this research in understanding plant biology. The first four chapters focus on spliceosome composition, genome-wide alternative splicing, and splice site requirements for U1 and U12 introns using computational and empirical approaches. Analysis of sequenced plant genomes has revealed that 80% of all protein-coding nuclear genes contain one or more introns. The lack of an in vitro plant splicing system has made it difficult to identify general and plant-specific components of splicing machinery in plants. The next three chapters focus on serine/arginine-rich (SR) proteins, a family of highly conserved proteins, which are known to play key roles in constitutive and regulated splicing of pre-mRNA and other aspects of RNA metabolism in metazoans. These proteins engage both in RNA binding and protein.protein interactions and function as splicing regulators at multiple stages of spliceosome assembly. This family of proteins has expanded considerably in plants with several plant-specific SR proteins. Several serendipitous discoveries made using forward genetics are indicating that RNA metabolism (alternative splicing, alternative polyadenylation, mRNA transport) plays an important role in many aspects of plant growth and development and in plant responses to biotic and abiotic stresses. The next seven chapters focus on these aspects of RNA metabolism. The plant hormone abscisic acid (ABA) regulates a number of physiological processes during plant growth and development. The next chapter or A.B. Rose discusses the ways introns affect gene expression both positively and

  14. Scientific Opportunities to Reduce Risk in Nuclear Process Science

    Energy Technology Data Exchange (ETDEWEB)

    Bredt, Paul R.; Felmy, Andrew R.; Gauglitz, Phillip A.; Poloski, Adam P.; Vienna, John D.; Moyer, Bruce A.; Hobbs, David; Wilmarth, B.; Mcilwain, Michael; Subramanian, K.; Krahn, Steve; Machara, N.

    2009-08-28

    Cleaning up the nation’s nuclear weapons complex remains as one of the most technologically challenging and financially costly problems facing the U.S. Department of Energy (DOE). Safety, cost, and technological challenges have often delayed progress in retrieval, processing, and final disposition of high-level waste, spent nuclear fuel, and challenging materials. Some of the issues result from the difficulty and complexity of the technological issues; others have programmatic bases, such as strategies that may provide undue focus on near-term goals or difficulty in developing and maintaining stakeholder confidence in the proposed solutions. We propose that independent basic fundamental science research, addressing the full cleanup life-cycle, offers an opportunity to help address these challenges by providing 1) scientific insight into the fundamental mechanisms involved in currently selected processing and disposal options, 2) a rational path to the development of alternative technologies should the primary options fail, 3) confidence that models that predict long-term performance of different disposal options are based upon the best available science, and 4) fundamental science discovery that enables transformational solutions to revolutionize the current baseline processes. Over the last 3 years, DOE’s Office of Environmental Management (EM) has experienced a fundamental shift in philosophy. The mission focus of driving to closure has been replaced by one of enabling the long-term needs of DOE and the nation. Resolving new challenges, such as the disposition of DOE spent nuclear fuel, have been added to EM’s responsibilities. In addition, the schedules for addressing several elements of the cleanup mission have been extended. As a result, EM’s mission is no longer focused only on driving the current baselines to closure. Meeting the mission will require fundamental advances over at least a 30-year window if not longer as new challenges are added. The

  15. Melt processed multiphase ceramic waste forms for nuclear waste immobilization

    Science.gov (United States)

    Amoroso, Jake; Marra, James C.; Tang, Ming; Lin, Ye; Chen, Fanglin; Su, Dong; Brinkman, Kyle S.

    2014-11-01

    Ceramic waste forms are promising hosts for nuclear waste immobilization as they have the potential for increased durability and waste loading compared with conventional borosilicate glass waste forms. Ceramics are generally processed using hot pressing, spark plasma sintering, and conventional solid-state reaction, however such methods can be prohibitively expensive or impractical at production scales. Recently, melt processing has been investigated as an alternative to solid-state sintering methods. Given that melter technology is currently in use for High Level Waste (HLW) vitrification in several countries, the technology readiness of melt processing appears to be advantageous over sintering methods. This work reports the development of candidate multi-phase ceramic compositions processed from a melt. Cr additions, developed to promote the formation and stability of a Cs containing hollandite phase were successfully incorporated into melt processed multi-phase ceramics. Control of the reduction-oxidation (Redox) conditions suppressed undesirable Cs-Mo containing phases, and additions of Al and Fe reduced the melting temperature.

  16. Used nuclear fuel separations process simulation and testing

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  17. Experimental Optimization and Modeling of Sodium Sulfide Production from H2S-Rich Off-Gas via Response Surface Methodology and Artificial Neural Network

    Directory of Open Access Journals (Sweden)

    Bashipour Fatemeh

    2017-03-01

    Full Text Available The existence of hydrogen sulfide (H2S in the gas effluents of oil, gas and petrochemical industries causes environmental pollution and equipment corrosion. These gas streams, called off-gas, have high H2S concentration, which can be used to produce sodium sulfide (Na2S by H2S reactive absorption. Na2S has a wide variety of applications in chemical industries. In this study, the reactive absorption process was performed using a spray column. Response Surface Methodology (RSM was applied to design and optimize experiments based on Central Composite Design (CCD. The individual and interactive effects of three independent operating conditions on the weight percent of the produced Na2S (Y were investigated by RSM: initial NaOH concentration (10-20% w/w, scrubbing solution temperature (40-60 °C and liquid-to-gas volumetric ratio (15 × 10−3 to 25 × 10−3. Furthermore, an Artificial Neural Network (ANN model was used to predict Y. The results from RSM and ANN models were compared with experimental data by the regression analysis method. The optimum operating conditions specified by RSM resulted in Y of 15.5% at initial NaOH concentration of 19.3% w/w, scrubbing solution temperature of 40 °C and liquid-to-gas volumetric ratio of 24.6 × 10−3 v/v.

  18. Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, John M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-29

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  19. Thermodynamic considerations on a boil-off gas recovery system; Considerations thermo-dynamiques sur le systeme de recuperation du gaz d'evaporation (boil-off)

    Energy Technology Data Exchange (ETDEWEB)

    Lanzi, D.; Vareschi, G. [SNAM S.p.A. (Italy); Sguera, O. [Snamprogetti S.p.A. (Italy)

    2000-07-01

    The Panigaglia LNG re-gasification terminal is equipped with a system recovering the boil-off gas produced during both the unloading and the re-gasification phases. The system consists of a compression station and an absorption column that condenses the boil-off by means of two packing bodies, made of INTALOX METAL ring-type, DN 1'' or 2'', depending on the quantity of boil-off available. The above configuration, which was built during the 1991 terminal revamping, was designed to optimise the previous plant facilities and compression works by compressing the boil-off up to 30 bar. At present, to reduce the operating costs deriving from the electric power consumption due to compression of the gas to high pressures, the possibility of running the column at a pressure less than 25 bar has been evaluated. A new processing solution has been studied. This configuration permits using the LNG coming from the low-pressure pumps and therefore an LNG temperature of 15 deg. C less than the one used in the original system. The maximum boil-off rate that can be absorbed has been calculated at different operating pressures. The thermodynamic (material and energy balance) behaviour of the column has been checked and the thermodynamic studies and statistical analysis have given satisfactory and consistent results. In parallel, the operating parameters referred to a period of about one year, during both the unloading and the operating phases, have been identified and analysed. (authors)

  20. Photoacoustic Spectroscopy for the Quantification of N2O in the Off-Gas of Wastewater Treatment Plants.

    Science.gov (United States)

    Thaler, Klemens M; Berger, Christoph; Leix, Carmen; Drewes, Jörg; Niessner, Reinhard; Haisch, Christoph

    2017-03-21

    Different configurations of photoacoustic (PA) setups for the online-measurement of gaseous N2O, employing semiconductor lasers at 2.9 and 4.5 μm, were developed and tested. Their performance was assessed with respect to the analysis of N2O emissions from wastewater treatment plants. For this purpose, the local N2O emissions of a wastewater treatment bioreactor was sampled by a dedicated mobile sampling device, and the total N2O emissions were analyzed in the gastight headspace of the bioreactor. We found that the use of a quantum-cascade laser emitting at about 4.53 μm, operated in a wavelength modulation mode, in combination with a conventional longitudinal PA cell yielded the highest sensitivity (<100 ppbv). However, we also observed a strong cross-sensitivity to humidity, which can be explained by increased V-T relaxation. This observation in combination with the limited dynamic range (max conc. ∼ 3000 ppmv) led us to the use of the less-sensitive but spectroscopically more robust 2.9 μm laser. A detection limit below 1 ppmv, a dynamic range of more than 4 orders of magnitude, no influence of humidity or any other substance relevant to the off-gas analysis, as well as a comparable low price of the laser source made it the ideal tool for N2O analyses of the off-gas of a wastewater treatment plant. Such a system was implemented successfully in a full-scale wastewater treatment plant. The results regarding the comparison of different PA setups can be transferred to other systems, and the optimum performance can be selected according to the specific demands.

  1. Applications of digital pulse processing in nuclear spectroscopy

    CERN Document Server

    Grzywacz, R

    2003-01-01

    Data acquisition systems for nuclear spectroscopy have traditionally been based on hybrid systems with analog shaping amplifiers followed by analog-to-digital converters. Recently, however, new systems based on digital signal processing concepts have been developed. For example, one specific design, the Digital Gamma Finder (DGF-4C), has been used extensively for particle- and gamma-spectroscopy of nuclei far from stability. Using the DGF-4C, a variety of data acquisition systems have been implemented and used for measurements with semiconductor and scintillator detectors at recoil separators like the RMS at ORNL, the FRS at GSI and LISE at GANIL. Some novel features and unique advantages, such as trigger-less operation and pulse shape recording, are discussed in the context of selected studies.

  2. Scientific Opportunities to Reduce Risk in Nuclear Process Science - 9279

    Energy Technology Data Exchange (ETDEWEB)

    Bredt, Paul R.; Felmy, Andrew R.; Gauglitz, Phillip A.; Poloski, Adam P.; Vienna, John D.; Moyer, Bruce A.; Hobbs, David; Wilmarth, B.; Mcilwain, Michael; Subramanian, K.; Krahn, Steve; Machara, N.

    2009-03-01

    In this document, we propose that scientific investments for the disposal of nuclear and hazardous wastes should not be focused solely on what may be viewed as current Department of Energy needs, but also upon longer-term investments in specific areas of science that underpin technologies presently in use. In the latter regard, we propose four science theme areas: 1) the structure and dynamics of materials and interfaces, 2) coupled chemical and physical processes, 3) complex solution phase phenomena, and 4) chemical recognition phenomena. The proposed scientific focus for each of these theme areas and the scientific opportunities are identified, along with links to major risks within the initiative areas identified in EM’s Engineering and Technology Roadmap.

  3. Management of the process of nuclear transport; Gestion del proceso de transporte nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Requejo, P.

    2015-07-01

    Since 1996 ETSA is the only Spanish logistics operator specialized on servicing the nuclear and radioactive industry. Nowadays ETSA has some technological systems specifically designed for the management of nuclear transports. These tools have been the result of the analysis of multiple factors involved in nuclear shipments, of ETSAs wide experience as a logistics operator and the search for continuous improvement. (Author)

  4. A survey of decontamination processes applicable to DOE nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chen, L.; Chamberlain, D.B.; Conner, C.; Vandegrift, G.F.

    1997-05-01

    The objective of this survey was to select an appropriate technology for in situ decontamination of equipment interiors as part of the decommissioning of U.S. Department of Energy nuclear facilities. This selection depends on knowledge of existing chemical decontamination methods. This report provides an up-to-date review of chemical decontamination methods. According to available information, aqueous systems are probably the most universally used method for decontaminating and cleaning metal surfaces. We have subdivided the technologies, on the basis of the types of chemical solvents, into acid, alkaline permanganate, highly oxidizing, peroxide, and miscellaneous systems. Two miscellaneous chemical decontamination methods (electrochemical processes and foam and gel systems) are also described. A concise technical description of various processes is given, and the report also outlines technical considerations in the choice of technologies, including decontamination effectiveness, waste handing, fields of application, and the advantages and limitations in application. On the basis of this survey, six processes were identified for further evaluation. 144 refs., 2 tabs.

  5. Diffusion processes in tumors: A nuclear medicine approach

    Energy Technology Data Exchange (ETDEWEB)

    Amaya, Helman, E-mail: haamayae@unal.edu.co [Grupo de Física Nuclear, Universidad Nacional de Colombia (Colombia)

    2016-07-07

    The number of counts used in nuclear medicine imaging techniques, only provides physical information about the desintegration of the nucleus present in the the radiotracer molecules that were uptaken in a particular anatomical region, but that information is not a real metabolic information. For this reason a mathematical method was used to find a correlation between number of counts and {sup 18}F-FDG mass concentration. This correlation allows a better interpretation of the results obtained in the study of diffusive processes in an agar phantom, and based on it, an image from the PETCETIX DICOM sample image set from OsiriX-viewer software was processed. PET-CT gradient magnitude and Laplacian images could show direct information on diffusive processes for radiopharmaceuticals that enter into the cells by simple diffusion. In the case of the radiopharmaceutical {sup 18}F-FDG is necessary to include pharmacokinetic models, to make a correct interpretation of the gradient magnitude and Laplacian of counts images.

  6. A Hydrogen Containment Process For Nuclear Thermal Engine Ground Testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    A hydrogen containment process was proposed for ground testing of a nuclear thermal engine. The hydrogen exhaust from the engine is contained in two unit operations: an oxygen-rich burner and a tubular heat exchanger. The burner burns off the majority of the hydrogen, and the remaining hydrogen is removed in the tubular heat exchanger through the species recombination mechanism. A multi-dimensional, pressure-based multiphase computational fluid dynamics methodology was used to conceptually sizing the oxygen-rich burner, while a one-dimensional thermal analysis methodology was used to conceptually sizing the heat exchanger. Subsequently, a steady-state operation of the entire hydrogen containment process, from pressure vessel, through nozzle, diffuser, burner and heat exchanger, was simulated numerically, with the afore-mentioned computational fluid dynamics methodology. The computational results show that 99% of hydrogen reduction is achieved at the end of the burner, and the rest of the hydrogen is removed to a trivial level in the heat exchanger. The computed flammability at the exit of the heat exchanger is less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process.

  7. Laboratory Optimization Tests of Technetium Decontamination of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Melter Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-23

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  8. Laboratory Optimization Tests of Decontamination of Cs, Sr, and Actinides from Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-01-06

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also substantially decrease the LAW vitrification mission duration and quantity of glass waste.

  9. Application of laser processing for disassembly of nuclear power plants

    Science.gov (United States)

    Baranov, Gennady A.; Zinchenko, A. V.; Arutyunyan, R. B.

    1998-12-01

    Provision of safety and drop of ecological risk at salvaging of nuclear submarines (NSM) of Russia Navy Forces represents one of the most actual problems of nowadays. It is necessary to remove from services of Russian Navy Forces 170 - 180 nuclear submarines by 2000. At salvaging of Russian Navy Forces NSM it should be necessary to cut out reactor compartments with more than 150 thousand tons of gross weight and to fragment terminal carcasses of submarines with gross weight of 2 million tons. Taking into account overall dimensions of salvaging objects and Euro-standard requirement on the sizes of carcass fragments, for salvaging of one NSM it is necessary to execute more than 10 km of cuts. Using of conventional methods of gas and plasma cutting of ship constructions and equipment polluted with radioactive oxides and bedding of insulation and paint and varnish materials causes contamination of working zones and environment by a mix of radioactive substances and highly toxic combustion products, nomenclature of which includes up to 50 names. Calculations carried out in the Institute of industrial and Marine Medicine have shown that salvage of just one NSM with using of gas and plasma cutting are accompanied by discharge into an environment of up to 11.5 kg of chromium oxides, up to 22.5 kg of manganese oxides, up to 97 kg of carbon oxides and up to 650 kg of nitrogen oxides. Fragmentation of such equipment by a method of directional explosion or hydraulic jet is problematic because of complexity of treated constructions and necessity to create special protective facilities, which will accumulate a bulk of radioactive and toxic discharges, as a consequence of the explosion and spreaded by shock waves and water deluges. In a number of new technological processes the cutting with using of high-power industrial lasers radiation stands out. As compared with other technological processes, laser cutting has many advantages determined by such unique properties of laser

  10. Nuclear structure theory for the astrophysical rp-process and r-process

    Energy Technology Data Exchange (ETDEWEB)

    Brown, B.A.; Clement, R.; Schatz, H.; Giansiracusa, J.; Richter, W.A.; Hjorth-Jensen, M.; Kratz, K.-L.; Pfeiffer, B.; Walters, W.B

    2003-05-19

    The astrophysical processes of rapid-proton capture and rapid-neutron capture require the knowledge of many nuclear properties which are not known from experiment. I will describe two examples of how theoretical models are used to provide this input. The first of these uses the Hartree-Fock method for displacement energies to obtain the masses of proton-rich nuclei needed for the rp-process. The second uses a model for configuration mixing near {sup 132}Sn to provide Q values and beta-decay lifetimes for the r-process.

  11. Development of the preparation technology of macroporous sorbent for industrial off-gas treatment including {sup 14}C

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Il Hoon; Cho, Young Hyun; Park, Guen Il; Kim, In Tae; Kim, June Hyung; Ahn, Byung Kil

    2001-01-01

    For environmental and health effects due to increasing levels of pollution in the atmosphere, it is necessary to develop environmentally sound technologies for the treatment of greenhouse gases (CO{sub 2}, CH{sub 4}, CFC, etc.) and acid gases (SOx, NOx, etc.). Specifically, advanced technology for CO{sub 2} capturing is currently one of the most important environmental issues in worldwide. {sup 14}CO{sub 2}, specially which has been gradually emerging issue in the nuclear facilities, is generated about 330 ppm from the CANDU (Canadian Deuterium Uranium Reactor) nuclear power plant and the DUPIC (Direct Use of spent PWR fuel in CANDU reactors) process which is the process of spent fuel treatment. For this purpose, it is necessary to develop the most efficient treatment technology of CO{sub 2} capture by various lime materials in semi- or dry process, it should be also considering a removal performance, waste recycling and safety of disposal. In order to develop a highly active slaked lime as a sorbent for CO{sub 2} and high temperature desulfurization, macroporous slaked lime is necessarily prepared by modified swelling process and equipment, which was developed under carrying out this project. And also for the optimal removal process of off-gases the removal performance tests of various sorbents and the effects of relative humidity and bed depth on the removal capacity must be considered.

  12. Low-Level waste phase 1 melter testing off gas and mass balance evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N.

    1996-06-28

    Commercially available melter technologies were tested during 1994-95 as part of a multiphase program to test candidate technologies for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of Hanford Site tank wastes. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes were also tested. Various feed material samples, product glass samples, and process offgas streams were characterized to provide data for evaluation of process decontamination factors and material mass balances for each vitrification technology. This report describes the melter mass balance evaluations and results for six of the Phase 1 LLW melter vendor demonstration tests.

  13. Two citizen task forces and the challenge of the evolving nuclear waste siting process

    Energy Technology Data Exchange (ETDEWEB)

    Peelle, E.B.

    1990-01-01

    Siting any nuclear waste facility is problematic in today's climate of distrust toward nuclear agencies and fear of nuclear waste. This study compares and contrasts the siting and public participation processes as two citizen task forces dealt with their difficult responsibilities. 10 refs., 3 tabs.

  14. 75 FR 63725 - Nuclear Energy Institute; Consideration of Petition in the Rulemaking Process

    Science.gov (United States)

    2010-10-18

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION 10 CFR Part 70 Nuclear Energy Institute; Consideration of Petition in the Rulemaking Process... raised in a petition submitted by the Nuclear Energy Institute (NEI), and is denying the remaining four...

  15. Establishment of the nuclear regulatory framework for the process of decommissioning of nuclear installations in Mexico; Establecimiento del marco regulador nuclear para el proceso de cierre de instalaciones nucleares en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Salmeron V, J. A.; Camargo C, R.; Nunez C, A., E-mail: juan.salmeron@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    Today has not managed any process of decommissioning of nuclear installations in the country; however because of the importance of the subject and the actions to be taken to long term, the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) in Mexico, accordance with its objectives is developing a National Nuclear Regulatory Framework and defined requirements to ensure the implementation of appropriate safety standards when such activities are performed. In this regard, the national nuclear regulatory framework for nuclear installations and the particular case of nuclear power reactors is presented, as well as a proposed licensing process for the nuclear power plant of Laguna Verde based on international regulations and origin country regulations of the existing reactors in nuclear facilities in accordance with the license conditions of operation to allow to define and incorporate such regulation. (Author)

  16. Nuclear energy an introduction to the concepts, systems, and applications of nuclear processes

    CERN Document Server

    Murray, Raymond L

    1993-01-01

    This expanded, revised, and updated fourth edition of Nuclear Energy maintains the tradition of providing clear and comprehensive coverage of all aspects of the subject, with emphasis on the explanation of trends and developments. As in earlier editions, the book is divided into three parts that achieve a natural flow of ideas: Basic Concepts, including the fundamentals of energy, particle interactions, fission, and fusion; Nuclear Systems, including accelerators, isotope separators, detectors, and nuclear reactors; and Nuclear Energy and Man, covering the many applications of radionuclides, r

  17. Nuclear energy an introduction to the concepts, systems, and applications of nuclear processes

    CERN Document Server

    Murray, Raymond

    2001-01-01

    Nuclear Energy, Fifth Edition provides nuclear engineers, plant designers and radiation physicists with a comprehensive overview of nuclear energy and its uses, discusses potential problems and provides an outlook for the futureNew and important trends are discussed including probabilistic safety analysis (PSA), deregulation of the electric power industry to permit competition in the supply of electricity; improvements in performance characteristics of nuclear power plants, such as capacity factor, production costs, and safety factors; storage and disposal of all types of radioactive w

  18. Critical assessment of methods for treating airborne effluents from high-level waste solidification processes

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J.D.; Pence, D.T.

    1977-06-01

    Off-gas treatment systems are reviewed for high-temperature processes which are being developed for the solidification of high-level liquid wastes from nuclear fuel reprocessing plants. A brief description of each of the processes is given and detailed analyses are made of the expected magnitudes of airborne effluent release rates from each system. The estimated release rates of the various processes are compared with present and anticipated regulatory limits. A number of recommendations are made for additional development studies to better understand and control certain airborne effluents from the solidification processes.

  19. Literature Review: Assessment of DWPF Melter and Melter Off-gas System Lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-30

    Testing to date for the MOC for the Hanford Waste Treatment and Immobilization Plant (WTP) melters is being reviewed with the lessons learned from DWPF in mind and with consideration to the changes in the flowsheet/feed compositions that have occurred since the original testing was performed. This information will be presented in a separate technical report that identifies any potential gaps for WTP processing.

  20. Nomographs for soil vapor extraction and off-gas treatment by activated carbon adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Egemen, E.; Nirmalakhandan, N. [New Mexico State Univ., Las Cruces, NM (United States). Civil, Agricultural, and Geological Engineering Dept.

    1997-12-31

    Soil vapor extraction (SVE) is a widely accepted in-place treatment technology that uses forced air to remove contaminant vapors from zones of permeable vapor flow, thereby enhancing the volatilization of contaminants from the subsurface. The resulting off-gases are contaminated with volatiles and semi-volatiles and have to treated by catalytic or thermal destruction systems, activated carbon adsorbers, or bioreactors. Of these, activated carbon adsorption is the most commonly used technology. From the theoretical foundation of SVE and carbon adsorption, two nomographs were developed for remedial investigation, feasibility studies, planning, operation, and preliminary design purposes. An advantage of such nomographs is that they graphically indicate the sensitivity of the remediation process to different design parameters and critical ranges within a given parameter. In effect, nomographs can help to foster an intuitive understanding of the SVE and adsorption processes itself, which is of considerable value to a process engineer. In addition, such a nomograph provides a utilitarian resource to those who do not have direct access to a comparable computer model. The purpose of this paper is to present the design equations and their use in the development of nomographs for the design of SVE systems and treatment of contaminated air streams by activated carbon canisters.

  1. The recovery of waste and off-gas in Large Combustion Plants subject to IPPC National Permit in Italy.

    Science.gov (United States)

    Di Marco, Giuseppe; Manuzzi, Raffaella

    2017-08-14

    The recovery of off-gas, waste, and biomass in Large Combustion Plants for energy production gives the opportunity to recycle waste and by-products and to recover materials produced in agricultural and industrial activities. The paper illustrates the Italian situation regarding the production of energy from off-gas, biomass, and waste in Large Combustion Plants subject to Integrated Pollution Prevention and Control (IPPC) National Permit. Moreover, it focuses on the 4 Italian Large Combustion Plants producing energy from biomass and waste. For these ones it illustrates the specific issues related to and provides a description of the solutions adopted in the 4 Italian plants. Given that air emission performance is the most relevant aspect of this kind of plants, the paper specifically focuses and reports results about this subject. In particular, in Italy among 113 LCPs subject to IPPC National Permit we have found that 4 plants use as fuel waste (i.e. solid or liquid biomasses and Solid Recovered Fuels), or a mixture of waste and traditional fuels (co-combustion of Solid Recovered Fuels and coal), and that 11 plants use as fuel off-gases listed in Annex X (i.e. Refinery Fuel Gas, Syngas, and gases produced in iron and steel industries). Moreover, there are 2 IPPC chemical plants that recovery energy from different off-gases not listed in Annex X. Regarding the 4 LCPs that produce energy from waste combustion or co-combustion, we find that they take into account all the specific issues related to this kind of plants (i.e. detailed waste characterization, waste acceptance procedures, waste handling and storage, waste pretreatment and emissions to air), and adopt solutions that are best available techniques to prevent pollution. Moreover for one of these plants, the only one for which we have a significant set of monitoring data because it obtained the IPPC National Permit in 2008, we find that energy efficiency and air emissions of the principal pollutants are in

  2. A novel digital pulse processing architecture for nuclear instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Moline, Yoann; Thevenin, Mathieu; Corre, Gwenole [CEA, LIST - Laboratoire Capteurs et Architectures electroniques, F-91191 Gif-sur-Yvette, (France); Paindavoine, Michel [CNRS, Universite de Bourgogne - Laboratoire d' Etude de l' Apprentissage et du Developpement, 21000 DIJON, (France)

    2015-07-01

    The field of nuclear instrumentation covers a wide range of applications, including counting, spectrometry, pulse shape discrimination and multi-channel coincidence. These applications are the topic of many researches, new algorithms and implementations are constantly proposed thanks to advances in digital signal processing. However, these improvements are not yet implemented in instrumentation devices. This is especially true for neutron-gamma discrimination applications which traditionally use charge comparison method while literature proposes other algorithms based on frequency domain or wavelet theory which show better performances. Another example is pileups which are generally rejected while pileup correction algorithms also exist. These processes are traditionally performed offline due to two issues. The first is the Poissonian characteristic of the signal, composed of random arrival pulses which requires to current architectures to work in data flow. The second is the real-time requirement, which implies losing pulses when the pulse rate is too high. Despite the possibility of treating the pulses independently from each other, current architectures paralyze the acquisition of the signal during the processing of a pulse. This loss is called dead-time. These two issues have led current architectures to use dedicated solutions based on re-configurable components like Field Programmable Gate Arrays (FPGAs) to overcome the need of performance necessary to deal with dead-time. However, dedicated hardware algorithm implementations on re-configurable technologies are complex and time-consuming. For all these reasons, a programmable Digital pulse Processing (DPP) architecture in a high level language such as Cor C++ which can reduce dead-time would be worthwhile for nuclear instrumentation. This would reduce prototyping and test duration by reducing the level of hardware expertise to implement new algorithms. However, today's programmable solutions do not meet

  3. Synthesis of tetrakis (hydroxymethyl) phosphonium chloride by high-concentration phosphine in industrial off-gas.

    Science.gov (United States)

    Huang, Xiaofeng; Wei, Yanfu; Zhou, Tao; Qin, Yangsong; Gao, Kunyang; Ding, Xinyue

    2013-01-01

    With increasing consumption of phosphate rock and acceleration of global phosphate production, the shortage of phosphate resources is increasing with the development and utilization of phosphate. China's Ministry of Land and Resources has classified phosphate as a mineral that cannot meet China's growing demand for phosphate rock in 2010. The phosphorus chemical industry is one of the important economic pillars for Yunnan province. Yellow phosphorus production in enterprises has led to a significant increase in the amount of phosphorus sludge. This paper focuses on phosphine generation in the process of phosphoric sludge utilization, where the flame retardant tetrakis (hydroxymethyl) phosphonium chloride (THPC) is synthesized by high concentrations of phosphine. The optimum conditions are determined at a space velocity of 150 h(-1), a reaction temperature of 60 °C, 0.75 g of catalyst, and a ratio of raw materials of 4:1. Because of the catalytic oxidation of copper chloride (CuCl2), the synthesis of THPC was accelerated significantly. In conclusion, THPC can be efficiently synthesized under optimal conditions and with CuCl2 as a catalyst.

  4. FRENDY: A new nuclear data processing system being developed at JAEA

    Science.gov (United States)

    Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

    2017-09-01

    JAEA has provided an evaluated nuclear data library JENDL and nuclear application codes such as MARBLE, SRAC, MVP and PHITS. These domestic codes have been widely used in many universities and industrial companies in Japan. However, we sometimes find problems in imported processing systems and need to revise them when the new JENDL is released. To overcome such problems and immediately process the nuclear data when it is released, JAEA started developing a new nuclear data processing system, FRENDY in 2013. This paper describes the outline of the development of FRENDY and both its capabilities and performances by the analyses of criticality experiments. The verification results indicate that FRENDY properly generates ACE files.

  5. Online gas composition estimation in solid oxide fuel cell systems with anode off-gas recycle configuration

    Science.gov (United States)

    Dolenc, B.; Vrečko, D.; Juričić, Ð.; Pohjoranta, A.; Pianese, C.

    2017-03-01

    Degradation and poisoning of solid oxide fuel cell (SOFC) stacks are continuously shortening the lifespan of SOFC systems. Poisoning mechanisms, such as carbon deposition, form a coating layer, hence rapidly decreasing the efficiency of the fuel cells. Gas composition of inlet gases is known to have great impact on the rate of coke formation. Therefore, monitoring of these variables can be of great benefit for overall management of SOFCs. Although measuring the gas composition of the gas stream is feasible, it is too costly for commercial applications. This paper proposes three distinct approaches for the design of gas composition estimators of an SOFC system in anode off-gas recycle configuration which are (i.) accurate, and (ii.) easy to implement on a programmable logic controller. Firstly, a classical approach is briefly revisited and problems related to implementation complexity are discussed. Secondly, the model is simplified and adapted for easy implementation. Further, an alternative data-driven approach for gas composition estimation is developed. Finally, a hybrid estimator employing experimental data and 1st-principles is proposed. Despite the structural simplicity of the estimators, the experimental validation shows a high precision for all of the approaches. Experimental validation is performed on a 10 kW SOFC system.

  6. Startup and initial operation of a DFGD and pulse jet fabric filter system on Cokenergy's Indiana Harbor coke oven off gas system

    Energy Technology Data Exchange (ETDEWEB)

    Morris, W.J.; Gansley, R.R.; Schaddell, J.G.

    1999-07-01

    This paper describes the design, initial operation and performance testing of a Dry Flue Gas Desulfurization (DFGD) and Modular Pulse Jet Fabric Filter (MPJFF) system installed at Cokenergy's site in East Chicago, Indiana. The combined flue gas from the sixteen (16) waste heat recovery boilers is processed by the system to control emissions of sulfur dioxide and particulates. These boilers recover energy from coke oven off gas from Indiana Harbor Coke Company's coke batteries. The DFGD system consists of two 100% capacity absorbers. Each absorber vessel uses a single direct drive rotary atomizer to disperse the lime slurry for SO{sub 2} control. The MPJFF consists of thirty two (32) modules arranged in twin sixteen-compartment (16) units. The initial start up of the DFGD/MPJFF posed special operational issues due to the low initial gas flows through the system as the four coke oven batteries were cured and put in service for the first time. This occurred at approximately monthly intervals beginning in March 1998. A plan was implemented to perform a staged startup of the DFGD and MPJFF to coincide with the staged start up of the coke batteries and waste heat boilers. Operational issues that are currently being addressed include reliability of byproduct removal. Performance testing was conducted in August and September 1998 at the inlet of the system and the outlet stack. During these tests, particulate, SO{sub 2}, SO{sub 3}, and HCI emissions were measured simultaneously at the common DFGD inlet duct and the outlet stack. Measurements were also taken for average lime, water, and power consumption during the tests as well as system pressure losses. These results showed that all guarantee parameters were achieved during the test periods. The initial operation and performance testing are described in this paper.

  7. WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008

    Energy Technology Data Exchange (ETDEWEB)

    Fellinger, A.

    2009-10-15

    The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

  8. Kinematics of high-energy nuclear processes; Cinematica de los procesos nucleares de alta energia

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez del Rio, C.

    1972-07-01

    This report is the first draft of one of the chapters of a book being prepared under the title:Topics on Practical Nuclear Physics. It is published as a report because of its immediate educational value and in order to include in its final draft the suggestions of the readers. (Author)

  9. Kinematics of low-energy nuclear processes; Cinematica de los procesos nucleares de baja energia

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez del Rio, C.

    1972-07-01

    This report is the first draft of one of the chapters of a book being prepared under the title: Topics on Practical Nuclear Physics. It is published as a report because of its immediate educational value and in order to include in its final draft the suggestions of the readers. (Author)

  10. Quark Energy Loss and Shadowing in Nuclear Drell-Yan Process

    Institute of Scientific and Technical Information of China (English)

    DUAN Chun-Gui; CUI Shu-Wen; YAN Zhan-Yuan

    2005-01-01

    The energy loss effect in nuclear matter is another nuclear effect apart from the nuclear effects on the parton distribution as in deep inelastic scattering process. The quark energy loss can be measured best by the nuclear dependence of the high energy nuclear Drell-Yan process. By means of three kinds of quark energy loss parameterizations given in literature and the nuclear parton distribution extracted only with lepton-nucleus deep inelastic scattering experimental data, measured Drell-Yan production cross sections are analyzed for 800 GeV proton incident on a variety of nuclear targets from FNAL E866. It is shown that our results with considering the energy loss effect are much different from those of the FNAL E866, who analyzes the experimental data with the nuclear parton distribution functions obtained by using the deep inelastic IA collisions and pA nuclear Drell-Yan data. Considering the existence of energy loss effect in Drell-Yan lepton pairs production, we suggest that the extraction of nuclear parton distribution functions should not include Drell-Yan experimental data.

  11. Quark Energy Loss and Shadowing in Nuclear Drell-Yan Process

    Institute of Scientific and Technical Information of China (English)

    DUANChun-Gui; CUIShu-Wen; YANZhan-Yuan

    2005-01-01

    The energy loss effect in nuclear matter is another nuclear effect apart from the nuclear effects on the parton distribution as in deep inelastic scattering process. The quark energy loss can be measured best by the nuclear dependence of the high energy nuclear Dre11-Yan process. By means of three kinds of quark energy loss parameterizations given in literature and the nuclear parton distribution extracted only with lepton-nucleus deep inelastic scattering experimental data, measured Dre11-Yan production cross sections are analyzed for 800 GeV proton incident on a variety of nuclear targets from FNAL E866. It is shown that our results with considering the energy loss effect are much different from those of the FNAL E866, who analyzes the experimental data with the nuclear parton distribution functions obtained by using the deep inelastic IA collisions and pA nuclear Drell-Yan data. Considering the existence of energy loss effect in Drell-Yan lepton pairs production, we suggest that the extraction of nuclear parton distribution functions shoul""""d not include Dre11-Yan experimental data.

  12. The NJOY nuclear data processing system: Volume 2, The NJOY, RECONR, BROADR, HEATR, and THERMR modules

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, R.E.; Muir, D.W.; Boicourt, R.M.

    1982-05-01

    The NJOY nuclear data processing system is a comprehensive computer code package for producing cross sections and related nuclear parameters from ENDF/B evaluated nuclear data. This volume provides detailed descriptions of the NJOY module, which contains the executive program and utility subroutines used by the other modules, and it discusses the theory and computational methods of four of the modules used for producing pointwise cross sections: RECONR, BROADR, HEATR, and THERMR.

  13. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  14. Spent Nuclear Fuel (SNF) Process Validation Technical Support Plan

    Energy Technology Data Exchange (ETDEWEB)

    SEXTON, R.A.

    2000-03-13

    The purpose of Process Validation is to confirm that nominal process operations are consistent with the expected process envelope. The Process Validation activities described in this document are not part of the safety basis, but are expected to demonstrate that the process operates well within the safety basis. Some adjustments to the process may be made as a result of information gathered in Process Validation.

  15. Urgent problems of the vector meson production in nuclear processes

    CERN Document Server

    Titov, A I

    2001-01-01

    A brief review of the topical problems related to dynamics of vector meson is given. In particular, in-medium modification of the vector meson properties in hot and nuclear matter, photoproduction of the phi-mesons as a probe for hidden strangeness in a nucleon phi-, omega-production and OZI-rule violation, omega-production as a tool for studying the baryon resonance properties are discussed

  16. Field demonstration for bioremediation treatment: Technology demonstration of soil vapor extraction off-gas at McClellan Air Force Base. Final report November 1997--April 1998

    Energy Technology Data Exchange (ETDEWEB)

    Magar, V.S.; Tonga, P.; Webster, T.; Drescher, E.

    1999-01-12

    McClellan Air Force Base (AFB) is a National Test Location designated through the Strategic Environmental Research and Development Program (SERDP), and was selected as the candidate test site for a demonstration of soil vapor extraction (SVE) off-gas treatment technology. A two-stage reactor system was employed for the treatment of the off-gas. The biological treatment was conducted at Operable Unit (OU) D Site S, located approximately 400 ft southwest of Building 1093. The SVE system at this area normally operates at a nominal volumetric flowrate of approximately 500 to 600 standard cubic feet per minute (scfm). The contaminated air stream from the SVE system that was fed to the reactor system operated at a flowrate of 5 to 10 scfm. The two-stage reactor system consisted of a fixed-film biofilter followed by a completely mixed (by continuous stirring), suspended-growth biological reactor. This reactor configuration was based on a review of the literature, on characterization of the off-gas from the SVE system being operated at McClellan AFB, and on the results of the laboratory study conducted by Battelle and Envirogen for this study.

  17. Processing used nuclear fuel with nanoscale control of uranium and ultrafiltration

    Science.gov (United States)

    Wylie, Ernest M.; Peruski, Kathryn M.; Prizio, Sarah E.; Bridges, Andrea N. A.; Rudisill, Tracy S.; Hobbs, David T.; Phillip, William A.; Burns, Peter C.

    2016-05-01

    Current separation and purification technologies utilized in the nuclear fuel cycle rely primarily on liquid-liquid extraction and ion-exchange processes. Here, we report a laboratory-scale aqueous process that demonstrates nanoscale control for the recovery of uranium from simulated used nuclear fuel (SIMFUEL). The selective, hydrogen peroxide induced oxidative dissolution of SIMFUEL material results in the rapid assembly of persistent uranyl peroxide nanocluster species that can be separated and recovered at moderate to high yield from other process-soluble constituents using sequestration-assisted ultrafiltration. Implementation of size-selective physical processes like filtration could results in an overall simplification of nuclear fuel cycle technology, improving the environmental consequences of nuclear energy and reducing costs of processing.

  18. The nuclear safeguards system and the process of global governance accountability

    Energy Technology Data Exchange (ETDEWEB)

    Xavier, Roberto Salles, E-mail: xavier@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Coordenacao Geral de Planejamento e Avaliacao

    2011-07-01

    Due to rising energy costs and climate concerns, nuclear energy is again being seriously considered as an energy source for several countries. Along with the resurgence of nuclear energy comes the concern of the world if these countries will develop their programs for the peaceful use of nuclear energy. If on one hand the growth potential of nuclear energy should not be stifled, on the other hand it is imperative that a climate of mutual trust is developed, respecting the right of each country to develop its nuclear program without taking a climate of mistrust to a possible 'intention' behind the pursuit of peaceful use of nuclear energy. Therefore, it is essential that appropriate mechanisms of accountability of global governance are institutionalized at the institutional architecture of the international process of nuclear safeguards, more specifically to the nuclear fuel cycle, so that abuses of power in this sphere does not happen, both by countries that aspire to develop projects nuclear, and by the suppliers of technology. In this context, the case study of Brazil and Argentina gained importance, because these two countries have a single binational organization of nuclear safeguards in the world: Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials - ABACC. In the theoretical question, the paper tries to understand what happens with the process of legitimacy and authority of the organizations of global governance by analyzing the degree of publicness and constrictiveness. This work intends to focus on the role of ABACC as an interstate institution of accountability, which has a key role to control the nation States of Brazil and Argentina regarding the appropriate use of nuclear material used in their programs, and analyze how this Agency behaves within of tension legitimacy-authority, taking into account existing studies on accountability in global governance. (author)

  19. People detection in nuclear plants by video processing for safety purpose

    Energy Technology Data Exchange (ETDEWEB)

    Jorge, Carlos Alexandre F.; Mol, Antonio Carlos A., E-mail: calexandre@ien.gov.b, E-mail: mol@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN), Rio de Janeiro, RJ (Brazil); Seixas, Jose M.; Silva, Eduardo Antonio B., E-mail: seixas@lps.ufrj.b, E-mail: eduardo@lps.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Eletrica; Cota, Raphael E.; Ramos, Bruno L., E-mail: brunolange@poli.ufrj.b [Universidade Federal do Rio de Janeiro (EP/UFRJ), RJ (Brazil). Dept. de Engenharia Eletronica e de Computacao

    2011-07-01

    This work describes the development of a surveillance system for safety purposes in nuclear plants. The final objective is to track people online in videos, in order to estimate the dose received by personnel, during the execution of working tasks in nuclear plants. The estimation will be based on their tracked positions and on dose rate mapping in a real nuclear plant at Instituto de Engenharia Nuclear, Argonauta nuclear research reactor. Cameras have been installed within Argonauta's room, supplying the data needed. Both video processing and statistical signal processing techniques may be used for detection, segmentation and tracking people in video. This first paper reports people segmentation in video using background subtraction, by two different approaches, namely frame differences, and blind signal separation based on the independent component analysis method. Results are commented, along with perspectives for further work. (author)

  20. Public participation processes related to nuclear research installations: what are the driving factors behind participation intention?

    Science.gov (United States)

    Turcanu, Catrinel; Perko, Tanja; Laes, Erik

    2014-04-01

    This article addresses organised public participation processes related to installations for nuclear research. The aim was to determine predictors that could provide an empirical insight into the motivations underlying people's intended level of involvement. The results highlight attitude towards participation and moral norm as the strongest predictors for participation intention. Other significant predictors were time constraints, attitude towards nuclear energy, subjective and descriptive norms, and knowledge. An opposing relationship is noted between participation intention and attitude towards nuclear energy. At the same time, people who are more knowledgeable about the nuclear domain seem more willing to get involved. The analysis also revealed that financial benefits do not influence people's intended involvement in participation processes related to nuclear research installations. The results reported here are based on empirical data from a large-scale public opinion survey (N = 1020) carried out in Belgium during May-June 2011.

  1. Impact of nuclear fission on r-process nucleosynthesis and origin of solar r-process elements

    Energy Technology Data Exchange (ETDEWEB)

    Shibagaki, Shota, E-mail: shota.shibagaki@nao.ac.jp [Department of Astronomy, Graduate School of Science, University of Tokyo, 2 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-0033, Japan and National Astronomical Observatory of Japan, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan); Kajino, Toshitaka [National Astronomical Observatory of Japan, 2-21-1 Osawa, Mitaka, Tokyo 181-8588, Japan and Department of Astronomy, Graduate School of Science, University of Tokyo, 2 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-0033 (Japan); Mathews, Grant J. [Center for Astrophysics, Department of Physics, University of Notre Dame, Notre Dame, IN 46556 (United States); Chiba, Satoshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo, 152-8850 (Japan)

    2015-02-24

    Binary neutron star mergers (NSMs) are expected to be main production sites of r-process elements. Their ejecta are extremely neutron-rich (Y{sub e}<0.1), and the r-process path proceeds along the neutron drip line and enters the region of fissile nuclei. In this situation, although superheavy nuclei may be synthesized and the r-process path may reach the island of stability, those are sensitive to theoretical models of nuclear masses and nuclear fission. In this study, we carry out r-process nucleosynthesis simulations in the NSMs. Our new nuclear reaction network code include new theoretical models of nuclear masses and nuclear fission. Our r-process simulation of a binary NSM shows that the final r-process elemental abundances exhibit flat pattern for A∼110-160, and several fission cycling operate in extremely neutron-rich conditions of the NSM. We find that the combination of the NSMs and the magnetorotational supernovae can reproduce the solar r-process elements. We discuss the validity of this interpretation.

  2. The nuclear processes responsible for the CNO synthesis

    CERN Document Server

    Arnould, M; Goriely, S

    2002-01-01

    The abundances of the isotopes of the elements C, N and O are mainly affected by the cold CNO cycles in non-explosive stellar situations, or by the hot CNO chains that can develop in certain explosive sites, like classical novae. Helium burning phases can modify the composition of the ashes of the CNO transmutations through several $\\alpha$-capture reactions, the most famed one being 12C(a,g)16O. This contribution presents a short review of the purely nuclear physics limitations imposed on the accuracy of the predicted C, N and O yields from H-burning in non-explosive stars or novae. This analysis makes largely use of the NACRE compilation for the rates of the reactions on stable targets making up the cold CNO cycle. Some more recent rate determinations are also considered. The analysis of the impact of the rate uncertainties on the abundance predictions is conducted in the framework of a simple parametric astrophysical model. These calculations have the virtue of being a guide in the selection of the nuclear...

  3. Reverse engineering nuclear properties from rare earth abundances in the r process

    Science.gov (United States)

    Mumpower, M. R.; McLaughlin, G. C.; Surman, R.; Steiner, A. W.

    2017-03-01

    The bulk of the rare earth elements are believed to be synthesized in the rapid neutron capture process or r process of nucleosynthesis. The solar r-process residuals show a small peak in the rare earths around A∼ 160, which is proposed to be formed dynamically during the end phase of the r process by a pileup of material. This abundance feature is of particular importance as it is sensitive to both the nuclear physics inputs and the astrophysical conditions of the main r process. We explore the formation of the rare earth peak from the perspective of an inverse problem, using Monte Carlo studies of nuclear masses to investigate the unknown nuclear properties required to best match rare earth abundance sector of the solar isotopic residuals. When nuclear masses are changed, we recalculate the relevant β-decay properties and neutron capture rates in the rare earth region. The feedback provided by this observational constraint allows for the reverse engineering of nuclear properties far from stability where no experimental information exists. We investigate a range of astrophysical conditions with this method and show how these lead to different predictions in the nuclear properties influential to the formation of the rare earth peak. We conclude that targeted experimental campaigns in this region will help to resolve the type of conditions responsible for the production of the rare earth nuclei, and will provide new insights into the longstanding problem of the astrophysical site(s) of the r process.

  4. Linking legacies: Connecting the Cold War nuclear weapons production processes to their environmental consequences

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    In the aftermath of the Cold War, the US has begun addressing the environmental consequences of five decades of nuclear weapons production. In support of this effort, the National Defense Authorization Act for Fiscal Year 1995 directed the Department of Energy (DOE) to describe the waste streams generated during each step in the production of nuclear weapons. Accordingly, this report responds to this mandate, and it is the Department`s first comprehensive analysis of the sources of waste and contamination generated by the production of nuclear weapons. The report also contains information on the missions and functions of nuclear weapons facilities, on the inventories of waste and materials remaining at these facilities, as well as on the extent and characteristics of contamination in and around these facilities. This analysis unites specific environmental impacts of nuclear weapons production with particular production processes. The Department used historical records to connect nuclear weapons production processes with emerging data on waste and contamination. In this way, two of the Department`s legacies--nuclear weapons manufacturing and environmental management--have become systematically linked. The goal of this report is to provide Congress, DOE program managers, non-governmental analysts, and the public with an explicit picture of the environmental results of each step in the nuclear weapons production and disposition cycle.

  5. A Hydrogen Containment Process for Nuclear Thermal Engine Ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    The objective of this study is to propose a new total hydrogen containment process to enable the testing required for NTP engine development. This H2 removal process comprises of two unit operations: an oxygen-rich burner and a shell-and-tube type of heat exchanger. This new process is demonstrated by simulation of the steady state operation of the engine firing at nominal conditions.

  6. Radioactive Dry Process Material Treatment Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Hung, I. H.; Kim, K. K. (and others)

    2007-06-15

    The project 'Radioactive Dry Process Material Treatment Technology Development' aims to be normal operation for the experiments at DUPIC fuel development facility (DFDF) and safe operation of the facility through the technology developments such as remote operation, maintenance and pair of the facility, treatment of various high level process wastes and trapping of volatile process gases. DUPIC Fuel Development Facility (DFDF) can accommodate highly active nuclear materials, and now it is for fabrication of the oxide fuel by dry process characterizing the proliferation resistance. During the second stage from march 2005 to February 2007, we carried out technology development of the remote maintenance and the DFDF's safe operation, development of treatment technology for process off-gas, and development of treatment technology for PWR cladding hull and the results was described in this report.

  7. Nuclear Transparency with the gamma + n -> pi- + p Process in 4He

    Energy Technology Data Exchange (ETDEWEB)

    Dipangkar Dutta; Feng Xiong; Lingyan Zhu; John Arrington; Todd Averett; Elizabeth Beise; John Calarco; Ting Chang; Jian-Ping Chen; Eugene Chudakov; Marius Coman; Benjamin Clasie; Christopher Crawford; Sonja Dieterich; Frank Dohrmann; Kevin Fissum; Salvatore Frullani; Haiyan Gao; Ronald Gilman; Charles Glashausser; Javier Gomez; Kawtar Hafidi; Jens-Ole Hansen; Douglas Higinbotham; R.J. Holt; Cornelis De Jager; Xiaochao Zheng; X. Jiang; Edward Kinney; Kevin Kramer; Gerfried Kumbartzki; John LeRose; Nilanga Liyanage; David Mack; Pete Markowitz; Kathy McCormick; Zein-Eddine Meziani; Robert Michaels; J. Mitchell; Sirish Nanda; David Potterveld; Ronald Ransome; Paul Reimer; Bodo Reitz; Arunava Saha; Elaine Schulte; Charles Seely; Simon Sirca; Steffen Strauch; Vincent Sulkosky; Branislav Vlahovic; Lawrence Weinstein; Krishni Wijesooriya; Claude Williamson; Bogdan Wojtsekhowski; Hong XIANG; Wang Xu; J. Zeng

    2003-08-01

    We have measured the nuclear transparency of the fundamental process {gamma} n {yields} {pi}{sup -} p in {sup 4}He. These measurements were performed at Jefferson Lab in the photon energy range of 1.6 to 4.5 GeV and at {theta}{sub cm}{sup {pi}} = 70{sup o} and 90{sup o}. These measurements are the first of their kind in the study of nuclear transparency in photoreactions. They also provide a benchmark test of Glauber calculations based on traditional models of nuclear physics. The transparency results suggest deviations from the traditional nuclear physics picture. The momentum transfer dependence of the measured nuclear transparency is consistent with Glauber calculations which include the quantum chromodynamics phenomenon of color transparency.

  8. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. [and others

    2002-05-01

    In this study, the DUPIC fuel fabrication technology for DUPIC pellet and element satisfying the standard specification was verified through (1) the improvement of fabrication technology and equipment, (2) remote operation of fuel manufacturing and inspection equipment installed at DFDF and (3) the study on the material properties of DUPIC fuel. The blending process was newly developed for making DUPIC powder composition homogeneous, and mixing process was added to the DUPIC process flow for fabricating crack-free pellets. A series of fabrication experiments were carried out in terms of various process conditions. Based on these experimental results, the optimal process flow and conditions for DUPIC fuel fabrication were established. 6 DUPIC elements and 6 mini-elements for irradiation test in HANARO was successfully fabricated using 7.4 kg of spent PWR fuel in 2000. The process qualification tests has been performed using 10 kg of spent PWR fuel since May 2001. The optimal DUPIC fuel fabrication process meeting AECL's quality requirements has been established and qualified. Quality assurance system for DUPIC fuel fabrication was also established in cooperation of AECL.

  9. Nuclear waste management. Quarterly progress report, April-June 1981

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.; Powell, J.A.

    1981-09-01

    Reports and summaries are presented for the following: high-level waste process development; alternative waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; and analysis of spent fuel policy implementation.

  10. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    D.E. Clark; D.C. Folz

    2010-08-29

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  11. Radioanalytical Chemistry for Automated Nuclear Waste Process Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Jay W. Grate; Timothy A. DeVol

    2006-07-20

    The objectives of our research were to develop the first automated radiochemical process analyzer including sample pretreatment methodoology, and to initiate work on new detection approaches, especially using modified diode detectors.

  12. GALILEE: A nuclear data processing system for transport, depletion and shielding codes

    Energy Technology Data Exchange (ETDEWEB)

    COSTE-DELCLAUX, Mireille [Commissariat a l' Energie Atomique, CEA Saclay, DEN/DANS/DM2S/SERMA/LLPR, 91191 Gif sur Yvette CEDEX (France)

    2008-07-01

    The Nuclear Data Processing System for Transport, Depletion and Shielding Codes GALILEE is part of a CEA global development program dedicated to fine modelling of nuclear systems. The other projects contributing to this aim are APOLLO3 inherited from DESCARTES (Calvin and Fedon-Magnaud, 2007) which treats deterministic transport, TRIPOLI-4 (Diop et al., 2006) which treats Monte Carlo transport and DARWIN3 (Tsilanizara et al., 1999) which solves all fuel cycle problems. GALILEE aims are: - To provide to application codes (deterministic or Monte Carlo transport codes, shielding codes or depletion codes), a tool-box allowing a consistent processing for nuclear data coming from any evaluation given in ENDF-6 format, - To carry out an automatic chain for creating application libraries, - To provide consistent application libraries for modelling a nuclear system. GALILEE project is carried out in synergy with application codes in order to be able to share 'objects' but also 'tools'. (author)

  13. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs.

  14. The impact of individual nuclear properties on $r$-process nucleosynthesis

    CERN Document Server

    Mumpower, M R; McLaughlin, G C; Aprahamian, A

    2015-01-01

    The astrophysical rapid neutron capture process or `$r$ process' of nucleosynthesis is believed to be responsible for the production of approximately half the heavy element abundances found in nature. This multifaceted problem remains one of the greatest open challenges in all of physics. Knowledge of nuclear physics properties such as masses, $\\beta$-decay and neutron capture rates, as well as $\\beta$-delayed neutron emission probabilities are critical inputs that go into calculations of $r$-process nucleosynthesis. While properties of nuclei near stability have been established, much still remains unknown regarding neutron-rich nuclei far from stability that may participate in the $r$ process. Sensitivity studies gauge the astrophysical response of a change in nuclear physics input(s) which allows for the isolation of the most important nuclear properties that shape the final abundances observed in nature. This review summarizes the extent of recent sensitivity studies and highlights how these studies play ...

  15. Synthesis and evaluation of potential ligands for nuclear waste processing

    NARCIS (Netherlands)

    Iqbal, M.

    2012-01-01

    The research presented in this thesis deals with the synthesis and evaluation of new potential ligands for the complexation of actinide and lanthanide ions either for their extraction from bulk radioactive waste or their stripping from an extracted organic phase for final processing of the waste. In

  16. Synthesis and evaluation of potential ligands for nuclear waste processing

    NARCIS (Netherlands)

    Iqbal, M.

    2012-01-01

    The research presented in this thesis deals with the synthesis and evaluation of new potential ligands for the complexation of actinide and lanthanide ions either for their extraction from bulk radioactive waste or their stripping from an extracted organic phase for final processing of the waste. In

  17. Reverse engineering nuclear properties from rare earth abundances in the $r$ process

    CERN Document Server

    Mumpower, M R; Surman, R; Steiner, A W

    2016-01-01

    The bulk of the rare earth elements are believed to be synthesized in the rapid neutron capture process or $r$ process of nucleosynthesis. The solar $r$-process residuals show a small peak in the rare earths around $A\\sim 160$, which is proposed to be formed dynamically during the end phase of the $r$ process by a pileup of material. This abundance feature is of particular importance as it is sensitive to both the nuclear physics inputs and the astrophysical conditions of the main $r$ process. We explore the formation of the rare earth peak from the perspective of an inverse problem, using Monte Carlo studies of nuclear masses to investigate the unknown nuclear properties required to best match rare earth abundance sector of the solar isotopic residuals. When nuclear masses are changed, we recalculate the relevant $\\beta$-decay properties and neutron capture rates in the rare earth region. The feedback provided by this observational constraint allows for the reverse engineering of nuclear properties far from ...

  18. CORROSION ISSUES ASSOCIATED WITH AUSTENITIC STAINLESS STEEL COMPONENTS USED IN NUCLEAR MATERIALS EXTRACTION AND SEPARATION PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.; Louthan, M.; Sindelar, R.

    2012-12-17

    This paper illustrated the magnitude of the systems, structures and components used at the Savannah River Site for nuclear materials extraction and separation processes. Corrosion issues, including stress corrosion cracking, pitting, crevice corrosion and other corrosion induced degradation processes are discussed and corrosion mitigation strategies such as a chloride exclusion program and corrosion release testing are also discussed.

  19. Nuclear regulatory process in the UK: the road to nuclear license; Proceso regulatorio nuclear en el Reino Unido: el camino hacia la licencia nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Prieto, M.

    2012-07-01

    Relaunching a nuclear-based electric power generating system in any country, even in one like the United Kingdom (UK) that already has an adequate regulatory system that is pioneering in Europe and the world and has had several plants in operation for many years, always involves a series of far-reaching reforms of an institutional and national nature which, at the same time, serve to raise the standards of nuclear safety to the highest level, in order to reduce the risk run by the investors in the program and to ensure that both regulatory bodies and governments fulfill their fundamental function; serve and protect the citizen and always strive to improve the country's well-being. (Author)

  20. Mock Nuclear Processing Facility-Safeguards Training Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Gibbs, Philip [Brookhaven National Lab. (BNL), Upton, NY (United States); Hasty, Tim [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johns, Rissell [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Baum, Gregory [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-08-31

    This document outlines specific training requirements in the topical areas of Material Control and Accounting (MC&A) and Physical Protection(PP) which are to be used as technical input for designing a mock Integrated Security Facility (ISF) at Sandia National Laboratories (SNL). The overall project objective for these requirements is to enhance the ability to deliver training on Material Protection Control and Accounting (MC&A) concepts regarding hazardous material such as irradiated materials with respect to bulk processing facilities.

  1. Solvent degradation products in nuclear fuel processing solvents

    Energy Technology Data Exchange (ETDEWEB)

    Shook, H.E. Jr.

    1988-06-01

    The Savannah River Plant uses a modified Purex process to recover enriched uranium and separate fission products. This process uses 7.5% tri-n-butyl phosphate (TBP) dissolved in normal paraffin hydrocarbons for the solvent extraction of a nitric acid solution containing the materials to be separated. Periodic problems in product decontamination result from solvent degradation. A study to improve process efficiency has identified certain solvent degradation products and suggested mitigation measures. Undecanoic acid, lauric acid, and tridecanoic acid were tentatively identified as diluent degradation products in recycle solvent. These long-chain organic acids affect phase separation and lead to low decontamination factors. Solid phase extraction (SPE) was used to concentrate the organic acids in solvent prior to analysis by high performance liquid chromatography (HPLC). SPE and HPLC methods were optimized in this work for analysis of decanoic acid, undecanoic acid, and lauric acid in solvent. Accelerated solvent degradation studies with 7.5% TBP in normal paraffin hydrocarbons showed that long-chain organic acids and long-chain alkyl butyl phosphoric acids are formed by reactions with nitric acid. Degradation of both tributyl phosphate and hydrocarbon can be minimized with purified normal paraffin replacing the standard grade presently used. 12 refs., 1 fig., 3 tabs.

  2. Application of telerobotic control to remote processing of nuclear material

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, R.D.; Grasz, E.L.; Herget, C.J.; Gavel, D.T.; Addis, R.B.; DeMinico, G.A.

    1991-07-08

    In processing radioactive material there are certain steps which have customarily required operators working at glove box enclosures. This can subject the operators to low level radiation dosages and the risk of accidental contamination, as well as generate significant radioactive waste to accommodate the human interaction. An automated system is being developed to replace the operator at the glove box and thus remove the human from these risks, and minimize waste. Although most of the processing can be automated with very little human operator interaction, there are some tasks where intelligent intervention is necessary to adapt to unexpected circumstances and events. These activities will require that the operator be able to interact with the process using a remote manipulator in a manner as natural as if the operator were actually in the work cell. This robot-based remote manipulation system, or telerobot, must provide the operator with an effective means of controlling the robot arm, gripper and tools. This paper describes the effort in progress in Lawrence Livermore National Laboratory to achieve this capability. 8 refs.

  3. Consideration of Command and Control Performance during Accident Management Process at the Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Nisrene M. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The accident at the Fukushima Daiichi nuclear power plants shifted the nuclear safety paradigm from risk management to on-site management capability during a severe accident. The kernel of on-site management capability during an accident at a nuclear power plant is situation awareness and agility of command and control. However, little consideration has been given to accident management. After the events of September 11, 2001 and the catastrophic Fukushima nuclear disaster, agility of command and control has emerged as a significant element for effective and efficient accident management, with many studies emphasizing accident management strategies, particularly man-machine interface, which is considered a key role in ensuring nuclear power plant safety during severe accident conditions. This paper proposes a conceptual model for evaluating command and control performance during the accident management process at a nuclear power plant. Communication and information processing while responding to an accident is one of the key issues needed to mitigate the accident. This model will give guidelines for accurate and fast communication response during accident conditions.

  4. Rate and peak concentrations of off-gas emissions in stored wood pellets--sensitivities to temperature, relative humidity, and headspace volume.

    Science.gov (United States)

    Kuang, Xingya; Shankar, Tumuluru Jaya; Bi, Xiaotao T; Lim, C Jim; Sokhansanj, Shahab; Melin, Staffan

    2009-11-01

    Wood pellets emit CO, CO(2), CH(4), and other volatiles during storage. Increased concentration of these gases in a sealed storage causes depletion of concentration of oxygen. The storage environment becomes toxic to those who operate in and around these storages. The objective of this study was to investigate the effects of temperature, moisture, and the relative size of storage headspace on emissions from wood pellets in an enclosed space. Twelve 10-l plastic containers were used to study the effects of headspace ratio (25, 50, and 75% of container volume) and temperatures (10-50 degrees C). Another eight containers were set in uncontrolled storage relative humidity (RH) and temperature. Concentrations of CO(2), CO, and CH(4) were measured by gas chromatography (GC). The results showed that emissions of CO(2), CO, and CH(4) from stored wood pellets are more sensitive to storage temperature than to RH and the relative volume of headspace. Higher peak emission factors are associated with higher temperatures. Increased headspace volume ratio increases peak off-gas emissions because of the availability of oxygen associated with pellet decomposition. Increased RH in the enclosed container increases the rate of off-gas emissions of CO(2), CO, and CH(4) and oxygen depletion.

  5. Efficient carbon dioxide utilization and simultaneous hydrogen enrichment from off-gas of acetone-butanol-ethanol fermentation by succinic acid producing Escherichia coli.

    Science.gov (United States)

    He, Aiyong; Kong, Xiangping; Wang, Chao; Wu, Hao; Jiang, Min; Ma, Jiangfeng; Ouyang, Pingkai

    2016-08-01

    The off-gas from acetone-butanol-ethanol (ABE) fermentation was firstly used to be CO2 source (co-substrate) for succinic acid production. The optimum ratio of H2/CO2 indicated higher CO2 partial pressures with presence of H2 could enhance C4 pathway flux and reductive product productivity. Moreover, when an inner recycling bioreactor was used for CO2 recycling at a high total pressure (0.2Mpa), a maximum succinic acid concentration of 65.7g·L(-1) was obtained, and a productivity of 0.76g·L(-1)·h(-1) and a high yield of 0.86g·g(-1) glucose were achieved. Furthermore, the hydrogen content was simultaneously enriched to 92.7%. These results showed one successful attempt to reuse the off-gas of ABE fermentation which can be an attractive CO2 source for succinic acid production. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. The development of a neuroscience-based methodology for the nuclear energy learning/teaching process

    Energy Technology Data Exchange (ETDEWEB)

    Barabas, Roberta de C.; Sabundjian, Gaiane, E-mail: robertabarabas@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    When compared to other energy sources such as fossil fuels, coal, oil, and gas, nuclear energy has perhaps the lowest impact on the environment. Moreover, nuclear energy has also benefited other fields such as medicine, pharmaceutical industry, and agriculture, among others. However, despite all benefits that result from the peaceful uses of nuclear energy, the theme is still addressed with prejudice. Education may be the starting point for public acceptance of nuclear energy as it provides pedagogical approaches, learning environments, and human resources, which are essential conditions for effective learning. So far nuclear energy educational researches have been conducted using only conventional assessment methods. The global educational scenario has demonstrated absence of neuroscience-based methods for the teaching of nuclear energy, and that may be an opportunity for developing new strategic teaching methods that will help demystifying the theme consequently improving public acceptance of this type of energy. This work aims to present the first step of a methodology in progress based on researches in neuroscience to be applied to Brazilian science teachers in order to contribute to an effective teaching/learning process. This research will use the Implicit Association Test (IAT) to verify implicit attitudes of science teachers concerning nuclear energy. Results will provide data for the next steps of the research. The literature has not reported a similar neuroscience-based methodology applied to the nuclear energy learning/teaching process; therefore, this has demonstrated to be an innovating methodology. The development of the methodology is in progress and the results will be presented in future works. (author)

  7. A STUDY OF NUCLEAR PHYSICS PROCESSES AT MIDDLE SCHOOL

    Directory of Open Access Journals (Sweden)

    Mykola I. Sadovyi

    2011-02-01

    Full Text Available The article discloses the problem of new technology usage for the physics’ experiment in the quantum physics modeling. Currency of investigation consists in the need of physics experiment organization and realization in high energy physics with the consistent usage of activity method in middle education institutions. This kind of method considerably stirs up the process of model usage and modeling, abstracting, idealization and analogy. Idealized objects’ creation, elementary part transmutation, in particular, that does not exist in the objective reality, but possesses definite prototypes in the real world that help in their first approximation to the truth. The program Macromedia Flesh has been used in the article. This program has a range of advantages comparing to other possible software according to their possibilities and usage simplicity. The program uses all kinds of computer graph (raster, vectorial, which gives great opportunities for graphic objects’ creation, and prepared files take minimum of the constant memory. A part of developed experiments of the modeling character is given in the article. Demonstrations are done in dynamic rate.

  8. NUMATH: a nuclear-material-holdup estimator for unit operations and chemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Krichinsky, A.M.

    1983-02-01

    A computer program, NUMATH (Nuclear Material Holdup Estimator), has been developed to estimate compositions of materials in vessels involved in unit operations and chemical processes. This program has been implemented in a remotely operated nuclear fuel processing plant. NUMATH provides estimates of the steady-state composition of materials residing in process vessels until representative samples can be obtained and chemical analyses can be performed. Since these compositions are used for inventory estimations, the results are determined for the cataloged in container-oriented files. The estimated compositions represent materials collected in applicable vessels - including consideration for materials previously acknowledged in these vessels. The program utilizes process measurements and simple performance models to estimate material holdup and distribution within unit operations. In simulated run-testing, NUMATH typically produced estimates within 5% of the measured inventories for uranium and within 8% of the measured inventories for thorium during steady-state process operation.

  9. The impact of global nuclear mass model uncertainties on r-process abundance predictions

    Directory of Open Access Journals (Sweden)

    Mumpower M.

    2015-01-01

    Full Text Available Rapid neutron capture or ‘r-process’ nucleosynthesis may be responsible for half the production of heavy elements above iron on the periodic table. Masses are one of the most important nuclear physics ingredients that go into calculations of r-process nucleosynthesis as they enter into the calculations of reaction rates, decay rates, branching ratios and Q-values. We explore the impact of uncertainties in three nuclear mass models on r-process abundances by performing global monte carlo simulations. We show that root-mean-square (rms errors of current mass models are large so that current r-process predictions are insufficient in predicting features found in solar residuals and in r-process enhanced metal poor stars. We conclude that the reduction of global rms errors below 100 keV will allow for more robust r-process predictions.

  10. Study on the high-precision laser welding technology of nuclear fuel elements processing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Sung; Yang, M. S.; Kim, W. K.; Lee, D. Y

    2001-01-01

    The proper welding method for appendage of bearing pads and spacers of PHWR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of welding defects of the appendage parts is mostly apt to occur and it is connected directly with the safty and life prediction of the nuclear reactor in operation. Recently there has been studied all over the world to develope welding technology by laser in nuclear fuel processing, and the appendage of bearing pads and spacers of PHWR nuclear fuel elements. Therefore, the purpose of this study is to investigate the characteristics of the laser welded specimens and make some samples for the appendage of bearing pads of PHWR nuclear fuel elements. This study will be also provide the basic data for the fabrications of the appendage of bearing pads and spacers. Especially the laser welding is supposed to be used in the practical application such as precise materials manufacturing fields. In this respect this technology is not only a basic advanced technology with wide applications but also likely to be used for the development of directly applicable technologies for industries, with high potential benefits derived in the view point of economy and industry.

  11. A sensitivity study of s-process: the impact of uncertainties from nuclear reaction rates

    Science.gov (United States)

    Vinyoles, N.; Serenelli, A.

    2016-01-01

    The slow neutron capture process (s-process) is responsible for the production of about half the elements beyond the Fe-peak. The production sites and the conditions under which the different components of s-process occur are relatively well established. A detailed quantitative understanding of s-process nucleosynthesis may yield light in physical processes, e.g. convection and mixing, taking place in the production sites. For this, it is important that the impact of uncertainties in the nuclear physics is well understood. In this work we perform a study of the sensitivity of s-process nucleosynthesis, with particular emphasis in the main component, on the nuclear reaction rates. Our aims are: to quantify the current uncertainties in the production factors of s-process elements originating from nuclear physics and, to identify key nuclear reactions that require more precise experimental determinations. In this work we studied two different production sites in which s-process occurs with very different neutron exposures: 1) a low-mass extremely metal-poor star during the He-core flash (nn reaching up to values of ∼ 1014cm-3); 2) the TP-AGB phase of a M⊙, Z=0.01 model, the typical site of the main s-process component (nn up to 108 — 109cm-3). In the first case, the main variation in the production of s-process elements comes from the neutron poisons and with relative variations around 30%-50%. In the second, the neutron poison are not as important because of the higher metallicity of the star that actually acts as a seed and therefore, the final error of the abundances are much lower around 10%-25%.

  12. Presentation of the process External communications on the nuclear facilities operation of the Adjunct Head Office of Nuclear Safety of Comision Nacional de Seguridad Nuclear y Salvaguardias; Presentacion del proceso Comunicaciones externas sobre el funcionamiento de instalaciones nucleares de la Direccion General Adjunta de Seguridad Nuclear de la Comision Nacional de Seguridad Nuclear y Salvaguardias

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa V, J. M., E-mail: jmespinosa@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2012-10-15

    The Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) in use of their attributions granted by the Regulation Law of the constitutional Art. 27 in nuclear matter began the development of the called process External communications on the nuclear facilities operation, with the purpose of negotiating the evaluation of the concerns related with the safety of the nuclear facilities received these of external people to the CNSNS. The process External communications on the nuclear facilities operation will allow to the public's members and the workers that carry out activities inside the mark regulator imposed by the CNSNS that report to this Commission their concerns related with safety for several means (for example, directly to the personnel of the assigned Office, official and public statements, phone communication, electronic mail, etc.) The present article presents the legal mark confers the CNSNS the attributions to develop the mentioned process and exposes the most important elements that compose it. The term External communication on the nuclear facilities operation is defined and also is described how these communications are received, evaluated and closed by the assigned Office. Of equal way the objectives that intents to reach this process are indicated. The intention of the mentioned process is to strengthen the actions that the CNSNS carries out in the execution of its functions to maintain the safety standards in the operation of the nuclear facilities in Mexico. (Author)

  13. Laser-based analytical monitoring in nuclear-fuel processing plants

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, J.P.

    1978-09-01

    The use of laser-based analytical methods in nuclear-fuel processing plants is considered. The species and locations for accountability, process control, and effluent control measurements in the Coprocessing, Thorex, and reference Purex fuel processing operations are identified and the conventional analytical methods used for these measurements are summarized. The laser analytical methods based upon Raman, absorption, fluorescence, and nonlinear spectroscopy are reviewed and evaluated for their use in fuel processing plants. After a comparison of the capabilities of the laser-based and conventional analytical methods, the promising areas of application of the laser-based methods in fuel processing plants are identified.

  14. Participation of local citizens groups in the Swedish nuclear waste process

    Energy Technology Data Exchange (ETDEWEB)

    Holmstrand, O. [The Waste Network, Lerum (Sweden)

    1999-12-01

    The Waste Network's conclusions and views on the nuclear waste issue are summarised in the following points: The management of the nuclear waste is not solved. In order to minimise the amount of waste, the further operating of nuclear reactors should be questioned. The choice of method should be made before the siting. The deadlock to the KBS method must come to an end. The choice of method must be based on clearly expressed functional conditions formulated in advance. The siting must be based on considerations to environment and security, not political acceptance. The pilot studies in municipalities must cease and should be replaced by a clear and understandable sieving process at a national scale. An independent authority must control and supervise the EIA process instead of the nuclear industry. A well performed EIA process is the necessary condition to give the choice of method and site enough legitimacy and acceptance. Environmental organisations being the representatives of the public must be given reasonable conditions and resources to take part in the EIA process to handle the question.

  15. A computational tool for selective pyrochemical processes based on molten salts in nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Walle, Eric [EDF R and D - MMC, Site des Renardieres, Avenue des Renardieres, Ecuelles, 77818 Moret Sur Loing cedex (France)]. E-mail: eric.walle@edf.fr; Finne, Jorgen [LECA - CNRS - UMR 7575, Ecole Nationale Superieure de Chimie de Paris, 11, rue Pierre et Marie Curie, 75005 Paris (France); Picard, Gerard [LECA - CNRS - UMR 7575, Ecole Nationale Superieure de Chimie de Paris, 11, rue Pierre et Marie Curie, 75005 Paris (France); Sanchez, Sylvie [LECA - CNRS - UMR 7575, Ecole Nationale Superieure de Chimie de Paris, 11, rue Pierre et Marie Curie, 75005 Paris (France); Boursier, Jean-Marie [EDF R and D - MMC, Site des Renardieres, Avenue des Renardieres, Ecuelles, 77818 Moret Sur Loing cedex (France); Noel, Didier [EDF R and D - MMC, Site des Renardieres, Avenue des Renardieres, Ecuelles, 77818 Moret Sur Loing cedex (France)

    2005-09-01

    In the framework of the development of pyrochemical techniques for future nuclear systems, we propose to investigate the development of computational tools in order to optimise the control of chemical parameters for pyrochemical processes (electrodeposition, oxide selective precipitation, liquid/liquid reductive extraction). This paper discusses how potential-oxoacidity diagrams can be automatically constructed and gives illustrative examples on how these diagrams may help in the choice and the optimisation of pyrochemical processes.

  16. 活性炭脱硫富集烟气制酸系统技术改造实践%Technological revamping of sulphuric acid plant based on off-gas enriched by activated carbon desulphurization

    Institute of Scientific and Technical Information of China (English)

    李强

    2012-01-01

    The process and problems of the sulphuric acid plant based on a new 450 m2 sintering machine off-gas which is enriched by activated carbon desulphurization are introduced. Some improvements were made in accordance with lower gas temperature at spray tower inlet, insufficient bubble column height, high dust content at the inlet of drying tower, serious acid accumulation in acid distributor groove and darken product acid color. After retrofitting, the system operated well and the product acid became clear and transparent.%介绍了新450m^2烧结机烟气活性炭脱硫富集烟气制酸系统流程和存在问题。针对喷淋塔进口烟气温度偏低、泡沫柱高度不足、干燥塔进口尘含量超高、分酸槽内酸泥堆积严重、成品酸色度差等问题采取有效改造措施。改造后系统运行正常,产品酸清澈透明。

  17. Nuclear DNA replication initiation in kinetoplastid parasites: new insights into an ancient process.

    Science.gov (United States)

    Tiengwe, Calvin; Marques, Catarina A; McCulloch, Richard

    2014-01-01

    Nuclear DNA replication is, arguably, the central cellular process in eukaryotes, because it drives propagation of life and intersects with many other genome reactions. Perhaps surprisingly, our understanding of nuclear DNA replication in kinetoplastids was limited until a clutch of studies emerged recently, revealing new insight into both the machinery and genome-wide coordination of the reaction. Here, we discuss how these studies suggest that the earliest acting components of the kinetoplastid nuclear DNA replication machinery - the factors that demarcate sites of the replication initiation, termed origins - are diverged from model eukaryotes. In addition, we discuss how origin usage and replication dynamics relate to the highly unusual organisation of transcription in the genome of Trypanosoma brucei.

  18. Simulation modeling of nuclear steam generator water level process--a case study

    Science.gov (United States)

    Zhao; Ou; Du

    2000-01-01

    Simulation modeling of the nuclear steam generator (SG) water level process in Qinshan Nuclear Power Plant (QNPP) is described in this paper. A practical methodology was adopted so that the model is both simple and accurate for control engineering implementation. The structure of the model is in the form of a transfer function, which was determined based on first-principles analysis and expert experience. The parameters of the model were obtained by taking advantage of the recorded historical response curves under the existing closed-loop control system. The results of process dimensional data verification and experimental tests demonstrate that the simulation model depicts the main dynamic characteristics of the SG water level process and is in accordance with the field recorded response curves. The model has been successfully applied to the design and test of an advanced digital feedwater control system in QNPP.

  19. Processing of mixed uranium/refractory metal carbide fuels for high temperature space nuclear reactors

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim

    2000-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for high performance, next generation space power and propulsion systems. These mixed carbides such as the pseudo-ternary, (U, Zr, Nb)C, hold significant promise because of their high melting points (typically greater than 3200 K), thermochemical stability in a hot hydrogen environment, and high thermal conductivity. However, insufficient test data exist under nuclear thermal propulsion conditions of temperature and hot hydrogen environment to fully evaluate their performance. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders while hypostoichiometric samples with carbon-to-metal (C/M) ratios of 0.95 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold pressing, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce dense (low porosity), homogeneous, single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for space power and propulsion applications. .

  20. The modernization of the process computer of the Trillo Nuclear Power Plant; Modernizacion del ordenador de proceso de la Central Nuclear de Trillo

    Energy Technology Data Exchange (ETDEWEB)

    Martin Aparicio, J.; Atanasio, J.

    2011-07-01

    The paper describes the modernization of the Process computer of the Trillo Nuclear Power Plant. The process computer functions, have been incorporated in the non Safety I and C platform selected in Trillo NPP: the Siemens SPPA-T2000 OM690 (formerly known as Teleperm XP). The upgrade of the Human Machine Interface of the control room has been included in the project. The modernization project has followed the same development process used in the upgrade of the process computer of PWR German nuclear power plants. (Author)

  1. Seawater desalination plant using nuclear heating reactor coupled with MED process

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10~200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

  2. The r-process of stellar nucleosynthesis: Astrophysics and nuclear physics achievements and mysteries

    Science.gov (United States)

    Arnould, M.; Goriely, S.; Takahashi, K.

    2007-09-01

    The r-process, or the rapid neutron-capture process, of stellar nucleosynthesis is called for to explain the production of the stable (and some long-lived radioactive) neutron-rich nuclides heavier than iron that are observed in stars of various metallicities, as well as in the solar system. A very large amount of nuclear information is necessary in order to model the r-process. This concerns the static characteristics of a large variety of light to heavy nuclei between the valley of stability and the vicinity of the neutron-drip line, as well as their beta-decay branches or their reactivity. Fission probabilities of very neutron-rich actinides have also to be known in order to determine the most massive nuclei that have a chance to be involved in the r-process. Even the properties of asymmetric nuclear matter may enter the problem. The enormously challenging experimental and theoretical task imposed by all these requirements is reviewed, and the state-of-the-art development in the field is presented. Nuclear-physics-based and astrophysics-free r-process models of different levels of sophistication have been constructed over the years. We review their merits and their shortcomings. The ultimate goal of r-process studies is clearly to identify realistic sites for the development of the r-process. Here too, the challenge is enormous, and the solution still eludes us. For long, the core collapse supernova of massive stars has been envisioned as the privileged r-process location. We present a brief summary of the one- or multidimensional spherical or non-spherical explosion simulations available to-date. Their predictions are confronted with the requirements imposed to obtain an r-process. The possibility of r-nuclide synthesis during the decompression of the matter of neutron stars following their merging is also discussed. Given the uncertainties remaining on the astrophysical r-process site and on the involved nuclear physics, any confrontation between predicted r-process

  3. The r-process of stellar nucleosynthesis: Astrophysics and nuclear physics achievements and mysteries

    Energy Technology Data Exchange (ETDEWEB)

    Arnould, M. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, CP226, B-1050 Brussels (Belgium)], E-mail: marnould@astro.ulb.ac.be; Goriely, S.; Takahashi, K. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, CP226, B-1050 Brussels (Belgium)

    2007-09-15

    The r-process, or the rapid neutron-capture process, of stellar nucleosynthesis is called for to explain the production of the stable (and some long-lived radioactive) neutron-rich nuclides heavier than iron that are observed in stars of various metallicities, as well as in the solar system. A very large amount of nuclear information is necessary in order to model the r-process. This concerns the static characteristics of a large variety of light to heavy nuclei between the valley of stability and the vicinity of the neutron-drip line, as well as their beta-decay branches or their reactivity. Fission probabilities of very neutron-rich actinides have also to be known in order to determine the most massive nuclei that have a chance to be involved in the r-process. Even the properties of asymmetric nuclear matter may enter the problem. The enormously challenging experimental and theoretical task imposed by all these requirements is reviewed, and the state-of-the-art development in the field is presented. Nuclear-physics-based and astrophysics-free r-process models of different levels of sophistication have been constructed over the years. We review their merits and their shortcomings. The ultimate goal of r-process studies is clearly to identify realistic sites for the development of the r-process. Here too, the challenge is enormous, and the solution still eludes us. For long, the core collapse supernova of massive stars has been envisioned as the privileged r-process location. We present a brief summary of the one- or multidimensional spherical or non-spherical explosion simulations available to-date. Their predictions are confronted with the requirements imposed to obtain an r-process. The possibility of r-nuclide synthesis during the decompression of the matter of neutron stars following their merging is also discussed. Given the uncertainties remaining on the astrophysical r-process site and on the involved nuclear physics, any confrontation between predicted r-process

  4. Nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Sang, David (Bishop Luffa Comprehensive School, Chichester (UK))

    1990-01-01

    Nuclear Physics covers the aspects of radioactivity and nuclear physics dealt with in the syllabuses of all the A-level examination boards; in particular, it provides detailed coverage of the Joint Matriculation Board option in nuclear physics. It deals with the discovery of the atomic nucleus, the physics of nuclear processes, and nuclear technology. (author).

  5. Process for estimating likelihood and confidence in post detonation nuclear forensics.

    Energy Technology Data Exchange (ETDEWEB)

    Darby, John L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craft, Charles M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Technical nuclear forensics (TNF) must provide answers to questions of concern to the broader community, including an estimate of uncertainty. There is significant uncertainty associated with post-detonation TNF. The uncertainty consists of a great deal of epistemic (state of knowledge) as well as aleatory (random) uncertainty, and many of the variables of interest are linguistic (words) and not numeric. We provide a process by which TNF experts can structure their process for answering questions and provide an estimate of uncertainty. The process uses belief and plausibility, fuzzy sets, and approximate reasoning.

  6. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric T.; Canabal, Francisco

    A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze the entire process on a three-dimensional domain. The computed flammability at the exit of the heat exchanger was less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process.

  7. Dynamic flowgraph modeling of process and control systems of a nuclear-based hydrogen production plant

    Energy Technology Data Exchange (ETDEWEB)

    Al-Dabbagh, Ahmad W. [Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario (Canada); Lu, Lixuan [Faculty of Energy Systems and Nuclear Science, Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario (Canada)

    2010-09-15

    Modeling and analysis of system reliability facilitate the identification of areas of potential improvement. The Dynamic Flowgraph Methodology (DFM) is an emerging discrete modeling framework that allows for capturing time dependent behaviour, switching logic and multi-state representation of system components. The objective of this research is to demonstrate the process of dynamic flowgraph modeling of a nuclear-based hydrogen production plant with the copper-chlorine (Cu-Cl) cycle. Modeling of the thermochemical process of the Cu-Cl cycle in conjunction with a networked control system proposed for monitoring and control of the process is provided. This forms the basis for future component selection. (author)

  8. Excitation of the isomeric ^{229m}Th nuclear state via an electronic bridge process in ^{229}Th^+

    CERN Document Server

    Porsev, S G; Peik, E; Tamm, Chr

    2010-01-01

    We consider the excitation of the nuclear transition ^{229g}Th - ^{229m}Th near 7.6 eV in singly ionized thorium via an electronic bridge process. The process relies on the excitation of the electron shell by two laser photons whose sum frequency is equal to the nuclear transition frequency. This scheme allows to determine the nuclear transition frequency with high accuracy. Based on calculations of the electronic level structure of Th^+ which combine the configuration-interaction method and many-body perturbation theory, we estimate that a nuclear excitation rate in the range of 10 s^{-1} can be obtained using conventional laser sources.

  9. The p-process of stellar nucleosynthesis: astrophysics and nuclear physics status

    Science.gov (United States)

    Arnould, M.; Goriely, S.

    2003-09-01

    The p-process of stellar nucleosynthesis is aimed at explaining the production of the stable neutron-deficient nuclides heavier than iron that are observed up to now in the solar system exclusively. Various scenarios have been proposed to account for the bulk p-nuclide content of the solar system, as well as for deviations (`anomalies') with respect to the bulk p-isotope composition of some elements discovered in primitive meteorites. The astrophysics of these models is reviewed, and the involved nuclear physics is discussed, including a brief account of recent experimental efforts. Already published results are complemented with new ones. A specific attention is paid to the very rare odd-odd nuclides 138La and 180Tam, as well as to the puzzling case of the light Mo and Ru isotopes. Astrophysics and nuclear physics prospects of improvements in the p-process modeling are presented.

  10. The p-process of stellar nucleosynthesis: astrophysics and nuclear physics status

    Energy Technology Data Exchange (ETDEWEB)

    Arnould, M.; Goriely, S

    2003-09-01

    The p-process of stellar nucleosynthesis is aimed at explaining the production of the stable neutron-deficient nuclides heavier than iron that are observed up to now in the solar system exclusively. Various scenarios have been proposed to account for the bulk p-nuclide content of the solar system, as well as for deviations ('anomalies') with respect to the bulk p-isotope composition of some elements discovered in primitive meteorites. The astrophysics of these models is reviewed, and the involved nuclear physics is discussed, including a brief account of recent experimental efforts. Already published results are complemented with new ones. A specific attention is paid to the very rare odd-odd nuclides {sup 138}La and {sup 180}Ta{sup m}, as well as to the puzzling case of the light Mo and Ru isotopes. Astrophysics and nuclear physics prospects of improvements in the p-process modeling are presented.

  11. Measurement of the energy spectra of fission fragments using nuclear track detectors and digital image processing

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, Guillermo; Golzarri, Jose I. [Instituto de Fisica, Universidad Nacional Autonoma de Mexico, A.P. 20-364, Mexico D.F. 01000 (Mexico); Castano, Victor M., E-mail: castano@fata.unam.m [Centro de Fisica Aplicada y Tecnologia Avanzada, Universidad Nacional Autonoma de Mexico, Boulevard Juriquilla 3001, Santiago de Queretaro, Queretaro 76230 (Mexico)

    2010-08-15

    Energy spectra of fission fragments were determined using a Nuclear Track Methodology (NTM) supported by digital image analysis and numerical data processing using a standard personal computer. The analysis of a californium ({sup 252}Cf) spectrum with this approach shows improvement compared with the values reported previously using the standard procedure, in terms of resolution and accuracy. This new method adds full automation to the technical advantages and cost effectiveness of an NTM.

  12. Lanthanides extraction processes in molten fluoride media. Application to nuclear spent fuel reprocessing

    OpenAIRE

    Taxil, Pierre; Massot, Laurent; Nourry, Christophe; Gibilaro, Mathieu; Chamelot, Pierre; Cassayre, Laurent

    2009-01-01

    This paper describes four techniques of extraction of lanthanides elements (Ln) from molten salts in the general frame of reprocessing nuclear wastes; One of them is chemical: the precipitation of Ln ions in insoluble compounds (oxides or oxifluorides); the others use electrochemical methodology in molten fluorides for extraction and measurement of the progress of the processes: first electrodeposition of pure Ln metals on an inert cathode material was proved to be incomplete and cause probl...

  13. The impact of nuclear mass models on r-process nucleosynthesis network calculations

    Science.gov (United States)

    Vaughan, Kelly

    2002-10-01

    An insight into understanding various nucleosynthesis processes is via modelling of the process with network calculations. My project focus is r-process network calculations where the r-process is nucleosynthesis via rapid neutron capture thought to take place in high entropy supernova bubbles. One of the main uncertainties of the simulations is the Nuclear Physics input. My project investigates the role that nuclear masses play in the resulting abundances. The code tecode, involves rapid (n,γ) capture reactions in competition with photodisintegration and β decay onto seed nuclei. In order to fully analyze the effects of nuclear mass models on the relative isotopic abundances, calculations were done from the network code, keeping the initial environmental parameters constant throughout. The supernova model investigated by Qian et al (1996) in which two r-processes, of high and low frequency with seed nucleus ^90Se and of fixed luminosity (fracL_ν_e(0)r_7(0)^2 ˜= 8.77), contribute to the nucleosynthesis of the heavier elements. These two r-processes, however, do not contribute equally to the total abundance observed. The total isotopic abundance produced from both events was therefore calculated using equation refabund. Y(H+L) = fracY(H)+fY(L)f+1 applicability of the P-Scheme in relation to the other mass models to the r-process network calculations. 02 Pscheme Aprahamian,A., Gadala-Maria,A. & Cuka,N. 1996, Revista Mexicana de Fisica,42,1 code Surman,R. & Engel,J. 1998, Phys.Rev. C,54,4 thebibliography

  14. The evolutionary processes of mitochondrial and chloroplast genomes differ from those of nuclear genomes

    Science.gov (United States)

    Korpelainen, Helena

    2004-11-01

    This paper first introduces our present knowledge of the origin of mitochondria and chloroplasts, and the organization and inheritance patterns of their genomes, and then carries on to review the evolutionary processes influencing mitochondrial and chloroplast genomes. The differences in evolutionary phenomena between the nuclear and cytoplasmic genomes are highlighted. It is emphasized that varying inheritance patterns and copy numbers among different types of genomes, and the potential advantage achieved through the transfer of many cytoplasmic genes to the nucleus, have important implications for the evolution of nuclear, mitochondrial and chloroplast genomes. Cytoplasmic genes transferred to the nucleus have joined the more strictly controlled genetic system of the nuclear genome, including also sexual recombination, while genes retained within the cytoplasmic organelles can be involved in selection and drift processes both within and among individuals. Within-individual processes can be either intra- or intercellular. In the case of heteroplasmy, which is attributed to mutations or biparental inheritance, within-individual selection on cytoplasmic DNA may provide a mechanism by which the organism can adapt rapidly. The inheritance of cytoplasmic genomes is not universally maternal. The presence of a range of inheritance patterns indicates that different strategies have been adopted by different organisms. On the other hand, the variability occasionally observed in the inheritance mechanisms of cytoplasmic genomes reduces heritability and increases environmental components in phenotypic features and, consequently, decreases the potential for adaptive evolution.

  15. A statistical approach for identifying nuclear waste glass compositions that will meet quality and processability requirements

    Energy Technology Data Exchange (ETDEWEB)

    Piepel, G.F.

    1990-09-01

    Borosilicate glass provides a solid, stable medium for the disposal of high-level radioactive wastes resulting from the production of nuclear materials for United States defense needs. The glass must satisfy various quality and processability requirements on properties such as chemical durability, viscosity, and electrical conductivity. These properties depend on the composition of the waste glass, which will vary during production due to variations in nuclear waste composition and variations in the glass-making process. This paper discusses the experimentally-based statistical approach being used in the Hanford Waste Vitrification Plant (HWVP) Composition Variability Study (CVS). The overall goal of the CVS is to identify the composition region of potential HWVP waste glasses that satisfy with high confidence the applicable quality and processability requirements. This is being accomplished by melting and obtaining property data for simulated nuclear waste glasses of various compositions, and then statistically developing models and other tools needed to meet the goal. 6 refs., 1 fig., 5 tabs.

  16. Beta processes in a high-temperature field and nuclear multibeta decays

    Energy Technology Data Exchange (ETDEWEB)

    Kopytin, I. V., E-mail: kopytin@yandex.ru; Hussain, Imad A. [Voronezh State University (Russian Federation)

    2013-11-15

    Sources of the temperature dependence of rates of nuclear beta processes in matter of massive stars are systematized. Electron and positron beta decays and electron capture (K capture and the capture of unbound electrons) fromexcited nuclear states (thermal decays) are considered along with the photobeta decays from ground and excited nuclear states. The possible quantum degeneracy of an electron gas in matter and the degree of ionization of an atomic K shell in a high-temperature field are taken into account. For a number of multidecay odd-nuclei, the temperature dependences of the ratios of the total rates of their {beta}{sup -} decays to the sum of the total rates over all of decay modes for the same nuclei are calculated in the range of nuclear temperature from 2 to 3 Multiplication-Sign 10{sup 9} K. It is shown that the deviation of this ratio from the experimental value obtained at 'normal' temperature may be quite sizable. This circumstance should be taken into account in models that consider the problem of synthesis of nuclei in matter of massive stars.

  17. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  18. Energetic electron processes fluorescence effects for structured nanoparticles X-ray analysis and nuclear medicine applications

    Energy Technology Data Exchange (ETDEWEB)

    Taborda, A.; Desbrée, A. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-HOM/SDI/LEDI, BP-17, 31, Avenue de la Division Leclerc, 92262 Fontenay-aux-Roses (France); Carvalho, A. [IEQUALTECS, Lda, Rua Dr. Francisco Sá Carneiro, 36, 2500-065 S. Gregório CLD (Portugal); Chaves, P.C. [C" 2TN, Campus Tecnológico e Nuclear, Instituto Superior Técnico, Universidade de Lisboa, EN10 km 139.7, 2685-066 Bobadela LRS (Portugal); Reis, M.A., E-mail: mareis@ctn.tecnico.ulisboa.pt [IEQUALTECS, Lda, Rua Dr. Francisco Sá Carneiro, 36, 2500-065 S. Gregório CLD (Portugal); C" 2TN, Campus Tecnológico e Nuclear, Instituto Superior Técnico, Universidade de Lisboa, EN10 km 139.7, 2685-066 Bobadela LRS (Portugal)

    2016-08-15

    Superparamagnetic iron oxide (SPIO) nanoparticles are widely used as contrast agents for nuclear magnetic resonance imaging (MRI), and can be modified for improved imaging or to become tissue-specific or even protein-specific. The knowledge of their detailed elemental composition characterisation and potential use in nuclear medicine applications, is, therefore, an important issue. X-ray fluorescence techniques such as particle induced X-ray emission (PIXE) or X-ray fluorescence spectrometry (XRF), can be used for elemental characterisation even in problematic situations where very little sample volume is available. Still, the fluorescence coefficient of Fe is such that, during the decay of the inner-shell ionised atomic structure, keV Auger electrons are produced in excess to X-rays. Since cross-sections for ionisation induced by keV electrons, for low atomic number atoms, are of the order of 10{sup 3} barn, care should be taken to account for possible fluorescence effects caused by Auger electrons, which may lead to the wrong quantification of elements having atomic number lower than the atomic number of Fe. Furthermore, the same electron processes will occur in iron oxide nanoparticles containing {sup 57}Co, which may be used for nuclear medicine therapy purposes. In the present work, simple approximation algorithms are proposed for the quantitative description of radiative and non-radiative processes associated with Auger electrons cascades. The effects on analytical processes and nuclear medicine applications are quantified for the case of iron oxide nanoparticles, by calculating both electron fluorescence emissions and energy deposition on cell tissues where the nanoparticles may be embedded.

  19. Energetic electron processes fluorescence effects for structured nanoparticles X-ray analysis and nuclear medicine applications

    Science.gov (United States)

    Taborda, A.; Desbrée, A.; Carvalho, A.; Chaves, P. C.; Reis, M. A.

    2016-08-01

    Superparamagnetic iron oxide (SPIO) nanoparticles are widely used as contrast agents for nuclear magnetic resonance imaging (MRI), and can be modified for improved imaging or to become tissue-specific or even protein-specific. The knowledge of their detailed elemental composition characterisation and potential use in nuclear medicine applications, is, therefore, an important issue. X-ray fluorescence techniques such as particle induced X-ray emission (PIXE) or X-ray fluorescence spectrometry (XRF), can be used for elemental characterisation even in problematic situations where very little sample volume is available. Still, the fluorescence coefficient of Fe is such that, during the decay of the inner-shell ionised atomic structure, keV Auger electrons are produced in excess to X-rays. Since cross-sections for ionisation induced by keV electrons, for low atomic number atoms, are of the order of 103 barn, care should be taken to account for possible fluorescence effects caused by Auger electrons, which may lead to the wrong quantification of elements having atomic number lower than the atomic number of Fe. Furthermore, the same electron processes will occur in iron oxide nanoparticles containing 57Co, which may be used for nuclear medicine therapy purposes. In the present work, simple approximation algorithms are proposed for the quantitative description of radiative and non-radiative processes associated with Auger electrons cascades. The effects on analytical processes and nuclear medicine applications are quantified for the case of iron oxide nanoparticles, by calculating both electron fluorescence emissions and energy deposition on cell tissues where the nanoparticles may be embedded.

  20. Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

    Directory of Open Access Journals (Sweden)

    Eun-Young Choi

    2015-12-01

    Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

  1. Recent advances of annular centrifugal extractor for hot test of nuclear waste partitioning process

    Institute of Scientific and Technical Information of China (English)

    HeXiang-Ming; YanYu-Shun; 等

    1998-01-01

    Advances are being made in the design of the annular centrifugal extractor fornuclear fuel reprocessing extraction process studies.The extractors have been built and tested.Twelve stages of this extractor and 50 stages are used toimplement the TRPO process for the cleanup ofcommercial and defense nuclear waste liquids,respectively.Following advances are available:(1) simple way of assembly and disassembly between rotor part and housing part of extractor,ease of manipulator operation;(2)automatic sampling from housing of extractor in hot cell;(3) compact multi-stage housing system;(4) easy interstage link;(5) computer data acquisition and monitoring system of speed.

  2. Advances in processing technologies for titanium heat exchanger tubes of fossil and nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Likhareva, T.P.; Tchizhik, A.A.; Chavchanidze, N.N. [Polzanov Central Boiler and Turbine Institute, St. Petersburg (Russian Federation)

    1998-12-31

    The advances in processing technologies for titanium heat exchangers with rolled and welded tubes of fossil and nuclear power plants in Russia are presented. The special methodology of investigations with constant small strain rate have been used to study the effects of mixed corrosion and creep processes in condensers cooled by sea or synthetic sea waters. The results of corrosion creep tests and K1scc calculations are given. The Russian science activities concerning condensers manufactured from titanium show the possibilities for designing structures with very high level service reliability in different corrosion aggressive mediums with high total salt, Cl-ion and oxygen contents. (orig.)

  3. The Role of Fe and Ni for S-process Nucleosynthesis and Innovative Nuclear Technologies

    CERN Document Server

    Giubrone, G; Perkowski, J; Andriamonje, S; Carrapico, C; Wallner, A; Vannini, G; Quesada, J M; Lederer, C; Tarrio, D; Berthier, B; Lozano, M; Krticka, M; Domingo-Pardo, C; Chiaveri, E; Jericha, E; Ferrari, A; Massimi, C; Avrigeanu, V; Martinez, T; Guerrero, C; Andrzejewski, J; Karadimos, D; Mendoza, E; Ganesan, S; Vlachoudis, V; Milazzo, P M; Cortes, G; Becares, V; Tain, J L; Variale, V; Quinones, J; Calvino, F; Kappeler, F; Gunsing, F; Gramegna, F; Colonna, N; Marrone, S; Lebbos, E; Paradela, C; Mastinu, P F; Vaz, P; Tassan-Got, L; Kadi, Y; Dillman, I; Cano-Ott, D; Brugger, M; Audouin, L; Fernandez-Ordonez, M; Sarmento, R; Becvar, F; Goncalves, I F; Martin-Fuertes, F; Cerutti, F; Pina, G; Mosconi, M; Tagliente, G; Duran, I; Berthoumieux, E; Praena, J; Ioannides, K; Weiss, C; Mirea, M; Gomez-Hornillos, M B; Vlastou, R; Calviani, M; Nolte, R; Mengoni, A; Gonzalez-Romero, E; Marganiec, J; Leeb, H; Heil, M; Meaze, M H; Pavlik, A; Belloni, F; Harrispopulos S

    2011-01-01

    The accurate measurement of neutron capture cross sections of all Fe and Ni isotopes is important for disentangling the contribution of the s-process and the r-process to the stellar nucleosynthesis of elements in the mass range 60 < A < 120. At the same time, Fe and Ni are important components of structural materials and improved neutron cross section data is relevant in the design of new nuclear systems. With the aim of obtaining improved capture data on all stable iron and nickel isotopes, a program of measurements has been launched at the CERN Neutron Time of Flight Facility n_TOF.

  4. FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200

  5. User input verification and test driven development in the NJOY21 nuclear data processing code

    Energy Technology Data Exchange (ETDEWEB)

    Trainer, Amelia Jo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCartney, Austin Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-21

    Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, and capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.

  6. Potential Signatures of Semi-volatile Compounds Associated With Nuclear Processing

    Energy Technology Data Exchange (ETDEWEB)

    Probasco, Kathleen M.; Birnbaum, Jerome C.; Maughan, A. D.

    2002-06-01

    Semi-volatile chemicals associated with nuclear processes (e.g., the reprocessing of uranium to produce plutonium for nuclear weapons, or the separation of actinides from processing waste streams), can provide sticky residues or signatures that will attach to piping, ducting, soil, water, or other surface media. Volatile compounds, that are more suitable for electro-optical sensing, have been well studied. However, the semi-volatile compounds have not been well documented or studied. A majority of these semi-volatile chemicals are more robust than typical gaseous or liquid chemicals and can have lifetimes of several weeks, months, or years in the environment. However, large data gaps exist concerning these potential signature compounds and more research is needed to fill these data gaps so that important signature information is not overlooked or discarded. This report investigates key semi-volatile compounds associated with nuclear separations, identifies available chemical and physical properties, and discusses the degradation products that would result from hydrolysis, radiolysis and oxidation reactions on these compounds.

  7. Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

    2005-01-01

    Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

  8. Hans A. Bethe Prize: Astrophysical, observational and nuclear-physics aspects of r-process nucleosynthesis

    Science.gov (United States)

    Kratz, Karl-Ludwig

    2014-03-01

    Guided by the Solar System (S.S.) abundance peaks at A ~= 130 and A ~= 195, the basic mechanisms for the rapid neutron-capture process (the r-process) have been known for over 50 years. However, even today, all proposed scenarios and sites face problems with astrophysical conditions as well as with the necessary nuclear-physics input. In my talk, I will describe efforts in experimental and theoretical nuclear-structure data for modeling today's three groups of r-process ``observables'', i.e. the bulk S.S. isotopic abundances, the elemental abundances in metal-poor halo stars, and peculiar isotopic patterns measured in certain cosmic stardust grains. To set a historical basis, I will briefly recall our site-independent ``waiting-point'' model, with superpositions of neutron-density components and the use of the first global, unified nuclear input based on the mass model FRDM(1992). This approach provided a considerable leap forward in the basic understanding of the required astrophysical conditions, as well as of specific shell-structure properties far from stability. Starting in the early millenium, the above simple model has been replaced by more realistic, dynamical parameter studies within the high-entropy wind scenario of core-collapse supernovae, now with superpositions of entropy (S) and electron-fraction (Ye) components. Furthermore, an improved, global set of nuclear-physics data is used today, based on the new mass model FRDM(2012). With this nuclear and astrophysics parameter combination, a new fit to the S.S. r-abundances will be shown, and its improvements and remaining deficiencies in terms of underlying shell structure will be discussed. Concerning the abundance patterns in metal-poor halo stars, an interpretation of the production of ``r-rich'' (e.g. CS 22892-052) and ``r-poor'' (e.g. HD 122563) stars in terms of different (Ye), S combinations will be presented. Finally, for the third group of ``r-observables'', a possible origin of the anomalous Xe

  9. Program Management at the National Nuclear Security Administration Office of Defense Nuclear Security: A Review of Program Management Documents and Underlying Processes

    Energy Technology Data Exchange (ETDEWEB)

    Madden, Michael S.

    2010-05-01

    The scope of this paper is to review the National Nuclear Security Administration Office of Defense Nuclear Security (DNS) program management documents and to examine the underlying processes. The purpose is to identify recommendations for improvement and to influence the rewrite of the DNS Program Management Plan (PMP) and the documentation supporting it. As a part of this process, over 40 documents required by DNS or its stakeholders were reviewed. In addition, approximately 12 other documents produced outside of DNS and its stakeholders were reviewed in an effort to identify best practices. The complete list of documents reviewed is provided as an attachment to this paper.

  10. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  11. Stress-induced Nuclear Bodies Are Sites of Accumulation of Pre-mRNA Processing Factors

    Science.gov (United States)

    Denegri, Marco; Chiodi, Ilaria; Corioni, Margherita; Cobianchi, Fabio; Riva, Silvano; Biamonti, Giuseppe

    2001-01-01

    Heterogeneous nuclear ribonucleoprotein (hnRNP) HAP (hnRNP A1 interacting protein) is a multifunctional protein with roles in RNA metabolism, transcription, and nuclear structure. After stress treatments, HAP is recruited to a small number of nuclear bodies, usually adjacent to the nucleoli, which consist of clusters of perichromatin granules and are depots of transcripts synthesized before stress. In this article we show that HAP bodies are sites of accumulation for a subset of RNA processing factors and are related to Sam68 nuclear bodies (SNBs) detectable in unstressed cells. Indeed, HAP and Sam68 are both present in SNBs and in HAP bodies, that we rename “stress-induced SNBs.” The determinants required for the redistribution of HAP lie between residue 580 and 788. Different portions of this region direct the recruitment of the green fluorescent protein to stress-induced SNBs, suggesting an interaction of HAP with different components of the bodies. With the use of the 580–725 region as bait in a two-hybrid screening, we have selected SRp30c and 9G8, two members of the SR family of splicing factors. Splicing factors are differentially affected by heat shock: SRp30c and SF2/ASF are efficiently recruited to stress-induced SNBs, whereas the distribution of SC35 is not perturbed. We propose that the differential sequestration of splicing factors could affect processing of specific transcripts. Accordingly, the formation of stress-induced SNBs is accompanied by a change in the splicing pattern of the adenovirus E1A transcripts. PMID:11694584

  12. Examining Quality Management Audits in Nuclear Medicine Practice as a lifelong learning process: opportunities and challenges to the nuclear medicine professional and beyond.

    Science.gov (United States)

    Pascual, Thomas N B

    2016-08-01

    This essay will explore the critical issues and challenges surrounding lifelong learning for professionals, initially exploring within the profession and organizational context of nuclear medicine practice. It will critically examine how the peer-review process called Quality Management Audits in Nuclear Medicine Practice (QUANUM) of the International Atomic Energy Agency (IAEA) can be considered a lifelong learning opportunity to instill a culture of quality to improve patient care and elevate the status of the nuclear medicine profession and practice within the demands of social changes, policy, and globalization. This will be explored initially by providing contextual background to the identity of the IAEA as an organization responsible for nuclear medicine professionals, followed by the benefits that QUANUM can offer. Further key debates surrounding lifelong learning, such as compulsification of lifelong learning and impact on professional change, will then be weaved through the discussion using theoretical grounding through a qualitative review of the literature. Keeping in mind that there is very limited literature focusing on the implications of QUANUM as a lifelong learning process for nuclear medicine professionals, this essay uses select narratives and observations of QUANUM as a lifelong learning process from an auditor's perspective and will further provide a comparative perspective of QUANUM on the basis of other lifelong learning opportunities such as continuing professional development activities and observe parallelisms on its benefits and challenges that it will offer to other professionals in other medical speciality fields and in the teaching profession.

  13. Contaminants of the bismuth phosphate process as signifiers of nuclear reprocessing history.

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Sweet, Lucas E.

    2012-10-01

    Reagents used in spent nuclear fuel recycling impart unique contaminant patterns into the product stream of the process. Efforts are underway at Pacific Northwest National Laboratory to characterize and understand the relationship between these patterns and the process that created them. A main challenge to this effort, recycling processes that were employed at the Hanford site from 1944-1989 have been retired for decades. This precludes direct measurements of the contaminant patterns that propagate within product streams of these facilities. In the absence of any operating recycling facilities at Hanford, we have taken a multipronged approach to cataloging contaminants of U.S. reprocessing activities using: (1) historical records summarizing contaminants within the final Pu metal button product of these facilities; (2) samples of opportunity that represent intermediate products of these processes; and (3) lab-scale experiments and model simulations designed to replicate contaminant patterns at each stage of nuclear fuel reprocessing. This report provides a summary of the progress and results from Fiscal Year (April 1, 2010-September 30) 2011.

  14. Medical Image Processing Server applied to Quality Control of Nuclear Medicine.

    Science.gov (United States)

    Vergara, C.; Graffigna, J. P.; Marino, E.; Omati, S.; Holleywell, P.

    2016-04-01

    This paper is framed within the area of medical image processing and aims to present the process of installation, configuration and implementation of a processing server of medical images (MIPS) in the Fundación Escuela de Medicina Nuclear located in Mendoza, Argentina (FUESMEN). It has been developed in the Gabinete de Tecnologia Médica (GA.TE.ME), Facultad de Ingeniería-Universidad Nacional de San Juan. MIPS is a software that using the DICOM standard, can receive medical imaging studies of different modalities or viewing stations, then it executes algorithms and finally returns the results to other devices. To achieve the objectives previously mentioned, preliminary tests were conducted in the laboratory. More over, tools were remotely installed in clinical enviroment. The appropiate protocols for setting up and using them in different services were established once defined those suitable algorithms. Finally, it’s important to focus on the implementation and training that is provided in FUESMEN, using nuclear medicine quality control processes. Results on implementation are exposed in this work.

  15. The r-process of stellar nucleosynthesis: Astrophysics and nuclear physics achievements and mysteries

    CERN Document Server

    Arnould, M; Takahashi, K

    2007-01-01

    The r-process, or the rapid neutron-capture process, of stellar nucleosynthesis is called for to explain the production of the stable (and some long-lived radioactive) neutron-rich nuclides heavier than iron that are observed in stars of various metallicities, as well as in the solar system. A very large amount of nuclear information is necessary in order to model the r-process. This concerns the static characteristics of a large variety of light to heavy nuclei between the valley of stability and the vicinity of the neutron-drip line, as well as their beta-decay branches or their reactivity. The enormously challenging experimental and theoretical task imposed by all these requirements is reviewed, and the state-of-the-art development in the field is presented. Nuclear-physics-based and astrophysics-free r-process models of different levels of sophistication have been constructed over the years. We review their merits and their shortcomings. For long, the core collapse supernova of massive stars has been envi...

  16. Nitrification inhibition by hexavalent chromium Cr(VI)--Microbial ecology, gene expression and off-gas emissions.

    Science.gov (United States)

    Kim, Young Mo; Park, Hongkeun; Chandran, Kartik

    2016-04-01

    The goal of this study was to investigate the responses in the physiology, microbial ecology and gene expression of nitrifying bacteria to imposition of and recovery from Cr(VI) loading in a lab-scale nitrification bioreactor. Exposure to Cr(VI) in the reactor strongly inhibited nitrification performance resulting in a parallel decrease in nitrate production and ammonia consumption. Cr(VI) exposure also led to an overall decrease in total bacterial concentrations in the reactor. However, the fraction of ammonia oxidizing bacteria (AOB) decreased to a greater extent than the fraction of nitrite oxidizing bacteria (NOB). In terms of functional gene expression, a rapid decrease in the transcript concentrations of amoA gene coding for ammonia oxidation in AOB was observed in response to the Cr(VI) shock. In contrast, transcript concentrations of the nxrA gene coding for nitrite oxidation in NOB were relatively unchanged compared to Cr(VI) pre-exposure levels. Therefore, Cr(VI) exposure selectively and directly inhibited activity of AOB, which indirectly resulted in substrate (nitrite) limitation to NOB. Significantly, trends in amoA expression preceded performance trends both during imposition of and recovery from inhibition. During recovery from the Cr(VI) shock, the high ammonia concentrations in the bioreactor resulted in an irreversible shift towards AOB populations, which are expected to be more competitive in high ammonia environments. An inadvertent impact during recovery was increased emission of nitrous oxide (N2O) and nitric oxide (NO), consistent with recent findings linking AOB activity and the production of these gases. Therefore, Cr(VI) exposure elicited multiple responses on the microbial ecology, gene expression and both aqueous and gaseous nitrogenous conversion in a nitrification process. A complementary interrogation of these multiple responses facilitated an understanding of both direct and indirect inhibitory impacts on nitrification.

  17. FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200

  18. 醇解工段尾气系统技术改造%The technical improvement of alcoholysis work sections off- gas system

    Institute of Scientific and Technical Information of China (English)

    邹耀; 周模勇

    2012-01-01

    To avoid organic superscalar in off - gas from alcoholysis work section polluting air environment, the offgas should be emptied after absorbing methanol, methyl acetate, acetaldehyde, using idle equipment. Thenabsorption liquid was returned to recovery section to recover methanol, methyl acetate and acetaldehyde. The improvement not only lowered methanol'sconsumption, raised acetic acid's recovery, and also ensured emptied offgas reaching the class B State discharge standard.%为了避免醇解工段放空尾气中有机物含量超标对大气环境造成的污染,利用闲置设备,用水将尾气中甲醇、醋酸甲酯、乙醛吸收后,再放空。吸收液送回收工段处理回收甲醇、醋酸甲酯、乙醛。改造不仅降低甲醇消耗、提高醋酸回收率,而且确保放空尾气达到国家大气二级排放标准。

  19. Studies of nuclear processes at the Triangle Universities Nuclear Laboratory. Progress report, 1 September 1994--31 August 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, E.J.

    1995-09-01

    The Triangle Universities Nuclear Laboratory (TUNL)--a collaboration of Duke University, North Carolina State University, and the University of North Carolina at Chapel Hill--has had a very productive year. This report covers the second year of a three-year grant between the US Department of Energy and the three collaborating universities. The TUNL research program focuses on the following areas of nuclear physics: parity violation in neutron and charged-particle resonances--the mass and energy dependence of the weak interaction spreading width; chaotic behavior in {sup 30}P from studies of eigenvalue fluctuations in nuclear level schemes; studies of few-body systems; nuclear astrophysics; nuclear data evaluation for A = 3--20, for which TUNL is now the international center; high-spin spectroscopy and superdeformation in nuclei, involving collaborations at Argonne National Laboratory. Developments in technology and instrumentation have been vital to the research and training program. In this progress report the author describes: a proposed polarized {gamma}-beam facility at the Duke Free Electron Laser Laboratory; cryogenic systems and microcalorimeter development; continuing development of the Low Energy Beam Facility. The research summaries presented in this progress report are preliminary.

  20. Studies of nuclear processes at the Triangle Universities Nuclear Laboratory. Progress report, 1 September 1995--31 August 1996

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, E.J.

    1996-09-01

    The Triangle Universities Nuclear Laboratory (TUNL)--a collaboration of Duke University, North Carolina State University, and the University of North Carolina at Chapel Hill--has had a very productive year. This report covers parts of the second and third year of a three-year grant between the US Department of Energy and the three collaborating universities. The TUNL research program focuses on the following areas: precision test of parity-invariance violation in resonance neutron scattering at LANSCE/LANL; parity violation measurements using charged-particle resonances in A = 20--40 targets and the A = 4 system at TUNL; chaotic behavior in the nuclei {sup 30}P and {sup 34}Cl from studies of eigenvalue fluctuations in nuclear level schemes; search for anomalies in the level density (pairing phase transition) in 1f-2p shell nuclei using GEANIE at LANSCE/LANL; parity-conserving time-reversal noninvariance tests using {sup 166}Ho resonances at Geel, ORELA, or LANSCE/LANL; nuclear astrophysics; few-body nuclear systems; Nuclear Data evaluation for A = 3--20 for which TUNL is now the international center. Developments in technology and instrumentation are vital to the research and training program. Innovative work was continued in: polarized beam development; polarized target development; designing new cryogenic systems; designing new detectors; improving high-resolution beams for the KN and FN accelerators; development of an unpolarized Low-Energy Beam Facility for radiative capture studies of astrophysical interest. Preliminary research summaries are presented.

  1. The Space Nuclear Thermal Propulsion Program Results of the Environmental Impact Analysis Process

    Science.gov (United States)

    Harmon, Charles D.; Kristensen, David H.; McCulloch, William H.

    1994-07-01

    The Space Nuclear Thermal Propulsion (SNTP) Program initiated an environmental impact analysis process (EIAP) in March of 1992 to support design and construction of a nuclear thermal rocket engine ground testing facility. The `` Notice of Intent'' appeared in the Federal Register on March 12, 1992 and Scoping Meetings occurred during April 1992. The Draft Environmental Impact Statement (EIS) was publicly available on August 21, 1993 and public hearings were conducted during September 1992. Comments were resolved and the Final EIS `` Notice of Availability'' appeared in the Federal Register on May 14, 1993. Although program termination negated the need for a Record of Decision, completion of this EIAP demonstrates that the National Environmental Policy Act (NEPA) provides an adequate framework for involving the general public in all governmental decisions irrespective of either the technical complexities or the potential for nonacceptance. This paper discusses the SNTP EIAP and the associated analyses which indicated that ground testing of nuclear rocket engine concepts could be accomplished without significantly affecting the surrounding environment.

  2. Staffing decision processes and issues: Case studies of seven US Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Melber, B.; Roussel, A.; Baker, K.; Durbin, N.; Hunt, P.; Hauth, J.; Forslund, C.; Terrill, E. [Battelle Human Affairs Research Centers, Seattle, WA (United States); Gore, B. [Pacific Northwest Lab., Richland, WA (United States)

    1994-03-01

    The objective of this report is to identify how decisions are made regarding staffing levels and positions for a sample of U.S. nuclear power plants. In this report, a framework is provided for understanding the major forces driving staffing and the implications of staffing decisions for plant safety. The focus of this report is on driving forces that have led to changes in staffing levels and to the establishment of new positions between the mid-1980s and the early 1990s. Processes used at utilities and nuclear power plants to make and implement these staffing decisions are also discussed in the report. While general trends affecting the plant as a whole are presented, the major emphasis of this report is on staffing changes and practices in the operations department, including the operations shift crew. The findings in this report are based on interviews conducted at seven nuclear power plants and their parent utilities. A discussion of the key findings is followed by a summary of the implications of staffing issues for plant safety.

  3. Geological Disposal Options for the Radioactive Wastes from a Recycling Process of Spent Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Choi, H. J.; Lee, M. S.; Jeong, J. T.; Choi, J. W.; Kim, S. K.; Cho, D. K.; Kuk, D. H.; Cha, J. H

    2008-10-15

    The electricity from the nuclear power plants is around 40 % of total required electricity in Korea and according to the energy development plan, the proportion will be raised about 60 % in near future. To implement this plan, the most important factor is the back-end fuel cycle, namely the safe management of the spent fuel or high level radioactive wastes from the nuclear power plants. Various researches are being carried out to manage the spent fuel effectively in the world. In our country, as one of the management alternatives which is more effective and non-proliferation, pyro-processing method is being developed actively to retrieve reusable uranium and TRU, and to reduce the volume of high level waste from a Nuclear power plant. This is a new dry recycling process. In this report, the amount of various wastes and their characteristics are estimated in a Pyro-process. Based on these information, the geological disposal alternatives are developed. According to the amount and the characteristics of each waste, the concepts of waste packages and the disposal container are developed. And also from the characteristics of the radioactivity and the heat generation, multi-layer of the depth is considered to dispose these wastes. The proposed various alternatives in this report can be used as input data for design of the deep geological disposal system. And they will be improved through the application of the real site data and safety assessment in the future. After then, the final disposal concept will be selected with various assessment and the optimization will be carried out.

  4. Startup of Pumping Units in Process Water Supplies with Cooling Towers at Thermal and Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Berlin, V. V., E-mail: vberlin@rinet.ru; Murav’ev, O. A., E-mail: muraviov1954@mail.ru; Golubev, A. V., E-mail: electronik@inbox.ru [National Research University “Moscow State University of Civil Engineering,” (Russian Federation)

    2017-03-15

    Aspects of the startup of pumping units in the cooling and process water supply systems for thermal and nuclear power plants with cooling towers, the startup stages, and the limits imposed on the extreme parameters during transients are discussed.

  5. Prioritizing the countries for BOT nuclear power project using Analytic Hierarchy Process

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Woo; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    This paper proposes factors influencing the success of BOT nuclear power projects and their weighting method using Analytic Hierarchy Process (AHP) to find the optimal country which developer intends to develop. To summarize, this analytic method enable the developer to select and focus on the country which has preferable circumstance so that it enhances the efficiency of the project promotion by minimizing the opportunity cost. Also, it enables the developer to quantify the qualitative factors so that it diversifies the project success strategy and policy for the targeted country. Although the performance of this study is insufficient due to the limitation of time, small sampling and security of materials, it still has the possibility to improve the analytic model more systematically through further study with more data. Developing Build-Own(or Operate)-Transfer (BOT) nuclear power project carrying large capital in the long term requires initially well-made multi-decision which it prevents sorts of risks from unexpected situation of targeted countries. Moreover, the nuclear power project in most case is practically implemented by Government to Government cooperation, so the key concern for such nuclear power project would be naturally focused on the country situation rather than project viability at planning stage. In this regard, it requires the evaluation of targeted countries before involving the project, comprehensive and proper decision making for complex judgment factors, and efficient integration of expert's opinions, etc. Therefore, prioritizing and evaluating the feasibility of country for identification of optimal project region is very meaningful study.

  6. Limitation of the EIA Process for the assessment of nuclear fuel waste disposal in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, B.L.; Kuhn, R.G. [Guelph Univ., ON (Canada). Dept. of Geography

    1999-12-01

    The Canadian environmental impact assessment process for the Nuclear Fuel Waste Management and Disposal Concept was completed in 1994. Almost four years later, in February 1998, the Review Panel released its report. The viewpoints of those who participated in the assessment process is archived in the thousands of pages of hearing testimony, meeting transcripts and written briefs. One of the most contentious issues raised, and one that continues to plague management in Canada, is the debate surrounding how the problem of NFW waste management should be defined. The purpose of this paper is to critically assess the problem frame of the Canadian NFW management disposal concept EIS. This will be accomplished through an analysis of stakeholder participation and views, and through an evaluation of the range and nature of the information considered legitimate or constrained in the Canadian process.

  7. A Spectroscopic Study of Nuclear Processing and Production of Anomalously Strong Lines in the Crab Nebula

    CERN Document Server

    MacAlpine, G M; Lester, W R; Vanderveer, S J; Strolger, L G

    2006-01-01

    We present and discuss correlations for optical and near-infrared (5500-10030 angstroms) line intensity measurements at many positions in the Crab Nebula. These correlations suggest the existence of gas produced by a range of nuclear processing, from material in which synthesis ended with the CNO-cycle, to some helium-burning and nitrogen depletion, to regions containing enriched products of oxygen-burning. The latter exhibit a gradual, linear rise of [Ni II] emission with increasing argon enrichment, whereas gas with less nuclear processing shows markedly different [Ni II] emission characteristics, including the highest derived abundances. This suggests two origins for stable, neutron-rich nickel in the nebula: a type of "alpha-rich freezeout" in the more highly processed material, and possibly removal of ions from the neutron star in other regions. In addition, the data indicate that anomalously strong observed [C I] emission comes from broad, low-ionization zones. Although the strongest He I emission could...

  8. Model description of non-Maxwellian nuclear processes in the solar interior

    CERN Document Server

    Voronchev, Victor T; Watanabe, Yukinobu

    2016-01-01

    A consistent model for the description of non-Maxwellian nuclear processes in the solar core triggered by fast reaction-produced particles is formulated. It essentially extends an approach to study suprathermal solar reactions discussed previously [Phys. Rev. C 91, 028801 (2015)] and refines its predictions. The model is applied to examine in detail the slowing-down of 8.7-MeV alpha particles produced in the 7Li(p,alpha)alpha reaction of the pp chain, and to study suprathermal processes in the solar CNO cycle induced by them. The influence of electron degeneracy and electron screening on suprathermal reactions through in-flight reaction probability and fast particle emission rate is clarified. In particular, these effects account for a 20% increase of the 14N(alpha,p)17O reaction rate at R 18F of nuclear flow transforms to abnormal sequential flow 14N --> 17O --> 18F, and the 14N(alpha,p)17O reaction rate exceeds the rate of 17O burn up through conventional 17O(p,alpha)14N and 17O(p,gamma)18F processes. It i...

  9. Lab-scale demonstration of the UREX+1a process using spent nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Vandegrift, G. F.; Regalbuto, M. C.; Bakel, A.; Bowers, D.; Gelia, A. V.; Hebden, A. S.; Maggos, L. E.; Stepinski, D.; Tsai, Y.; Laidler, J. J.; Chemical Engineering

    2007-01-01

    The Global Nuclear Energy Partnership (GNEP) is developing technologies to greatly expand repository capacity, improve proliferation resistance, and recover valuable energy that would otherwise be discarded; thus assuring a stable energy supply for the future. An important element of this initiative is the separation of key radionuclides followed by either superior waste-disposal forms and/or transmutation of long-lived isotopes. To that end, the GNEP is developing advanced fuel reprocessing systems that separate key radionuclides from spent fuel. One of these systems is the UREX+1a process. The UREX+1a process is a series of four solvent-extraction flowsheets that perform the following operations: (1) recovery of U and Tc (UREX), (2) recovery of Cs and Sr (CCD-PEG), and (3) recovery of TRU and rare earth elements (TRUEX), and (4) separation of TRU elements from the rare earths (TALSPEAK). This paper discusses the results of the demonstration of the UREX, TRUEX, and TALSPEAK processes using spent nuclear fuel, as well as future development needs and plans.

  10. Recent Advances in Computational Methods for Nuclear Magnetic Resonance Data Processing

    KAUST Repository

    Gao, Xin

    2013-01-11

    Although three-dimensional protein structure determination using nuclear magnetic resonance (NMR) spectroscopy is a computationally costly and tedious process that would benefit from advanced computational techniques, it has not garnered much research attention from specialists in bioinformatics and computational biology. In this paper, we review recent advances in computational methods for NMR protein structure determination. We summarize the advantages of and bottlenecks in the existing methods and outline some open problems in the field. We also discuss current trends in NMR technology development and suggest directions for research on future computational methods for NMR.

  11. Nuclear Effects in Polarized Proton-Deuteron Drell-Yan Processes

    Institute of Scientific and Technical Information of China (English)

    DUAN Chun-Gui; SHI Li-Jie; SHEN Peng-Nian; LI Guang-Lie

    2004-01-01

    @@ The longitudinally polarized Drell- Yan process is one of the most powerful tools to probe the structure of hadrons.By means of the recent formalism of the polarized proton-deuteron (pd) Drell-Yan, we calculate the ratio of the proton-deuteron Drell-Yan cross section to the proton-proton (pp) one △σpd/2△σpp in the polarized case. The theoretical results can be compared with future experimental data to confirm the nuclear effect due to the 6-quark cluster in deuteron.

  12. Off-gas catalyst. Abgaskatalysator

    Energy Technology Data Exchange (ETDEWEB)

    Saris, L.; Kloeck, H.

    1987-02-19

    The invention deals with a waste gas catalyst with a thermo-resistant SiO{sub 2} and Al{sub 2}O{sub 3} containing carrier of snarled ceramic fibres which form between themselves the flow paths for the waste gas to be purified and which are coated with platinum, palladium and/or rhodium. The ceramic fibres forming the carrier consist of SiO{sub 2} and Al{sub 2}O{sub 3} and have a diameter of 1 to 10 {mu}m. (orig./RB).

  13. Evaluating Safeguards Benefits of Process Monitoring as compared with Nuclear Material Accountancy

    Energy Technology Data Exchange (ETDEWEB)

    Humberto Garcia; Wen-Chiao Lin; Reed Carlson

    2014-07-01

    This paper illustrates potential safeguards benefits that process monitoring (PM) may have as a diversion deterrent and as a complementary safeguards measure to nuclear material accountancy (NMA). This benefit is illustrated by quantifying the standard deviation associated with detecting a considered material diversion scenario using either an NMA-based method or a PM-based approach. To illustrate the benefits of PM for effective safeguards, we consider a reprocessing facility. We assume that the diversion of interest for detection manifests itself as a loss of Pu caused by abnormally operating a dissolver for an extended period to accomplish protracted diversion (or misdirection) of Pu to a retained (unconditioned) waste stream. For detecting the occurrence of this diversion (which involves anomalous operation of the dissolver), we consider two different data evaluation and integration (DEI) approaches, one based on NMA and the other based on PM. The approach based on PM does not directly do mass balance calculations, but rather monitors for the possible occurrence of anomaly patterns related to potential loss of nuclear material. It is thus assumed that the loss of a given mass amount of nuclear material can be directly associated with the execution of proliferation-driven activities that trigger the occurrence of an anomaly pattern consisting of series of events or signatures occurring at different unit operations and time instances. By effectively assessing these events over time and space, the PM-based DEI approach tries to infer whether this specific pattern of events has occurred and how many times within a given time period. To evaluate the goodness of PM, the 3 Sigma of the estimated mass loss is computed under both DEI approaches as function of the number of input batches processed. Simulation results are discussed.

  14. Simulation of coupled THM process in surrounding rock mass of nuclear waste repository in argillaceous formation

    Institute of Scientific and Technical Information of China (English)

    蒋中明; 陈永贵

    2015-01-01

    To investigate and analyze the thermo-hydro-mechanical (THM) coupling phenomena of a surrounding rock mass in an argillaceous formation, a nuclear waste disposal concept in drifts was represented physically in an in-situ test way. A transversely isotropic model was employed to reproduce the whole test process numerically. Parameters of the rock mass were determined by laboratory and in-situ experiments. Based on the numerical simulation results and in-situ test data, the variation processes of pore water pressure, temperature and deformation of surrounding rock were analyzed. Both the measured data and numerical results reveal that the thermal perturbation is the principal driving force which leads to the variation of pore water pressure and deformations in the surrounding rock. The temperature, pore pressure and deformation of rock mass change rapidly at each initial heating stage with a constant heating power. The temperature field near the heater borehole is relatively steady in the subsequent stages of the heating phase. However, the pore pressure and deformation fields decrease gradually with temperature remaining unchanged condition. It also shows that a transversely isotropic model can reproduce the THM coupling effects generating in the near-field of a nuclear waste repository in an argillaceous formation.

  15. The Optimization of Radioactive Waste Management in the Nuclear Installation Decommissioning Process

    Energy Technology Data Exchange (ETDEWEB)

    Zachar, Matej; Necas, Vladimir [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Department of Nuclear Physics and Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia)

    2008-07-01

    The paper presents a basic characterization of nuclear installation decommissioning process especially in the term of radioactive materials management. A large amount of solid materials and secondary waste created after implementation of decommissioning activities have to be managed considering their physical, chemical, toxic and radiological characteristics. Radioactive materials should be, after fulfilling all the conditions defined by the authorities, released to the environment for the further use. Non-releasable materials are considered to be a radioactive waste. Their management includes various procedures starting with pre-treatment activities, continuing with storage, treatment and conditioning procedures. Finally, they are disposed in the near surface or deep geological repositories. Considering the advantages and disadvantages of all possible ways of releasing the material from nuclear installation area, optimization of the material management process should be done. Emphasis is placed on the radiological parameters of materials, availability of waste management technologies, waste repositories and on the radiological limits and conditions for materials release or waste disposal. Appropriate optimization of material flow should lead to the significant savings of money, disposal capacities or raw material resources. Using a suitable calculation code e.g. OMEGA, the evaluation of the various material management scenarios and selection of the best one, based on the multi-criterion analysis, should be done. (authors)

  16. Transient boundary conditions in the frame of THM-processes at nuclear waste repositories

    Directory of Open Access Journals (Sweden)

    Schanz Tom

    2016-01-01

    Full Text Available In nuclear waste repositories, initially unsaturated buffer is subjected to constant heat emitted by waste canister in conjunction with peripheral hydration through water from host rock. The transient hydration process can be potraied as transformation of initial heterogeneity towards homogeneity as final stage. In this context, this paper addresses the key issue of hydro mechanical behaviour of compacted buffer in context of clay microstructure and its evolution under repository relevant loading paths and material heterogeneity. This paper also introduces a unique column experiment facility available at Ruhr Universität Bochum, Germany. The facility has been designed as a forerunner of field scale testing program to simulate the transient hydration process of compacted buffer as per German reference disposal concept. The device is unique in terms of having proficiency to capture the transient material response under various possible repository relevant loading paths with higher precision level by monitor the key parameters like temperature, total suction, water content and axial & radial swelling pressure at three different sections along the length of compacted soil sample. In general, a larger spectrum of loading paths/scenarios, which may arise in the nuclear repository, can be covered precisely with existing device.

  17. Experiment on the improvement of OREOX process for fabrication of dry recycling nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I. [and others

    2004-01-01

    The OREOX(Oxidation and REduction of OXide fuel) process has been performed to fabricate dry recycling(DUPIC ; Direct Use of spent PWR fuel In CANDU reactor) nuclear fuel pellets by using spent PWR fuel. Generally, sinterable DUPIC powder has been manufactured from spent PWR fuel pellets by the 3 cycles of oxidation and reduction treatment. The OREOX process is one of the most important processes for DUPIC pellet fabrication. A lot of time more than 37 hours as well as a lot of reaction gas is required to perform 3 cycles of OREOX treatments. In this experiment, 1 cycle OREOX process was adopted to improve the powdering process of DUPIC pellet manufacturing processes. As a result of experiment, the densities of pellets sintered at 1800 .deg. C for 10 hours ranged from 10.15 to 10.22 g/cm{sup 3}(93.8{approx}94.5 % of T.D.). The pellets were sintered again to increase the sintered density. The sintered densities of pellets re-sintered at 1850 .deg. C for 7 hours ranged from 10.27 to 10.33 g/cm{sup 3}(94.9{approx} 95.5 % of T.D)

  18. Duval Corporation application study: nuclear process energy from PE-CNSG. [Sulfur mining

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    The technical and economic studies were performed to examine the possible installation of a small, integral pressurized water reactor as an industrial energy source in the Duval Corporation's Frasch Process sulfur mining operation located in Culberson County, Texas. Since this is the first industrial application study attempted for this type of reactor, it has been a learning process on the nuclear plant side as well as the industrial side, particularly in the area of economic analysis. The importance of considering inflationary effects, the significance of plant financing form, and the annualized, after-tax cash flow and incremental rate-of-return methods of comparison in determing energy costs have all been recognized during the course of the study.

  19. Absolute quantification for benzoic acid in processed foods using quantitative proton nuclear magnetic resonance spectroscopy.

    Science.gov (United States)

    Ohtsuki, Takashi; Sato, Kyoko; Sugimoto, Naoki; Akiyama, Hiroshi; Kawamura, Yoko

    2012-09-15

    The absolute quantification method of benzoic acid (BA) in processed foods using solvent extraction and quantitative proton nuclear magnetic resonance spectroscopy was developed and validated. BA levels were determined using proton signals (δ(H) 7.53 and 7.98) referenced to 2-dimethyl-2-silapentane-5-sulfonate-d(6) sodium salt (DSS-d(6)) after simple solvent extraction from processed foods. All recoveries from several kinds of processed foods, spiked at their specified maximum Japanese usage levels (0.6-2.5 g kg(-1)) and at 0.13 g kg(-1) and 0.063 g kg(-1), were greater than 80%. The limit of quantification was confirmed as 0.063 g kg(-1) in processed foods, which was sufficiently low for the purposes of monitoring BA. The accuracy of the proposed method is equivalent to the conventional method using steam-distillation extraction and high-performance liquid chromatography. The proposed method was both rapid and simple. Moreover, it provided International System of Units traceability without the need for authentic analyte standards. Therefore, the proposed method is a useful and practical tool for determining BA levels in processed foods.

  20. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  1. On-line calibration of process instrumentation channels in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hashemian, H.M.; Farmer, J.P. [Analysis and Measurement Services Corp., Knoxville, TN (United States)

    1995-04-01

    An on-line instrumentation monitoring system was developed and validated for use in nuclear power plants. This system continuously monitors the calibration status of instrument channels and determines whether or not they require manual calibrations. This is accomplished by comparing the output of each instrument channel to an estimate of the process it is monitoring. If the deviation of the instrument channel from the process estimate is greater than an allowable limit, then the instrument is said to be {open_quotes}out of calibration{close_quotes} and manual adjustments are made to correct the calibration. The success of the on-line monitoring system depends on the accuracy of the process estimation. The system described in this paper incorporates both simple intercomparison techniques as well as analytical approaches in the form of data-driven empirical modeling to estimate the process. On-line testing of the calibration of process instrumentation channels will reduce the number of manual calibrations currently performed, thereby reducing both costs to utilities and radiation exposure to plant personnel.

  2. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  3. Vanishing current hysteresis under competing nuclear spin pumping processes in a quadruplet spin-blockaded double quantum dot

    Energy Technology Data Exchange (ETDEWEB)

    Amaha, S., E-mail: s-amaha@riken.jp [Quantum Spin Information Project, Japan Science and Technology Agency, ICORP, 3-1, Morinosato Wakamiya, Atsugi-shi, Kanagawa 243-0198 (Japan); Quantum Functional System Research Group, RIKEN Center for Emergent Matter Science, RIKEN, 3-1 Wako-shi, Saitama 351-0198 (Japan); Hatano, T. [Quantum Spin Information Project, Japan Science and Technology Agency, ICORP, 3-1, Morinosato Wakamiya, Atsugi-shi, Kanagawa 243-0198 (Japan); Department of Physics, Tohoku University, Sendai-shi, Miyagi 980-8578 (Japan); Tarucha, S. [Quantum Spin Information Project, Japan Science and Technology Agency, ICORP, 3-1, Morinosato Wakamiya, Atsugi-shi, Kanagawa 243-0198 (Japan); Quantum Functional System Research Group, RIKEN Center for Emergent Matter Science, RIKEN, 3-1 Wako-shi, Saitama 351-0198 (Japan); Department of Applied Physics, School of Engineering, University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Gupta, J. A.; Austing, D. G. [National Research Council of Canada, M50, Montreal Road, Ottawa, Ontario K1A 0R6 (Canada)

    2015-04-27

    We investigate nuclear spin pumping with five-electron quadruplet spin states in a spin-blockaded weakly coupled vertical double quantum dot device. Two types of hysteretic steps in the leakage current are observed on sweeping the magnetic field and are associated with bidirectional polarization of nuclear spin. Properties of the steps are understood in terms of bias-voltage-dependent conditions for the mixing of quadruplet and doublet spin states by the hyperfine interaction. The hysteretic steps vanish when up- and down-nuclear spin pumping processes are in close competition.

  4. Pentagalloylglucose Blocks the Nuclear Transport and the Process of Nucleocapsid Egress to Inhibit HSV-1 Infection.

    Science.gov (United States)

    Jin, Fujun; Ma, Kaiqi; Chen, Maoyun; Zou, Muping; Wu, Yanting; Li, Feng; Wang, Yifei

    2016-01-01

    Herpes simplex virus type 1 (HSV-1), a widespread virus, causes a variety of human viral diseases worldwide. The serious threat of drug-resistance highlights the extreme urgency to develop novel antiviral drugs with different mechanisms of action. Pentagalloylglucose (PGG) is a natural polyphenolic compound with significant anti-HSV activity; however, the mechanisms underlying its antiviral activity need to be defined by further studies. In this study, we found that PGG treatment delays the nuclear transport process of HSV-1 particles by inhibiting the upregulation of dynein (a cellular major motor protein) induced by HSV-1 infection. Furthermore, PGG treatment affects the nucleocapsid egress of HSV-1 by inhibiting the expression and disrupting the cellular localization of pEGFP-UL31 and pEGFP-UL34, which are indispensable for HSV-1 nucleocapsid egress from the nucleus. However, the over-expression of pEGFP-UL31 and pEGFP-UL34 could decrease the antiviral effect of PGG. In this study, for the first time, the antiviral activity of PGG against acyclovir-resistant virus was demonstrated in vitro, and the possible mechanisms of its anti-HSV activities were identified based on the inhibition of nuclear transport and nucleocapsid egress in HSV-1. It was further confirmed that PGG could be a promising candidate for HSV therapy, especially for drug-resistant strains.

  5. Industrial Qualification Process for Optical Fibers Distributed Strain and Temperature Sensing in Nuclear Waste Repositories

    Directory of Open Access Journals (Sweden)

    S. Delepine-Lesoille

    2012-01-01

    Full Text Available Temperature and strain monitoring will be implemented in the envisioned French geological repository for high- and intermediate-level long-lived nuclear wastes. Raman and Brillouin scatterings in optical fibers are efficient industrial methods to provide distributed temperature and strain measurements. Gamma radiation and hydrogen release from nuclear wastes can however affect the measurements. An industrial qualification process is successfully proposed and implemented. Induced measurement uncertainties and their physical origins are quantified. The optical fiber composition influence is assessed. Based on radiation-hard fibers and carbon-primary coatings, we showed that the proposed system can provide accurate temperature and strain measurements up to 0.5 MGy and 100% hydrogen concentration in the atmosphere, over 200 m distance range. The selected system was successfully implemented in the Andra underground laboratory, in one-to-one scale mockup of future cells, into concrete liners. We demonstrated the efficiency of simultaneous Raman and Brillouin scattering measurements to provide both strain and temperature distributed measurements. We showed that 1.3 μm working wavelength is in favor of hazardous environment monitoring.

  6. Fission product iodine during early Hanford-Site operations: Its production and behavior during fuel processing, off-gas treatment and release to the atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Burger, L.L.

    1991-05-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project was established to estimate the radiological dose impact that Hanford Site operations may have made on the local and regional population. This impact is estimated by examining operations involving radioactive materials that were conducted at the Hanford Site from the startup of the first reactor in 1944 to the present. HEDR Project work is divided among several technical tasks. One of these tasks, Source Terms, is designed to develop quantitative estimates of all significant emissions of radionuclides by Hanford Site operations since 1944. Radiation doses can be estimated from these emissions by accounting for specific radionuclide transport conditions and population demography. This document provides technical information to assist in the evaluation of iodine releases. 115 refs., 5 figs., 3 tabs.

  7. Systematic and correlated nuclear uncertainties in the i-process at the neutron shell closure N = 82

    CERN Document Server

    Bertolli, M G; Pignatari, M; Kawano, T

    2013-01-01

    Nuclear astrophysics simulations aiming to study the origin of the elements in stars require a multitude of nuclear physics input. Both systematic model dependent and statistically correlated uncertainties need to be considered. An application where realistic uncertainty assessments are especially important is the intermediate neutron capture process or i process: a neutron capture regime with neutron densities intermediate between the slow and rapid processes. Accordingly, the main network flux proceeds on the neutron-rich unstable isotopes up to 4-5 species off the valley of stability. The i process has been clearly identified to be active in post-AGB stars during the Very Late Thermal Pulse H-ingestion event, and a recent work infers about its important role in early generations of stars. Here we demonstrate the effect of propagating systematic nuclear uncertainties from different theoretical models to final abundances for a region around the 2$^\\mathrm{nd}$ peak at $A-Z=80$ for elemental ratio predictions...

  8. Innovation in the processes of formation and training of nuclear professionals; La innovacion en los procesos de formacion y entranamiento de los profesionales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Martinez, F. J.; Lambistos Agustin, A.

    2015-07-01

    Innovation is the intoduction of new products and services, new processes, new sources of supply and changes in industrial organization, and continuous customer, consumer or user oriented (J. A. Schumpeter). According to this idea, three mental restrictions usually apply to the innovative break: not only are new products, not only are technological developments, not only are revolutionary ideas so also. From the innovative tradition of Tecnatom Formacion Nuclear materailized in examples like the SGI or Human Factors simulators, in recent years has made considerable progress in the function with innovative solutions to improve the results of nuclear power plants, made available to our customers, as significant as the Training Programs for Shift Supervisors, the OJT/TPE processes, seminars Diagnostic Techniques, EDMG Simulator or ROI and ROIF projects. (Author)

  9. Modeling Electronic-Nuclear Interactions for Excitation Energy Transfer Processes in Light-Harvesting Complexes.

    Science.gov (United States)

    Lee, Mi Kyung; Coker, David F

    2016-08-18

    An accurate approach for computing intermolecular and intrachromophore contributions to spectral densities to describe the electronic-nuclear interactions relevant for modeling excitation energy transfer processes in light harvesting systems is presented. The approach is based on molecular dynamics (MD) calculations of classical correlation functions of long-range contributions to excitation energy fluctuations and a separate harmonic analysis and single-point gradient quantum calculations for electron-intrachromophore vibrational couplings. A simple model is also presented that enables detailed analysis of the shortcomings of standard MD-based excitation energy fluctuation correlation function approaches. The method introduced here avoids these problems, and its reliability is demonstrated in accurate predictions for bacteriochlorophyll molecules in the Fenna-Matthews-Olson pigment-protein complex, where excellent agreement with experimental spectral densities is found. This efficient approach can provide instantaneous spectral densities for treating the influence of fluctuations in environmental dissipation on fast electronic relaxation.

  10. Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Theodore E. Madey

    2001-10-01

    Characterize the effects of temperature and radiation processes on the interactions of H20 with oxide surfaces. Our experiments focused on the fundamental interaction of H20 molecules with surfaces of U02. We characterized the surface chemistry of adsorbed H2O using thermal desorption methods and radiotracer methods, as well as x-ray photoelectron spectroscopy (XPS) and low energy ion scattering (LEIS). In parallel with these measurements of thermal effects, we examined the effects of secondary electrons and high-energy photons on hydrogen and oxygen generation and, and how this related to corrosion of spent nuclear fuel. These studies concentrated on neutral and ionic (cation and anion) desorption products of low-energy electron irradiation of water-covered UO2.

  11. Developments in quantum information processing by nuclear magnetic resonance: Use of quadrupolar and dipolar couplings

    Indian Academy of Sciences (India)

    Anil Kumar; K V Ramanathan; T S Mahesh; Neeraj Sinha; K V R Murali

    2002-08-01

    Use of dipolar and quadrupolar couplings for quantum information processing (QIP) by nuclear magnetic resonance (NMR) is described. In these cases, instead of the individual spins being qubits, the 2 energy levels of the spin-system can be treated as an -qubit system. It is demonstrated that QIP in such systems can be carried out using transition-selective pulses, in CH3CN, 13CH3CN, 7Li ( = 3/2) and 133Cs ( = 7/2), oriented in liquid crystals yielding 2 and 3 qubit systems. Creation of pseudopure states, implementation of logic gates and arithmetic operations (half-adder and subtractor) have been carried out in these systems using transition-selective pulses.

  12. Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Robert T [ORNL; Collins, Jack Lee [ORNL; Hunt, Rodney Dale [ORNL; Ladd-Lively, Jennifer L [ORNL; Patton, Kaara K [ORNL; Hickman, Robert [NASA Marshall Space Flight Center, Huntsville, AL

    2014-01-01

    The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

  13. Process of Judging Significant Modifications for Different Transportation Systems compared to the Approach for Nuclear Installations

    Directory of Open Access Journals (Sweden)

    Nicolas Petrek

    2015-12-01

    Full Text Available The implementation of the CSM regulation by the European Commission in 2009 which harmonizes the risk assessment process and introduces a rather new concept of judging changes within the European railway industry. This circumstance has risen the question how other technology sectors handle the aspect of modifications and alterations. The paper discusses the approaches for judging the significance of modifications within the three transport sectors of European railways, aviation and maritime transportation and the procedure which is used in the area of nuclear safety. We will outline the similarities and differences between these four methods and discuss the underlying reasons. Finally, we will take into account the role of the European legislator and the fundamental idea of a harmonization of the different approaches.

  14. Performance based seismic qualification of reinforced concrete nuclear materials processing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mertz, G.E.; Loceff, F.; Houston, T.; Rauls, G. [Westinghouse Savannah River Company, Aiken, SC (United States); Mulliken, J. [LPA Group Inc., SC (United States)

    1997-09-01

    A seismic qualification of a reinforced concrete nuclear materials processing facility using performance based acceptance criteria is presented. Performance goals are defined in terms of a minimum annual seismic failure frequency. Pushover analyses are used to determine the building`s ultimate capacity and relate the capacity to roof drift and joint rotation. Nonlinear dynamic analyses are used to quantify the building`s drift using a suite of ground motion intensities representing varying soil conditions and levels of seismic hazard. A correlation between joint rotation and building drift to damage state is developed from experimental data. The damage state and seismic hazard are convolved to determine annual seismic failure frequency. The results of this rigorous approach is compared to those using equivalent force methods and pushover techniques recommended by ATC-19 and FEMA-273.

  15. Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

  16. Absolute quantitative analysis for sorbic acid in processed foods using proton nuclear magnetic resonance spectroscopy.

    Science.gov (United States)

    Ohtsuki, Takashi; Sato, Kyoko; Sugimoto, Naoki; Akiyama, Hiroshi; Kawamura, Yoko

    2012-07-13

    An analytical method using solvent extraction and quantitative proton nuclear magnetic resonance (qHNMR) spectroscopy was applied and validated for the absolute quantification of sorbic acid (SA) in processed foods. The proposed method showed good linearity. The recoveries for samples spiked at the maximum usage level specified for food in Japan and at 0.13 g kg(-1) (beverage: 0.013 g kg(-1)) were larger than 80%, whereas those for samples spiked at 0.063 g kg(-1) (beverage: 0.0063 g kg(-1)) were between 56.9 and 83.5%. The limit of quantification was 0.063 g kg(-1) for foods (and 0.0063 g kg(-1) for beverages containing Lactobacillus species). Analysis of the SA content of commercial processed foods revealed quantities equal to or greater than those measured using conventional steam-distillation extraction and high-performance liquid chromatography quantification. The proposed method was rapid, simple, accurate, and precise, and provided International System of Units traceability without the need for authentic analyte standards. It could therefore be used as an alternative to the quantification of SA in processed foods using conventional method.

  17. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels.

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, S. F.

    1999-03-24

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns.

  18. Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

    Science.gov (United States)

    Al-Jobouri, Hussain A.; Rajab, Mustafa Y.

    2016-03-01

    CR-39 detector which covered with boric acid (H3Bo3) pellet was irradiated by thermal neutrons from (241Am - 9Be) source with activity 12Ci and neutron flux 105 n. cm-2. s-1. The irradiation times -TD for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C˚ temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLAB software were analyzed and found the following relationships: (a) The irradiation time -TD has behavior linear relationships with following nuclear track parameters: i) total track number - NT ii) maximum track number - MRD (relative to track diameter - DT) at response region range 2.5 µm to 4 µm iii) maximum track number - MD (without depending on track diameter - DT). (b) The irradiation time -TD has behavior logarithmic relationship with maximum track number - MA (without depending on track area - AT). The image processing technique principally track diameter - DT can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.

  19. Final disposal of spent nuclear fuel - regulatory system and roles of different actors during the decision process

    Energy Technology Data Exchange (ETDEWEB)

    2009-03-15

    In November 2006 Swedish Nuclear Fuels Co. applied for a license to build a plant for encapsulation of spent nuclear fuels at Oskarshamn, Sweden. The company also have plans to apply, in 2009, for a license to construct a underground repository for spent nuclear fuels. KASAM arranged a seminar in November 2006 in order to describe and discuss the licensing rules and regulations and the roles of different parties in the decision making. Another objective of the seminar was to point out possible ambiguities in this process. Another interesting question under discussion was in what ways the basic data for the decision should be produced. The seminar covered the part of the process beginning with the application for a license and ending with the government approval/rejection of the application. Most time was spent on the legal aspects of the process

  20. Absolute quantitative analysis for sorbic acid in processed foods using proton nuclear magnetic resonance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsuki, Takashi, E-mail: ohtsuki@nihs.go.jp [National Institute of Health Sciences, 1-18-1 Kamiyoga, Setagaya-ku, Tokyo 158-8501 (Japan); Sato, Kyoko; Sugimoto, Naoki; Akiyama, Hiroshi; Kawamura, Yoko [National Institute of Health Sciences, 1-18-1 Kamiyoga, Setagaya-ku, Tokyo 158-8501 (Japan)

    2012-07-13

    Highlights: Black-Right-Pointing-Pointer A method using qHNMR was applied and validated to determine SA in processed foods. Black-Right-Pointing-Pointer This method has good accuracy, precision, selectiveness, and linearity. Black-Right-Pointing-Pointer The proposed method is more rapid and simple than the conventional method. Black-Right-Pointing-Pointer We found that the proposed method is reliable for the accurate determination of SA. Black-Right-Pointing-Pointer This method can be used for the monitoring of SA in processed foods. - Abstract: An analytical method using solvent extraction and quantitative proton nuclear magnetic resonance (qHNMR) spectroscopy was applied and validated for the absolute quantification of sorbic acid (SA) in processed foods. The proposed method showed good linearity. The recoveries for samples spiked at the maximum usage level specified for food in Japan and at 0.13 g kg{sup -1} (beverage: 0.013 g kg{sup -1}) were larger than 80%, whereas those for samples spiked at 0.063 g kg{sup -1} (beverage: 0.0063 g kg{sup -1}) were between 56.9 and 83.5%. The limit of quantification was 0.063 g kg{sup -1} for foods (and 0.0063 g kg{sup -1} for beverages containing Lactobacillus species). Analysis of the SA content of commercial processed foods revealed quantities equal to or greater than those measured using conventional steam-distillation extraction and high-performance liquid chromatography quantification. The proposed method was rapid, simple, accurate, and precise, and provided International System of Units traceability without the need for authentic analyte standards. It could therefore be used as an alternative to the quantification of SA in processed foods using conventional method.

  1. Explosive Nuclear Burning in the pp-Chain Region and the Breakout Processes

    Directory of Open Access Journals (Sweden)

    Kubono S.

    2016-01-01

    Full Text Available The nuclear reactions in the pp-chain region and on the breakout process from the pp-chain region under very high temperature conditions are reviewed, and some possibilities for experimental investigation are discussed. The reactions discussed could play an important role typically for the primordial nucleosynthesis and supernova nucleosynthesis. Specifically, I discuss here the reactions starting from the two key nuclei, 7Be and 7Li. The 7Be(n,α reaction, which destroys 7Be, is considered to have a large impact to the primordial 7Li problem. Our recent estimate of the reaction rate indicates that the reaction rate can be about one order of magnitude smaller than the rate currently adopted, suggesting this channel has a minor effect for the 7Li problem. Under a proton-rich environment at high temperature like the νp-process, the 7Be(α,γ11C(α,p14N pathway is expected to play a majpr role for heavy element synthesis, comparable to the triple alpha process. These two reactions on the pathway were investigated by using low-energy, high-intensity RI beams of 7Be and 11C. The results support the theoretical prediction of heavy nucleus production at around mass 90-100 by the νp-process, where the anomalously abundant p-nuclei exist. The reactions on the breakout sequence of 7Li(n,γ8Li(α,n11B are also discussed which could paly a crucial role in nuetron-rich envirnments, like in the primirdial universe as well as the early stage of the r-process. The cross sections of the first step reaction 7Li(n,γ8Li seems well confirmed, but the second step reaction 8Li(α,n11B still is not well known yet, whose status of the study is discussed.

  2. Energy analysis of a desalination process of sea water with nuclear energy; Analisis energetico de un proceso de desalinizacion de agua de mar con energia nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Martinez L, G.; Valle H, J., E-mail: julfi_jg@yahoo.com.mx [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca No. 1009, Ex-hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2016-09-15

    In the present work, is theoretically proven that the residual heat, removed by the chillers in the stage prior to the compression of the recuperative Brayton cycle with which nuclear power plants operate with high temperature gas reactors (HTGR), can be used to produce stem and desalinate seawater. The desalination process selected for the analysis, based on its operating characteristics, is the Multi-Stage Distillation (Med). The Med process will use as energy source, for the flash evaporation process in the flash trap, the residual heat that the reactor coolant dissipates to the environment in order to increase the compression efficiency of the same; the energy dissipated depends on the operating conditions of the reactor. The Med distillation process requires saturated steam at low pressure which can be obtained by means of a heat exchanger, taking advantage of the residual heat, where the relative low temperatures with which the process operates make the nuclear plants with HTGR reactors ideal for desalination of sea water, because they do not require major modifications to their design of their operation. In this work the energy analysis of a six-stage Med module coupled to the chillers of an HTGR reactor of the Pebble Bed Modular Reactor type is presented. Mathematical modeling was obtained by differential equations of mass and energy balances in the system. The results of the analysis are presented in a table for each distillation stage, estimating the pure water obtained as a function of the heat supplied. (Author)

  3. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  4. Hydrogen production coupled to nuclear waste treatment: the safe treatment of alkali metals through a well-demonstrated process

    Energy Technology Data Exchange (ETDEWEB)

    Rahier, A. [Studie Centrum Voor Kernenergie (SCK-CEN), Boeretang, Mol (Belgium)]. E-mail: arahier@sckcen.be; Mesrobian, G. [Euro Chlor, Brussels (Belgium)]. E-mail: guy.mesrobian@wanadoo.fr

    2006-07-01

    In 1992, the United Nations emphasised the urgent need to act against the perpetuation of disparities between and within nations, the worsening of poverty, hunger, ill health and illiteracy and the continuing deterioration of ecosystems on which we depend for our well-being. In this framework, taking into account the preservation of both worldwide energy resources and ecosystems, the use of nuclear energy to produce clean energy carriers, such as hydrogen, is undoubtedly advisable. However, coping fully with the Agenda 21 statements requires defining adequate treatment processes for nuclear wastes. This paper discusses the possible use of a well-demonstrated process to convert radioactively contaminated alkali metals into sodium hydroxide while producing hydrogen. We conclude that a synergy between Chlor-Alkali specialists and nuclear specialists may help find an acceptable solution for radioactively contaminated sodium waste. (author)

  5. CORROSION OF LEAD SHIELDING IN NUCLEAR MATERIALS PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    Subramanian, K; Kerry Dunn, K

    2007-11-16

    Inspection of United States-Department of Energy (US-DOE) model 9975 nuclear materials shipping package revealed corrosion of the lead shielding induced by off-gas constituents from organic components in the package. Experiments were performed to determine the corrosion rate of lead when exposed to off-gas or degradation products of these organic materials. The results showed that the room temperature vulcanizing (RTV) sealant was the most corrosive organic species followed by the polyvinyl acetate (PVAc) glue. The fiberboard material induced corrosion to a much lesser extent than the PVAc glue and RTV, and only in the presence of condensed water. The results indicated faster corrosion at temperatures higher than ambient and with condensed water as expected. A corrosion rate of 0.05 mm/year measured for coupons exposed to the most aggressive conditions was recommended as a conservative estimate for use in package performance calculations.

  6. CORROSION OF LEAD SHIELDING IN NUCLEAR MATERIALS PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    Subramanian, K; Kerry Dunn, K; Joseph Murphy, J

    2008-07-18

    Inspection of United States-Department of Energy (US-DOE) model 9975 nuclear materials shipping package revealed corrosion of the lead shielding that was induced by off-gas constituents from organic components in the package. Experiments were performed to determine the corrosion rate of lead when exposed to off-gas or degradation products of these organic materials. The results showed that the room temperature vulcanizing (RTV) sealant was the most corrosive organic species used in the construction of the packaging, followed by polyvinyl acetate (PVAc) glue. Fiberboard material, also used in the construction of the packaging induced corrosion to a much lesser extent than the PVAc glue and RTV sealant, and only in the presence of condensed water. The results indicated faster corrosion at temperatures higher than ambient and with condensed water. In light of these corrosion mechanisms, the lead shielding was sheathed in a stainless steel liner to mitigate against corrosion.

  7. Processing of nuclear viroids in vivo: an interplay between RNA conformations.

    Directory of Open Access Journals (Sweden)

    María-Eugenia Gas

    2007-11-01

    Full Text Available Replication of viroids, small non-protein-coding plant pathogenic RNAs, entails reiterative transcription of their incoming single-stranded circular genomes, to which the (+ polarity is arbitrarily assigned, cleavage of the oligomeric strands of one or both polarities to unit-length, and ligation to circular RNAs. While cleavage in chloroplastic viroids (family Avsunviroidae is mediated by hammerhead ribozymes, where and how cleavage of oligomeric (+ RNAs of nuclear viroids (family Pospiviroidae occurs in vivo remains controversial. Previous in vitro data indicated that a hairpin capped by a GAAA tetraloop is the RNA motif directing cleavage and a loop E motif ligation. Here we have re-examined this question in vivo, taking advantage of earlier findings showing that dimeric viroid (+ RNAs of the family Pospiviroidae transgenically expressed in Arabidopsis thaliana are processed correctly. Using this methodology, we have mapped the processing site of three members of this family at equivalent positions of the hairpin I/double-stranded structure that the upper strand and flanking nucleotides of the central conserved region (CCR can form. More specifically, from the effects of 16 mutations on Citrus exocortis viroid expressed transgenically in A. thaliana, we conclude that the substrate for in vivo cleavage is the conserved double-stranded structure, with hairpin I potentially facilitating the adoption of this structure, whereas ligation is determined by loop E and flanking nucleotides of the two CCR strands. These results have deep implications on the underlying mechanism of both processing reactions, which are most likely catalyzed by enzymes different from those generally assumed: cleavage by a member of the RNase III family, and ligation by an RNA ligase distinct from the only one characterized so far in plants, thus predicting the existence of at least a second plant RNA ligase.

  8. Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

    Energy Technology Data Exchange (ETDEWEB)

    Al-Jobouri, Hussain A., E-mail: hahmed54@gmail.com; Rajab, Mustafa Y., E-mail: mostafaheete@gmail.com [Department of Physics, College of Science, AL-Nahrain University, Baghdad (Iraq)

    2016-03-25

    CR-39 detector which covered with boric acid (H{sub 3}Bo{sub 3}) pellet was irradiated by thermal neutrons from ({sup 241}Am - {sup 9}Be) source with activity 12Ci and neutron flux 10{sup 5} n. cm{sup −2}. s{sup −1}. The irradiation times -T{sub D} for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C° temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLAB software were analyzed and found the following relationships: (a) The irradiation time -T{sub D} has behavior linear relationships with following nuclear track parameters: i) total track number - N{sub T} ii) maximum track number - MRD (relative to track diameter - D{sub T}) at response region range 2.5 µm to 4 µm iii) maximum track number - M{sub D} (without depending on track diameter - D{sub T}). (b) The irradiation time -T{sub D} has behavior logarithmic relationship with maximum track number - M{sub A} (without depending on track area - A{sub T}). The image processing technique principally track diameter - D{sub T} can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.

  9. On the Future of Global Nuclear Arms Control and Disarmament Process

    Institute of Scientific and Technical Information of China (English)

    Shi Jianbin; Zhu Jianyu

    2016-01-01

    Because of the amazing destructions and great damages,nuclear weapons have been the objects the international community devotes itself to restrict and eliminate since their emergence.Compared to the height of the Cold War,although currently the size of the U.S.and Russian nuclear arsenals have been greatly

  10. Economic impact associated with the decommissioning process of Vandellos I Nuclear Power Plant; Informe final. Impacto economico del desmantelamiento de la central nuclear Vandellos I

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Silva, M.

    2005-07-01

    This economic study examines the economic impact associated with the decommissioning process of the Vandellos I Nuclear Power Plant, measured in terms of the global income that generated the ending of the Nuclear Power Plant activity, on the territory. To this end, we will take into account the total investment that has been necessary to complete the process of decommissioning. The economic impact is calculated using the Input- Output methodology. Briefly, the Input-Output model defines a group of accounting relationships that reflect the links taking place within the production system. The Input-Output model is based on the assumption that given an increase (decrease) in the final demand of one sector, this sector should produce more (less) to satisfy this new demand. At the same time, this will lead to demand more (less) intermediate consumption goods from the remainder sectors of the economy. Then, these sectors should produce more (less) and use more (less) intermediate inputs, and so on. Therefore, an increase (decrease) in the final demand of one sector multiplies the effect throughout the economy, following the interdependency relationships that exist among the productive activities. We will start by collecting an exhaustive economic information. This information covers the whole decommissioning process and the whole economic and productive activity of the province of Tarragona. Next, this information is used with the objective of building an Input-Output table of the province that will serve as a base to establish the global economic impact of Vandellos I. The incomes and employment generation has been evaluated in the province of Tarragona that, following the main assumptions, correspond to the global effects of the decommissioning. In addition, we have evaluated the income and employment generation within the region where the nuclear power plant is located. The total income impacts show a high multiplier effect due to the investment carried out during the

  11. Overview of U.S. nuclear launch safety approval process, supporting launch vehicle databook and probabilistic risk assessment methods

    Science.gov (United States)

    Reinhart, L. E.

    2001-01-01

    This paper provides an overview of the U.S. space nuclear power system launch approval process as defined by the two separate requirements of the National Environmental Policy Act (NEPA) and Presidential Directive/National Security Council Memorandum No. 25 (PD/NSC-25).

  12. Overview of U.S. nuclear launch safety approval process, supporting launch vehicle databook and probabilistic risk assessment methods

    Science.gov (United States)

    Reinhart, L. E.

    2001-01-01

    This paper provides an overview of the U.S. space nuclear power system launch approval process as defined by the two separate requirements of the National Environmental Policy Act (NEPA) and Presidential Directive/National Security Council Memorandum No. 25 (PD/NSC-25).

  13. Semiclassical analysis of the electron-nuclear coupling in electronic non-adiabatic processes

    CERN Document Server

    Agostini, Federica; Gross, E K U

    2015-01-01

    In the context of the exact factorization of the electron-nuclear wave function, the coupling between electrons and nuclei beyond the adiabatic regime is encoded (i) in the time-dependent vector and scalar potentials and (ii) in the electron-nuclear coupling operator. The former appear in the Schroedinger-like equation that drives the evolution of the nuclear degrees of freedom, whereas the latter is responsible for inducing non-adiabatic effects in the electronic evolution equation. As we have devoted previous studies to the analysis of the vector and scalar potentials, in this paper we focus on the properties of the electron-nuclear coupling operator, with the aim of describing a numerical procedure to approximate it within a semiclassical treatment of the nuclear dynamics.

  14. A review for identification of initiating events in event tree development process on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Riyadi, Eko H., E-mail: e.riyadi@bapeten.go.id [Center for Regulatory Assessment of Nuclear Installation and Materials, Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada 8 Jakarta 10120 (Indonesia)

    2014-09-30

    Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.

  15. Symmetry based frequency domain processing to remove harmonic noise from surface nuclear magnetic resonance measurements

    Science.gov (United States)

    Hein, Annette; Larsen, Jakob Juul; Parsekian, Andrew D.

    2017-02-01

    Surface nuclear magnetic resonance (NMR) is a unique geophysical method due to its direct sensitivity to water. A key limitation to overcome is the difficulty of making surface NMR measurements in environments with anthropogenic electromagnetic noise, particularly constant frequency sources such as powerlines. Here we present a method of removing harmonic noise by utilizing frequency domain symmetry of surface NMR signals to reconstruct portions of the spectrum corrupted by frequency-domain noise peaks. This method supplements the existing NMR processing workflow and is applicable after despiking, coherent noise cancellation, and stacking. The symmetry based correction is simple, grounded in mathematical theory describing NMR signals, does not introduce errors into the data set, and requires no prior knowledge about the harmonics. Modelling and field examples show that symmetry based noise removal reduces the effects of harmonics. In one modelling example, symmetry based noise removal improved signal-to-noise ratio in the data by 10 per cent. This improvement had noticeable effects on inversion parameters including water content and the decay constant T2*. Within water content profiles, aquifer boundaries and water content are more accurate after harmonics are removed. Fewer spurious water content spikes appear within aquifers, which is especially useful for resolving multilayered structures. Within T2* profiles, estimates are more accurate after harmonics are removed, especially in the lower half of profiles.

  16. D-D nuclear fusion processes induced in polyethylene foams by TW Laser-generated plasma

    Science.gov (United States)

    Torrisi, L.; Cutroneo, M.; Cavallaro, S.; Ullschmied, J.

    2015-06-01

    Deuterium-Deuterium fusion processes were generated by focusing the 3 TW PALS Laser on solid deuterated polyethylene targets placed in vacuum. Deuterium ion acceleration of the order of 4 MeV was obtained using laser irradiance Iλ2 ˜ 5 × 1016 W μm2/cm2 on the target. Thin and thick targets, at low and high density, were irradiated and plasma properties were monitored "on line" and "off line". The ion emission from plasma was monitored with Thomson Parabola Spectrometer, track detectors and ion collectors. Fast semiconductor detectors based on SiC and fast plastic scintillators, both employed in time-of-flight configuration, have permitted to detect the characteristic 3.0 MeV protons and 2.45 MeV neutrons emission from the nuclear fusion reactions. From massive absorbent targets we have evaluated the neutron flux by varying from negligible values up to about 5 × 107 neutrons per laser shot in the case of foams targets, indicating a reaction rate of the order of 108 fusion events per laser shot using "advanced targets".

  17. D-D nuclear fusion processes induced in polyethylene foams by TW Laser-generated plasma

    Directory of Open Access Journals (Sweden)

    Torrisi L.

    2015-01-01

    Full Text Available Deuterium-Deuterium fusion processes were generated by focusing the 3 TW PALS Laser on solid deuterated polyethylene targets placed in vacuum. Deuterium ion acceleration of the order of 4 MeV was obtained using laser irradiance Iλ2 ∼ 5 × 1016 W μm2/cm2 on the target. Thin and thick targets, at low and high density, were irradiated and plasma properties were monitored “on line” and “off line”. The ion emission from plasma was monitored with Thomson Parabola Spectrometer, track detectors and ion collectors. Fast semiconductor detectors based on SiC and fast plastic scintillators, both employed in time-of-flight configuration, have permitted to detect the characteristic 3.0 MeV protons and 2.45 MeV neutrons emission from the nuclear fusion reactions. From massive absorbent targets we have evaluated the neutron flux by varying from negligible values up to about 5 × 107 neutrons per laser shot in the case of foams targets, indicating a reaction rate of the order of 108 fusion events per laser shot using “advanced targets”.

  18. The role of non-elastic nuclear processes for intermediate-energy protons in silicon targets

    Energy Technology Data Exchange (ETDEWEB)

    Hormaza, Joel Mesa, E-mail: jmesa@ibb.unesp.br [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil); Garcia, Cesar E., E-mail: cgarcia@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), Havana (Cuba); Arruda Neto, Joao D.T.; Rodrigues, Tulio E., E-mail: arruda@if.usp.br, E-mail: tulio@if.usp.br [Universidade de Sao Paulo (USP), Sao Paulo, SP (Brazil). Instituto de Fisica; Schelin, Hugo R.; Denyak, Valery, E-mail: schelin@utfpr.edu.br, E-mail: denyak@gmail.com [Instituto de Pesquisa Pele Pequeno Principe, Curitiba, PR (Brazil); Paschuck, Sergei A.; Evseev, Ivan, E-mail: sergei@utfpr.edu.br, E-mail: evseev@utfpr.edu.br [Universidade Tecnologica Federal do Parana (UTFPR), Curitiba, PR (Brazil)

    2013-07-01

    The transportation of energetic ions in bulk matter is of direct interest in several areas including shielding against ions originating from either space radiations or terrestrial accelerators, cosmic ray propagation studies in galactic medium, or radiobiological effects resulting from the work place or clinical exposures. For carcinogenesis, terrestrial radiation therapy, and radiobiological research, knowledge of beam composition and interactions is necessary to properly evaluate the effects on human and animal tissues. For the proper assessment of radiation exposures both reliable transport codes and accurate input parameters are needed. In the last years efforts have been increasing in order to develop more effective models to describe and predict the damages induced by radiation in electronic devices. In this sense, the interaction of protons with those devices, particularly which operate in space, is a topic of paramount importance, mainly because although the majority of them are made with silicon, experimental data on p+Si nuclear processes is very sparse. In this work we have used a new quite sophisticated Monte Carlo multicollisional intranuclear cascade (MCMC) code for pre-equilibrium emission, plus de-excitation of residual nucleus by two ways: evaporation of particles (mainly nucleons, but also composites) and possibly fragmentation/fission in the case of heavy residues, in order to study some observable of nuclear interaction of protons between 100-200 MeV in a {sup 28}Si target. The code has been developed with very recent improvements that take into account Pauli blocking effects in a novel and more precise way, as well as a more rigorous energy balance, an energy stopping time criterion for pre-equilibrium emission and the inclusion of deuteron, triton and 3He emissions in the evaporation step, which eventually concurs with fragmentation/break-up stage. The fragment mass distributions, as well as the multiplicities and the spectra of secondary

  19. The multi-isotope process monitor: Non-destructive, near-real-time nuclear safeguards monitoring at a reprocessing facility

    Science.gov (United States)

    Orton, Christopher Robert

    The IAEA will require advanced technologies to effectively safeguard nuclear material at envisioned large scale nuclear reprocessing plants. This dissertation describes results from simulations and experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel safeguards approach for process monitoring in reprocessing plants. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams, nondestructively and in near-real time (NRT). Three different models were used to predict spent nuclear fuel composition, estimate chemical distribution during separation, and simulate spectra from a variety of gamma detectors in product and raffinate streams for processed fuel. This was done for fuel with various irradiation histories and under a variety of plant operating conditions. Experiments were performed to validate the results from the model. Three segments of commercial spent nuclear fuel with variations in burnup and cooling time were dissolved and subjected to a batch PUREX method to separate the uranium and plutonium from fission and activation products. Gamma spectra were recorded by high purity germanium (HPGe) and cadmium zinc telluride (CZT) detectors. Hierarchal Cluster Analysis (HCA) and Principal Component Analysis (PCA) were applied to spectra from both model and experiment to investigate spectral variations as a function of acid concentration, burnup level and cooling time. Partial Least Squares was utilized to extract quantitative information about process variables, such as acid concentration or burnup. The MIP Monitor was found to be sensitive to the induced variations of the process and was capable of extracting quantitative process information from the analyzed spectra.

  20. Microemulsions and Aggregation Formation in Extraction Processes for Used Nuclear Fuel: Thermodynamic and Structural Studies

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Mikael [Univ. of California, Irvine, CA (United States)

    2016-05-04

    Advanced nuclear fuel cycles rely on successful chemical separation of various elements in the used fuel. Numerous solvent extraction (SX) processes have been developed for the recovery and purification of metal ions from this used material. However, the predictability of process operations has been challenged by the lack of a fundamental understanding of the chemical interactions in several of these separation systems. For example, gaps in the thermodynamic description of the mechanism and the complexes formed will make predictions very challenging. Recent studies of certain extraction systems under development and a number of more established SX processes have suggested that aggregate formation in the organic phase results in a transformation of its selectivity and efficiency. Aggregation phenomena have consistently been interfering in SX process development, and have, over the years, become synonymous with an undesirable effect that must be prevented. This multiyear, multicollaborative research effort was carried out to study solvation and self-organization in non-aqueous solutions at conditions promoting aggregation phenomena. Our approach to this challenging topic was to investigate extraction systems comprising more than one extraction reagent where synergy of the metal ion could be observed. These systems were probed for the existence of stable microemulsions in the organic phase, and a number of high-end characterization tools were employed to elucidate the role of the aggregates in metal ion extraction. The ultimate goal was to find connections between synergy of metal ion extraction and reverse micellar formation. Our main accomplishment for this project was the expansion of the understanding of metal ion complexation in the extraction system combining tributyl phosphate (TBP) and dibutyl phosphoric acid (HDBP). We have found that for this system no direct correlation exists for the metal ion extraction and the formation of aggregates, meaning that the

  1. Systems and methods for processing irradiation targets through a nuclear reactor

    Science.gov (United States)

    Dayal, Yogeshwar; Saito, Earl F.; Berger, John F.; Brittingham, Martin W.; Morales, Stephen K.; Hare, Jeffrey M.

    2016-05-03

    Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

  2. Processing of three-dimensional structures of Nuclear Medicine in PET modality; Processamento de estruturas tridimensionais de medicina nuclear na modalidade PET

    Energy Technology Data Exchange (ETDEWEB)

    Pacheco, Edward Florez; Furuie, Sergio Shiguemi, E-mail: edward.florez@usp.br [Universidade de Sao Paulo (EP/USP), SP (Brazil). Dept. de Engenharia de Telecomunicacoes e Controle. Lab. de Engenharia Biomedica

    2013-03-15

    The nuclear medicine, as a specialty to obtain medical images is very important, and it has became one of the main procedures utilized in health care centers to analyze the metabolic behavior of the patient. This project was based on medical images obtained by the PET modality (Positron Emission Tomography). Thus, we developed a framework for processing Nuclear Medicine three-dimensional images of the PET modality, which is composed of consecutive steps that start with the generation of standard images (gold standard) by using simulated images of the Left Ventricular Heart, such as phantoms obtained from the NCAT-4D software. Then, Poisson quantum noise was introduced into the whole volume to simulate the characteristic noises in PET images. Subsequently, the pre-processing step was executed by using specific 3D filters, such as the median filter, the weighted Gaussian filter, and the Anscombe/Wiener filter. Then the segmentation process, which is based on the fuzzy connectedness theory, was implemented. For that purpose four different 3D approaches were implemented: Generic, LIFO, kTetaFOEMS, and dynamic weight algorithm. Finally, an assessment procedure was used as a measurement tool to quantify three parameters (true positive, false positive and maximum distance) that determined the level of efficiency and precision of our process. It was found that the pair filter - segmenter formed by the Anscombe/Wiener filter together with the Fuzzy segmenter based on dynamic weights provided the best results, with VP and FP rates of 98.49 {+-}0.27% and 2.19 {+-}0.19%, respectively, for the simulation of the left ventricular volume. Along with the set of choices made during the processing structure, the project was finished with the analysis of a small number of volumes that belonged to a real PET test, thus the quantification of the volumes was obtained. (author)

  3. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  4. Nuclear dynamics investigation of the initial electron transfer in the cyclobutane pyrimidine dimer lesion repair process by photolyases

    CERN Document Server

    Joubert-Doriol, Loic; Olivucci, Massimo; Izmaylov, Artur F

    2016-01-01

    Photolyases are proteins capable of harvesting the sunlight to repair DNA damages caused by UV light. In this work we focus on the first step in the repair process of the cyclobutane pyrimidine dimer photoproduct (CPD) lesion, which is an electron transfer (ET) from a flavine cofactor to CPD, and study the role of various nuclear degrees of freedom (DOF) in this step. The ET step has been experimentally studied using transient spectroscopy and the corresponding data provide excellent basis for testing the quality of quantum dynamical models. Based on previous theoretical studies of electronic structure and conformations of the protein active site, we present a procedure to build a diabatic Hamiltonian for simulating the ET reaction in a molecular complex mimicking the enzyme's active site. We generate a reduced nuclear dimensional model that provides a first non-empirical quantum dynamical description of the structural features influencing the ET rate. By varying the nuclear DOF parametrization in the model t...

  5. Basic nuclear processes affected by histone acetyltransferases and histone deacetylase inhibitors

    NARCIS (Netherlands)

    Legartová, Soňa; Stixová, Lenka; Strnad, Hynek; Kozubek, Stanislav; Martinet, Nadine; Dekker, Frank J; Franek, Michal; Bártová, Eva

    2013-01-01

    AIM: The optimal balance between histone acetylation and deacetylation is important for proper gene function. Therefore, we addressed how inhibitors of histone-modifying enzymes can modulate nuclear events, including replication, transcription, splicing and DNA repair. MATERIALS & METHODS: Changes i

  6. Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

    Energy Technology Data Exchange (ETDEWEB)

    Steven R. Sherman

    2007-05-01

    The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant.

  7. Nuclear waste management. Quarterly progress report, January-March, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.; Powell, J.A. (comp.)

    1981-06-01

    Reports and summaries are provided for the following programs: high-level waste process development; alternative waste forms; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclide in soils; low-level waste generation reduction handbook; waste management system studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

  8. The BSC implanting process in a nuclear research center in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Cezar Augusto de; Guimaraes, Regia Ruth Ramirez; Filgueiras, Sergio Almeida Cunha [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Planejamento Estrategico e Qualidade - PE]. E-mail: cao@cdtn.br; rrrg@cdtn.br; sacf@cdtn.br

    2007-07-01

    The dynamics of the economical development founded due to the markets globalization and also to the increasing of the competition based on innovation, whose leadership belongs to the richest countries, presses for changes and moves the national and regional innovating systems. In a world of constant change, getting along with the external changes became one of the most relevant factors of the organizational success. Knowing and interpreting the external reality; monitoring the transformations; finding the opportunities and being able to answer fast and adequately; neutralize or minimize the threats: these and other abilities are constantly done by the most successful organizations, as part of a structured and conscious process focused on results . The technological research institutes were created in order to support the industries in their effort to overcome the competition by innovating. It is related, in last instance, to be an integrating part of the national or local innovating system, essential to the economical development and also to the improvement of life quality. However, they are put in this mutation atmosphere and fight for adapting to the new premises of the organizational success in order to have their mission fulfilled. In this context, the Development Center for Nuclear Technology - CDTN, makes an effort to adequate its strategically planning, by introducing and adapting the best administrating practices known nowadays. Among them, the Balanced Scorecard - BSC. This paper presents a brief form of each elaborating form of the strategic planning and also of the BSC implantation, it also clears up the level achieved by the organization and discusses the difficulties it faced. (author)

  9. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  10. Processes for separating the noble fission gases xenon and krypton from waste gases from nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Henrich, E.; Hufner, R.; Weirich, F.

    1983-08-23

    A process is claimed for separating the noble fission gases xenon and krypton from a prepurified waste gas from a nuclear plant. The prepurified waste gas is brought into contact with liquid Cl/sub 2/CF/sub 2/ as an absorption agent in a first column at an operating pressure which is less than or equal to normal pressure, whereby Xe, Kr, N/sub 2/O, CO/sub 2/, O/sub 2/ and N/sub 2/ are absorbed by the agent. Subsequently, the liquid absorption agent containing the absorbed gases is heated to substantially the boiling temperature of Cl/sub 2/CF/sub 2/ at the operating pressure for vaporizing part of the liquid absorption agent and desorbing the absorbed Kr, N/sub 2/ and O/sub 2/ to thereby separate the Kr and Xe from one another. The desorbed Kr, N/sub 2/ and O/sub 2/ gases are separated from the vaporized absorption agent. The liquid absorption agent which has not been vaporized is treated to recover Xe, N/sub 2/O and CO/sub 2/. Waste gas containing Kr, N/sub 2/ and O/sub 2/ from the head of the first column is brought into contact with liquid Cl/sub 2/CF/sub 2/ as an absorption agent in a second column, at an operating pressure which is less than or equal to normal pressure, whereby Kr, N/sub 2/ and O/sub 2/ are absorbed. Subsequently, the liquid absorption agent in the second column containing the absorbed Kr, N/sub 2/ and O/sub 2/ is heated substantially the boiling temperature of the Cl/sub 2/CF/sub 2/ at the operating pressure for vaporizing part of the liquid absorption agent and desorbing the absorbed N/sub 2/ and O/sub 2/. The liquid Cl/sub 2/CF/sub 2/ which has not been vaporized is treated to recover KR. An apparatus is provided for performing the process.

  11. Integrated Process Monitoring based on Systems of Sensors for Enhanced Nuclear Safeguards Sensitivity and Robustness

    Energy Technology Data Exchange (ETDEWEB)

    Humberto E. Garcia

    2014-07-01

    This paper illustrates safeguards benefits that process monitoring (PM) can have as a diversion deterrent and as a complementary safeguards measure to nuclear material accountancy (NMA). In order to infer the possible existence of proliferation-driven activities, the objective of NMA-based methods is often to statistically evaluate materials unaccounted for (MUF) computed by solving a given mass balance equation related to a material balance area (MBA) at every material balance period (MBP), a particular objective for a PM-based approach may be to statistically infer and evaluate anomalies unaccounted for (AUF) that may have occurred within a MBP. Although possibly being indicative of proliferation-driven activities, the detection and tracking of anomaly patterns is not trivial because some executed events may be unobservable or unreliably observed as others. The proposed similarity between NMA- and PM-based approaches is important as performance metrics utilized for evaluating NMA-based methods, such as detection probability (DP) and false alarm probability (FAP), can also be applied for assessing PM-based safeguards solutions. To this end, AUF count estimates can be translated into significant quantity (SQ) equivalents that may have been diverted within a given MBP. A diversion alarm is reported if this mass estimate is greater than or equal to the selected value for alarm level (AL), appropriately chosen to optimize DP and FAP based on the particular characteristics of the monitored MBA, the sensors utilized, and the data processing method employed for integrating and analyzing collected measurements. To illustrate the application of the proposed PM approach, a protracted diversion of Pu in a waste stream was selected based on incomplete fuel dissolution in a dissolver unit operation, as this diversion scenario is considered to be problematic for detection using NMA-based methods alone. Results demonstrate benefits of conducting PM under a system

  12. Studies of nuclear processes; Progress report, 1 September 1992--31 August 1993

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, E.J.

    1993-09-01

    Results for the period 1 Sep 92 through 31 Aug 93 are presented in nearly a hundred brief papers, some of which present new but preliminary data. Activities reported may be grouped as follows: Fundamental symmetries in the nucleus (parity-mixing measurements, time reversal invariance measurements, signatures of quantum chaos in nuclei), Internucleon reactions (neutron -- proton interactions, the neutron -- neutron scattering length, reactions between deuterons and very light nuclei), Dynamics of very light nuclei (measurements of D states of very light nuclei by transfer reactions, nuclear reactions between very light nuclei, radiative capture reactions with polarized sources), The many-nucleon problem (nuclear astrophysics, high-spin spectroscopy and superdeformation, the nuclear mean field: Dispersive relations and nucleon scattering, configuration mixing in {sup 56}Co and {sup 46}Sc using (d,{alpha}) reactions, radiative capture studies, high energy resolution resonance studies at 100--400 keV, nuclear data evaluation for A=3--20), Nuclear instruments and methods (FN tandem accelerator operation, KN accelerator operation and maintenance, atomic beam polarized ion source, development of techniques for determining the concentration of SF{sub 6} in the accelerator insulating gas mixture, production of beams and targets, detector systems, updating of TeX, Psprint, and associated programs on the VAX cluster), and Educational Activities.

  13. Design and operation of Taiyuan Iron & Steel' s sulphuric acid plant based on SO_2 from desulphurization and enrichment of sintering off-gas%太钢烧结烟气脱硫富集SO_2烟气制酸装置的设计与运行

    Institute of Scientific and Technical Information of China (English)

    涂瑞; 李强; 葛帅华

    2012-01-01

    太原钢铁(集团)有限公司炼铁厂2台烧结机烟气采用活性炭吸附法脱硫工艺。针对脱硫富集SO2烟气流量小、温度高、SO2浓度高、尘含量高并含有氟、氯、氨、汞等有害杂质的特点,烟气制酸设计采用喷淋塔—一级泡沫柱洗涤器—气体冷却塔—二级泡沫柱洗涤器—2级电除雾器稀酸洗净化、"3+1"ⅢⅠ-ⅣⅡ二转二吸工艺流程。2套制酸装置投产近1年时间,装置运行稳定,各项工艺指标均达到设计值,硫酸产量分别达到26,38 t/d,制酸尾气ρ(SO2)均小于或等于450 mg/m3,工业硫酸品质达到国家优等品标准。%The off-gas from two sintering machines was treated for desulphurization by activated carbon adsorption process in Iron Plant of Taiyuan Iron Steel.For the characteristics of low flow,high temperature,high SO2 and dust contents and existence of ammonia,fluorine,chlorine,mercury and other harmful impurities in desulphurization enriched off-gas,the plant adopted a cleaning section consisting of spray tower,1st stage froth column scrubber,gas cooling tower,2nd stage froth column scrubber and two-stage wet electrostatic precipitatores,and a "3+1" Ⅲ Ⅰ-Ⅳ Ⅱ double absorption process.The two plants have been operating for more than one year and every index reached design value,with tail gas SO2 concentration of both lower than 450 mg/m3,sulphuric acid production of 26 t/d and 38 t/d,respectively,and sulphuric acid quality up to the national premium grade requirement.

  14. The nuclear physics input to astrophysics modelling, and the r- and p-processes: Where do we stand 50 years after B^2FH and Cameron?

    Science.gov (United States)

    Arnould, M.

    2008-11-01

    This is a brief review of the progress made since the seminal contributions to the foundations of the theory of nucleosynthesis by M. Burbidge, G. Burbidge, Fowler and Hoyle, and by Cameron. The reviewed topics are (1) the nuclear physics input to the nucleosynthesis models (nuclear masses, fission, rates of β-decays, neutrino reactions, photoreactions, and nuclear charged particle-induced or neutron-induced reactions), (2) the nuclear physics and astrophysics aspects of the r-process, and (3) the same items for the p-process.

  15. Application of fuzzy neural network to the nuclear power plant in process fault diagnosis

    Institute of Scientific and Technical Information of China (English)

    LIU Yong-kuo; XIA Hong; XIE Chun-li

    2005-01-01

    The fuzzy logic and neural networks are combined in this paper,setting up the fuzzy neural network (FNN); meanwhile, the distinct differences and connections between the fuzzy logic and neural network are compared. Furthermore, the algorithm and structure of the FNN are introduced. In order to diagnose the faults of nuclear power plant, the FNN is applied to the nuclear power plant, and the intelligence fault diagnostic system of the nuclear power plant is built based on the FNN . The fault symptoms and the possibility of the inverted U-tube break accident of steam generator are discussed. In order to test the system's validity, the inverted U-tube break accident of steam generator is used as an example and many simulation experiments are performed. The test result shows that the FNN can identify the fault.

  16. Induction of micronuclei and nuclear abnormalities in Oreochromis niloticus following exposure to petroleum refinery and chromium processing plant effluents

    Energy Technology Data Exchange (ETDEWEB)

    Cavas, Tolga [Mersin University, Faculty of Sciences and Letters, Department of Biology, 33342 Mersin (Turkey)]. E-mail: tcavas@mersin.edu.tr; Ergene-Goezuekara, Serap [Mersin University, Faculty of Sciences and Letters, Department of Biology, 33342 Mersin (Turkey)

    2005-09-10

    The genotoxic effects of effluents from a petroleum refinery and a chromium processing plant were evaluated in Oreochromis niloticus (Pisces: Perciformes) using the micronucleus test. Fish were exposed to different concentrations (5, 10 and 20%, v/v) of the effluents for 3, 6 and 9 days. Micronucleus analyses were carried out on gill epithelial cells and peripheral blood erythrocytes. Nuclear abnormalities other than micronuclei, considered as genetic damage indicators, were also evaluated on erythrocytes. Cyclophosphamide at a single dose of 4 mg/L was used as a positive control. The results of this study showed that both effluents had genotoxic potential. On the other hand, the level of genetic damage induced by petroleum refinery effluent was considerably higher than that of chromium processing plant effluent. Our results further indicate that nuclear abnormalities other than micronuclei, such as blebbed and lobed nuclei, may also be used as indicators of genotoxic damage.

  17. Development of melt dilute technology for disposition of aluminum based spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swift, W.F. [Nuclear Material Management Division Westinghouse Savannah River Company, Savannah River Site Building 707-C, Aiken, SC 29808 (United States)

    2002-07-01

    The US Department of Energy (DOE) has for many years had a program for receipt and disposition of spent nuclear fuels of US origin from research reactors around the world. The research reactor spent nuclear fuel that consists of aluminum alloy composition has historically been returned to the Savannah River Site (SRS) and dispositioned via chemical reprocessing. In 1995, the DOE evaluated a number of alternatives to chemical reprocessing. In 2000, the DOE selected the melt-dilute alternative as the primary disposition path and direct disposal as the backup path. The melt-dilute technology has been developed from lab-scale demonstration up through the construction of a pilot-scale facility. The pilot-scale L-Area Experimental Facility (LEF) has been constructed and is ready for operation. The LEF will be used primarily, to confirm laboratory research on zeolite media for off- gas trapping and remote operability. Favorable results from the LEF are expected to lead to final design of the production melt-dilute facility identified as the Treatment and Storage Facility (TSF). This paper will describe the melt-dilute process and provide a status of the program development. (author)

  18. Development of the process of energy transfer from a nuclear Power Plant to an intermediate temperature electrolyse; Desarrollo del proceso de transferencia de energia desde una central nuclear a un electrolizador de temperatura intermedia

    Energy Technology Data Exchange (ETDEWEB)

    Munoz Cervantes, A.; Cuadrado Garcia, P.; Soraino Garcia, J.

    2013-07-01

    Fifty million tons of hydrogen are consumed annually in the world in various industrial processes. Among them, the ammonia production, oil refining and the production of methanol. One of the methods to produce it is the electrolysis of water, oxygen and hydrogen. This process needs electricity and steam which a central nuclear It can be your source; Hence the importance of developing the transfer process energy between the two. The objective of the study is to characterize the process of thermal energy transfer from a nuclear power plant to an electrolyzer of intermediate temperature (ITSE) already defined. The study is limited to the intermediate engineering process, from the central to the cell.

  19. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

    Energy Technology Data Exchange (ETDEWEB)

    Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

  20. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

    Energy Technology Data Exchange (ETDEWEB)

    Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

  1. 78 FR 33995 - Nuclear Proliferation Assessment in Licensing Process for Enrichment or Reprocessing Facilities

    Science.gov (United States)

    2013-06-06

    ...; denial. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying a petition for rulemaking (PRM), PRM-70-9, submitted by the American Physical Society (APS or the petitioner). The petitioner requested... members of the public. DATES: The docket for PRM-70-9 closed on June 6, 2013. ADDRESSES: Please refer...

  2. Sealed magic angle spinning nuclear magnetic resonance probe and process for spectroscopy of hazardous samples

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Herman M.; Washton, Nancy M.; Mueller, Karl T.; Sears, Jr., Jesse A.; Townsend, Mark R.; Ewing, James R.

    2016-06-14

    A magic-angle-spinning (MAS) nuclear magnetic resonance (NMR) probe is described that includes double containment enclosures configured to seal and contain hazardous samples for analysis. The probe is of a modular design that ensures containment of hazardous samples during sample analysis while preserving spin speeds for superior NMR performance and convenience of operation.

  3. Processing of solid solution, mixed uranium/refractory metal carbides for advanced space nuclear power and propulsion systems

    Science.gov (United States)

    Knight, Travis Warren

    Nuclear thermal propulsion (NTP) and space nuclear power are two enabling technologies for the manned exploration of space and the development of research outposts in space and on other planets such as Mars. Advanced carbide nuclear fuels have been proposed for application in space nuclear power and propulsion systems. This study examined the processing technologies and optimal parameters necessary to fabricate samples of single phase, solid solution, mixed uranium/refractory metal carbides. In particular, the pseudo-ternary carbide, UC-ZrC-NbC, system was examined with uranium metal mole fractions of 5% and 10% and corresponding uranium densities of 0.8 to 1.8 gU/cc. Efforts were directed to those methods that could produce simple geometry fuel elements or wafers such as those used to fabricate a Square Lattice Honeycomb (SLHC) fuel element and reactor core. Methods of cold uniaxial pressing, sintering by induction heating, and hot pressing by self-resistance heating were investigated. Solid solution, high density (low porosity) samples greater than 95% TD were processed by cold pressing at 150 MPa and sintering above 2600 K for times longer than 90 min. Some impurity oxide phases were noted in some samples attributed to residual gases in the furnace during processing. Also, some samples noted secondary phases of carbon and UC2 due to some hyperstoichiometric powder mixtures having carbon-to-metal ratios greater than one. In all, 33 mixed carbide samples were processed and analyzed with half bearing uranium as ternary carbides of UC-ZrC-NbC. Scanning electron microscopy, x-ray diffraction, and density measurements were used to characterize samples. Samples were processed from powders of the refractory mono-carbides and UC/UC 2 or from powders of uranium hydride (UH3), graphite, and refractory metal carbides to produce hypostoichiometric mixed carbides. Samples processed from the constituent carbide powders and sintered at temperatures above the melting point of UC

  4. Site selection process for new nuclear power plants - a method to support decision making and improving public participation

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Vivian B.; Cunha, Tatiana S. da; Simoes Filho, Francisco Fernando Lamego, E-mail: vbmartins@ien.gov.br, E-mail: flamego@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratorio de Impactos Ambientais; Lapa, Celso Marcelo F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares

    2011-07-01

    The Brazilian Energy Plan (PNE 2030) that guides the Government in formulating its strategy for expanding energy supply by 2030 highlights the need for the Brazilian electrical system have more than 4,000 MW from nuclear sources by 2025. Therefore, the Government presented a proposal to build four more nuclear power plants with capacity of 1,000 MW each, at first, two in the Northeast and two in Southeast. The selection and site assessment are key parts of the installation process of a nuclear plant and may significantly affect the cost, public acceptance and safety of the facility during its entire life cycle. The result of this initial stage, it can even seriously affect program success. Wrong decisions in the process of site selection may also require a financial commitment to higher planned in a later phase of the project, besides causing extensive and expensive downtime. Select the location where these units will be built is not a trivial process, because involves the consideration of multiple criteria and judgments in addition to obtaining, organizing and managing a diverse range of data, both qualitative and quantitative, to assist in decision making and ensure that the site selected is the most appropriate in relation to safety and technical, economic and environmental feasibility. This paper presents an overview of the site selection process and its stages, the criteria involved in each step, the tools to support decision making that can be used and the difficulties in applying a formal process of decision making. Also discussed are ways to make the process more transparent and democratic, increasing public involvement as a way to improve acceptance and reduce opposition from various sectors of society, trying to minimize the expense and time involved in the implementation of undertakings of this kind. (author)

  5. Idaho Nuclear Technology and Engineering Center Sodium-Bearing Waste Treatment Research and Development FY-2002 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Deldebbio, John Anthony; Mc Cray, John Alan; Kirkham, Robert John; Olson, Lonnie Gene; Scholes, Bradley Adams

    2002-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is considering several optional processes for disposal of liquid sodium-bearing waste. During fiscal year 2002, immobilization-related research included of grout formulation development for sodium-bearing waste, absorption of the waste on silica gel, and off-gas system mercury collection and breakthrough using activated carbon. Experimental results indicate that sodium-bearing waste can be immobilized in grout at 70 weight percent and onto silica gel at 74 weight percent. Furthermore, a loading of 11 weight percent mercury in sulfur-impregnated activated carbon was achieved with 99.8% off-gas mercury removal efficiency.

  6. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  7. Situation Concerning Public Information about and Involvement in the Decision-Making Processes in the Nuclear Sector. Public Opinion Review.

    Energy Technology Data Exchange (ETDEWEB)

    Prades, A.; Sala, R.; Lopez, M.

    2006-07-01

    This report summarizes the CIEMAT's contribution to the study {sup S}ituation concerning Public Information about and Involvement in the Decision-Making Processes in the Nuclear Sector{sup ,} contract number TREN/ 04/NUC/ S07.39556 between the European Commission and Mutadis Consultants. The research was composed by Mutadis Consultants and CEPN (Nuclear Protection Evaluation Centre) (France), University of Aberdeen (UK) and CIEMAT (Spain). The objective of the project was to build a detailed overview of the EU situation regarding information and participation practices in the nuclear domain, provide an elaborated assessment, and to produce reporting and recommendations in the field. CIEMAT contribution' focused on the review of public opinion polis. Thus, Eurobarometers Standard Surveys (EBs) were analysed to report about the European citizens' public opinion regarding public Information and participation in the nuclear field. Additionally, the International Social Survey Program (ISSP), and some additional national polis were analysed. In terms of the EU public opinion, the follow up of the public information and participation domains receiving as much attention as necessary. Extremely few questions dealing with the subject were identified in the Eurobarometers, the national polis and the ISSP (International Social Survey Program) surveys reviewed in this study. An unambiguous illustration of this lack of attention is the fact that no questions dealing with public participation issues emerged in the {sup n}uclear EBs{sup u}ntil 1998. Even though, Eurobarometers (EBs) still provide an invaluable source of information on the topics we are interested on at the EU allowing longitudinal descriptions (trend analysis) of some key issues in our area of interest. (Author) 11 refs.

  8. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

    2014-10-02

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

  9. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hui Zhang; Raman P. Singh

    2008-11-30

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  10. Public acceptability of the use of gamma rays from spent nuclear fuel as a hazardous waste treatment process

    Energy Technology Data Exchange (ETDEWEB)

    Mincher, B.J.; Wells, R.P.; Reilly, H.J.

    1992-01-01

    Three methods were used to estimate public reaction to the use of gamma irradiation of hazardous wastes as a hazardous waste treatment process. The gamma source of interest is spent nuclear fuel. The first method is Benefit-Risk Decision Making, where the benefits of the proposed technology are compared to its risks. The second analysis compares the proposed technology to the other, currently used nuclear technologies and estimates public reaction based on that comparison. The third analysis is called Analysis of Public Consent, and is based on the professional methods of the Institute for Participatory Management and Planning. The conclusion of all three methods is that the proposed technology should not result in negative public reaction sufficient to prevent implementation.

  11. Systematic approach for assessment of accident risks in chemical and nuclear processing; Abordagem sistematica para avaliacao de riscos de acidentes em instalacoes de processamento quimico e nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Senne Junior, Murillo

    2003-07-15

    The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of

  12. Some aspects of the nuclear fission process; Quelques aspects du processus de fission nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Netter, F. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U{sup 233}, U{sup 235}, Pu{sup 239}, U{sup 238} are described at the beginning of this work. It appears that for U{sup 233} there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U{sup 239} than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U{sup 235}. Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [French] Le present travail debute par un apercu de l'etat actuel de nos connaissances sur le processus de fission nucleaire, notamment sur le passage par le point-seuil. Puis sont evoques des aspects lies au niveau d'energie d'excitation auquel est porte le noyau qui subit la fission. Les mesures de sections efficaces de fission induite dans {sup 233}U, {sup 235}U, {sup 239}Pu et {sup 238}U par des neutrons rapides effectuees a Saclay sont decrites en premier lieu; elles font apparaitre pour {sup 233}U une ondulation caracteristique du role des etats collectifs d'excitation du noyau deforme au point-seuil. Des experiences sur la fission avec emission de particules de long parcours confirment cet aspect tout en demontrant que

  13. Nuclear Fission

    Science.gov (United States)

    Denschlag, J. O.

    This chapter first gives a survey on the history of the discovery of nuclear fission. It briefly presents the liquid-drop and shell models and their application to the fission process. The most important quantities accessible to experimental determination such as mass yields, nuclear charge distribution, prompt neutron emission, kinetic energy distribution, ternary fragment yields, angular distributions, and properties of fission isomers are presented as well as the instrumentation and techniques used for their measurement. The contribution concentrates on the fundamental aspects of nuclear fission. The practical aspects of nuclear fission are discussed in http://dx.doi.org/10.1007/978-1-4419-0720-2_57 of Vol. 6.

  14. Nuclear Safety

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1989-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  15. Report of the second meeting of the consultants on coupled processes associated with geological disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Tsang, Chin-Fu; Mangold, D.C.

    1985-09-01

    The second meeting of the Consultants on Coupled Processes Associated with Geological Disposal of Nuclear Waste occurred on January 15-16, 1985 at Lawrence Berkeley Laboratory (LBL). All the consultants were present except Dr. K. Kovari, who presented comments in writing afterward. This report contains a brief summary of the presentations and discussions from the meeting. The main points of the speakers' topics are briefly summarized in the report. Some points that emerged during the discussions of the presentations are included in the text related to the respective talks. These comments are grouped under the headings: Comments on Coupled Processes in Unsaturated Fractured Porous Media, Comments on Overview of Coupled Processes, Presentations by Consultants on Selected Topics of Current Interest in Coupled Processes, and Recommendations for Underground Field Tests with Applications to Three Geologic Environments.

  16. DECOVALEX III PROJECT. Mathematical Models of Coupled Thermal-Hydro-Mechanical Processes for Nuclear Waste Repositories. Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Jing, L.; Stephansson, O. [Royal Inst. of Technology, Stockholm (Sweden). Engineering Geology; Tsang, C.F. [Lawrence Berkely National Laboratory, Berkeley, CA (United States). Earth Science Div.; Mayor, J.C. [ENRESA, Madrid (Spain); Kautzky, F. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)] (eds.)

    2005-02-15

    DECOVALEX is an international consortium of governmental agencies associated with the disposal of high-level nuclear waste in a number of countries. The consortium's mission is the DEvelopment of COupled models and their VALidation against EXperiments. Hence the acronym/name DECOVALEX. Currently, agencies from Canada, Finland, France, Germany, Japan, Spain, Switzerland, Sweden, United Kingdom, and the United States are in DECOVALEX. Emplacement of nuclear waste in a repository in geologic media causes a number of physical processes to be intensified in the surrounding rock mass due to the decay heat from the waste. The four main processes of concern are thermal, hydrological, mechanical and chemical. Interactions or coupling between these heat-driven processes must be taken into account in modeling the performance of the repository for such modeling to be meaningful and reliable. DECOVALEX III is organized around four tasks. The FEBEX (Full-scale Engineered Barriers EXperiment) in situ experiment being conducted at the Grimsel site in Switzerland is to be simulated and analyzed in Task 1. Task 2, centered around the Drift Scale Test (DST) at Yucca Mountain in Nevada, USA, has several sub-tasks (Task 2A, Task 2B, Task 2C and Task 2D) to investigate a number of the coupled processes in the DST. Task 3 studies three benchmark problems: a) the effects of thermal-hydrologic-mechanical (THM) coupling on the performance of the near-field of a nuclear waste repository (BMT1); b) the effect of upscaling THM processes on the results of performance assessment (BMT2); and c) the effect of glaciation on rock mass behavior (BMT3). Task 4 is on the direct application of THM coupled process modeling in the performance assessment of nuclear waste repositories in geologic media. This executive summary presents the motivation, structure, objectives, approaches, and the highlights of the main achievements and outstanding issues of the tasks studied in the DECOVALEX III project

  17. Development of pyro-processing technology at CRIEPI for carving out the future of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Iizuka, M.; Koyama, T.; Sakamura, Y.; Uozumi, K.; Fujihata, K.; Kato, T.; Murakami, T.; Tsukada, T. [Central Research Institute of Electric Power Industry, Komae-shi, Tokyo 201-8511 (Japan); Glatz, J.P. [European Commission, JRC, Institute for Transuranium Elements (Germany)

    2013-07-01

    Pyro-processing has been attracting increasing attention as a promising candidate as an advanced nuclear fuel cycle technology. It provides economic advantage as well as reduction in proliferation risk and burden of long live radioactive waste, especially when it is combined with advanced fuels such as metallic or nitride fuel which gives excellent burning efficiency of minor actinides (MA). CRIEPI has been developing pyro-processing technology since late eighties with both domestic and international collaborations. In the early stage, electrochemical and thermodynamic properties in LiCl-KCl eutectic melt, and fundamental feasibility of core technology like electrorefining were chiefly investigated. Currently, stress in the process chemistry development is also placed on supporting technologies, such as treatment of anode residue and high temperature distillation for cathode product from electrorefining, and so on. Waste treatment process development, such as studies on adsorption behavior of various FP elements into zeolite and conditions for the fabrication of glass-bonded sodalite waste form, are steadily improved as well. In parallel, dedicated pyro-processing equipment such as zeolite column for treatment of spent electro-refiner salt is currently in progress. Recently, an integrated engineering-scale fuel cycle tests were performed funded by Japanese government (MEXT) as an important step before proceeding to large scale hot demonstration of pyro-processing. Oxide fuels can be readily introduced into the pyro-processing by reducing them to metals by adoption of electrochemical reduction technique. Making use of this advantage, the pyro-processing is currently under preliminary evaluation for its applicability to the treatment of the corium, mainly consisting of (U,Zr)O{sub 2}, formed in different composition during the accident of the Fukushima Daiichi nuclear power plant. (authors)

  18. On the robustness of the r-process in neutron-star mergers against variations of nuclear masses

    Science.gov (United States)

    Mendoza-Temis, J. J.; Wu, M. R.; Martínez-Pinedo, G.; Langanke, K.; Bauswein, A.; Janka, H.-T.; Frank, A.

    2016-07-01

    r-process calculations have been performed for matter ejected dynamically in neutron star mergers (NSM), such calculations are based on a complete set of trajectories from a three-dimensional relativistic smoothed particle hydrodynamic (SPH) simulation. Our calculations consider an extended nuclear reaction network, including spontaneous, β- and neutron-induced fission and adopting fission yield distributions from the ABLA code. In this contribution we have studied the sensitivity of the r-process abundances to nuclear masses by using diferent mass models for the calculation of neutron capture cross sections via the statistical model. Most of the trajectories, corresponding to 90% of the ejected mass, follow a relatively slow expansion allowing for all neutrons to be captured. The resulting abundances are very similar to each other and reproduce the general features of the observed r-process abundance (the second and third peaks, the rare-earth peak and the lead peak) for all mass models as they are mainly determined by the fission yields. We find distinct differences in the predictions of the mass models at and just above the third peak, which can be traced back to different predictions of neutron separation energies for r-process nuclei around neutron number N = 130.

  19. On-line testing of calibration of process instrumentation channels in nuclear power plants. Phase 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hashemian, H.M. [Analysis and Measurement Services Corp., Knoxville, TN (United States)

    1995-11-01

    The nuclear industry is interested in automating the calibration of process instrumentation channels; this report provides key results of one of the sponsored projects to determine the validity of automated calibrations. Conclusion is that the normal outputs of instrument channels in nuclear plants can be monitored over a fuel cycle while the plant is operating to determine calibration drift in the field sensors and associated signal conversion and signal conditioning equipment. The procedure for on-line calibration tests involving calculating the deviation of each instrument channel from the best estimate of the process parameter that the instrument is measuring. Methods were evaluated for determining the best estimate. Deviation of each signal from the best estimate is updated frequently while the plant is operating and plotted vs time for entire fuel cycle, thereby providing time history plots that can reveal channel drift and other anomalies. Any instrument channel that exceeds allowable drift or channel accuracy band is then scheduled for calibration during a refueling outage or sooner. This provides calibration test results at the process operating point, one of the most critical points of the channel operation. This should suffice for most narrow-range instruments, although the calibration of some instruments can be verified at other points throughout their range. It should be pointed out that the calibration of some process signals such as the high pressure coolant injection flow in BWRs, which are normally off- scale during plant operation, can not be tested on-line.

  20. Developing Process and Applying Prospects of Nuclear Power Plant%核电站的发展历程及应用前景

    Institute of Scientific and Technical Information of China (English)

    林宗虎

    2012-01-01

    自1954年苏联建成第1个核电站以来,全球已出现了为数众多装有各种反应堆的核电站.本文简述了反应堆的主要结构及核电站的发展历程.根据核电站的特点,论述了核电站在未来能源中的地位及应用前景.%Since the first nuclear power plant established by the So.viet Union in 1954. many nuclear power plants with different types of nuclear reactors have been presented on the world.Main constructions of nuclear reactors and the developing process of the nuclear power plant are briefly described in this paper. According to features of the nuclear power plant, its position in the future energy and applying prospects are also mentioned.

  1. How to have the shortest level emission of a furnace off-gas; Como alcanzar el nivel mas bajo de emision de un horno de arco electrico en el mundo

    Energy Technology Data Exchange (ETDEWEB)

    Naaby, H.; Abildgaards, O.; Pedersen, J.

    1996-06-01

    With a starting point in the requisite of an improved internal work environment and an expected demand for very low dust emission limits for Danish Steel works Ltd (DDS), the furnace off-gas from the two 110t EAF`s was led to a new filter system with a continuously registered exhaust through a stack. At the same time a survey was carried out regarding formation and possible removal of dioxin and furans in scrap based steel making. The results of the improvements have shown, that it is possible to reduce the dust emissions to less than 1 mg/m{sup 3} - STP (the local limit is 5 mg/m{sup 3} - STP) corresponding to app. 1 g./t. produced steel. The dioxin emission in less than lg./year, corresponding to 1500 TCDD-eq ng/t. produced steel. It has been proved that the metallurgy-, energy-, and furnace off-gas conditions are of the greatest importance for the lowering of dust and dioxin emissions, as well as for the optimization of the energy consumption. Therefore investigations of the energy conditions in the furnace are made and the furnace has been rebuild to ensure a sealed furnace. These improvements have reduced the quantity of gas to be filtered from the previous 90.000 m{sup 3} - STP/h to 40.000 m{sup 3} - STP/h from each furnace. It has also been shown, that a continuous monitoring and control of the consumption with approximately 10%. danish Steel Works Ltd. has led a project in an ECSC project which is going to find a economically and energy-wisw optimised way for further reduction of the dioxin emission. the results of the latest improvements are reported in this paper. (Author) 4 refs.

  2. Biosorption of Strontium from Simulated Nuclear Wastewater by Scenedesmus spinosus under Culture Conditions: Adsorption and Bioaccumulation Processes and Models

    Directory of Open Access Journals (Sweden)

    Mingxue Liu

    2014-06-01

    Full Text Available Algae biosorption is an ideal wastewater treatment method when coupled with algae growth and biosorption. The adsorption and bioaccumulation of strontium from simulated nuclear wastewater by Scenedesmus spinosus were investigated in this research. One hundred mL of cultured S. spinosus cells with a dry weight of 1.0 mg in simulated nuclear wastewater were used to analyze the effects on S. spinosus cell growth as well as the adsorption and bioaccumulation characters under conditions of 25 ± 1 °C with approximately 3,000 lux illumination. The results showed that S. spinosus had a highly selective biosorption capacity for strontium, with a maximum bioremoval ratio of 76%. The adsorbed strontium ion on cell walls was approximately 90% of the total adsorbed amount; the bioaccumulation in the cytoplasm varied by approximately10%. The adsorption quantity could be described with an equilibrium isotherm. The pseudo-second-order kinetic model suggested that adsorption was the rate-limiting step of the biosorption process. A new bioaccumulation model with three parameters was proposed and could give a good fit with the experiment data. The results suggested that S. spinosus may be a potential biosorbent for the treatment of nuclear wastewater in culture conditions.

  3. Biosorption of Strontium from Simulated Nuclear Wastewater by Scenedesmus spinosus under Culture Conditions: Adsorption and Bioaccumulation Processes and Models

    Science.gov (United States)

    Liu, Mingxue; Dong, Faqin; Kang, Wu; Sun, Shiyong; Wei, Hongfu; Zhang, Wei; Nie, Xiaoqin; Guo, Yuting; Huang, Ting; Liu, Yuanyuan

    2014-01-01

    Algae biosorption is an ideal wastewater treatment method when coupled with algae growth and biosorption. The adsorption and bioaccumulation of strontium from simulated nuclear wastewater by Scenedesmus spinosus were investigated in this research. One hundred mL of cultured S. spinosus cells with a dry weight of 1.0 mg in simulated nuclear wastewater were used to analyze the effects on S. spinosus cell growth as well as the adsorption and bioaccumulation characters under conditions of 25 ± 1 °C with approximately 3,000 lux illumination. The results showed that S. spinosus had a highly selective biosorption capacity for strontium, with a maximum bioremoval ratio of 76%. The adsorbed strontium ion on cell walls was approximately 90% of the total adsorbed amount; the bioaccumulation in the cytoplasm varied by approximately10%. The adsorption quantity could be described with an equilibrium isotherm. The pseudo-second-order kinetic model suggested that adsorption was the rate-limiting step of the biosorption process. A new bioaccumulation model with three parameters was proposed and could give a good fit with the experiment data. The results suggested that S. spinosus may be a potential biosorbent for the treatment of nuclear wastewater in culture conditions. PMID:24919131

  4. Measurement of the energy spectrum of {sup 252}Cf fission fragments using nuclear track detectors and digital image processing

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, G.; Golzarri, J. I. [UNAM, Instituto de Fisica, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Castano, V. M. [UNAM, Centro de Fisica Aplicada y Tecnologia Avanzada, Boulevard Juriquilla 3001, Santiago de Queretaro, 76230 Queretaro (Mexico); Gaso, I. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Mena, M.; Segovia, N. [UNAM, Instituto de Geofisica, Circuito de la Investigacion Cientifica, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2010-02-15

    The energy spectrum of {sup 252}Cf fission fragments was measured using nuclear track detectors and digital image analysis system. The detection material was fused silica glass. The detectors were chemically etched in an 8% HF solution. After experimenting with various etching time, it was found that the best resolution of the track diameter distribution was obtained after 30 minutes of etching. Both Gaussian and Lorentzian curves were fit to the track diameter distribution histograms and used to determine the basic parameters of the distribution of the light (N{sub L}) and heavy (N{sub H}) formed peaks and the minimum of the central valley (N{sub V}). Advantages of the method presented here include the fully-automated analysis process, the low cost of the nuclear track detectors and the simplicity of the nuclear track method. The distribution resolution obtained by this method is comparable with the resolution obtained by electronic analysis devices. The descriptive variables calculated were very close to those obtained by other methods based on the use of semiconductor detectors. (Author)

  5. Comment on "138La-138Ce-136Ce nuclear cosmochronometer of the supernova neutrino process"

    CERN Document Server

    Von Neumann-Cosel, P; Byelikov, A

    2009-01-01

    The nuclear chosmochronometer suggested by Hayakawa et al. [Phys. Rev.C 77, 065802 (2008)] based on the 138La-138Ce-136Ce abundance ratio in presolar grains would be affected by the existence of a hitherto unknown low-energy 1+ state in 138La. Results of a recent high-resolution study of the 138Ba(3He,t) reaction under kinematics selectively populating 1+ states in 138La through Gamow-Teller transitions provides strong evidence against the existence of such a hypothetical state.

  6. A feasibility study on regulatory analysis in the nuclear regulatory decision making process in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. W.; Lee, D. K.; Lim, C. Y.; Choi, Y. S.; Kim, K. K. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-12-15

    Nuclear safety regulation has different implications from economic regulation. In recent survey, it was found that Korea people want more regulatory effort than current one. This indicates that scientific method is necessary to improve the efficiency of regulation with limited resources of regulators. The main reason to conduct the regulatory analysis like cost-benefit analysis is obtaining the rationality of regulation to be implemented and persuading the licensees to comply under the quantified condition. That is, the ultimate goal is to verify the necessity and justice of regulation and to achieve the safety goal with minimum impacts. Guidelines and procedures developed in this study should help regulators to reach that goal.

  7. Survey of potential process-heat and reject-heat utilization at a Green River nuclear-energy center

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.M.; Sandquist, G.M.

    1982-03-01

    Potential uses of process heat and reject heat from a nuclear-energy center at Green River, Utah have been investigated. The remoteness of the Green River site would preclude many potential industrial uses for economical reasons such as transportation costs and lack of local markets. Water-consumption requirements would also have serious impact on some applications due to limitations imposed by other contractual agreements upon the water in the region. Several processes were identified which could be considered for the Green River site; including the use of heat to separate bitumens from tar sands, district heating, warming of greenhouses and soil, and the production of fish for game and commercial sales. The size of these industries would be limited and no single process or industry can be identified at this time which could use the full amount of low-temperature reject heat that would be generated at a NEC.

  8. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Rogers, Adam Zachary; Simmons, R. F.; Palethorpe, S. J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  9. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  10. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

    1999-10-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  11. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

    1999-09-30

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  12. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

    2000-11-01

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  13. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

    2000-10-31

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  14. THE TESTING OF COMMERCIALLY AVAILABLE ENGINEERING AND PLANT SCALE ANNULAR CENTRIFUGAL CONTACTORS FOR THE PROCESSING OF SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Jack D. Law; David Meikrantz; Troy Garn; Nick Mann; Scott Herbst

    2006-10-01

    Annular centrifugal contactors are being evaluated for process scale solvent extraction operations in support of United State Advanced Fuel Cycle Initiative goals. These contactors have the potential for high stage efficiency if properly employed and optimized for the application. Commercially available centrifugal contactors are being tested at the Idaho National Laboratory to support this program. Hydraulic performance and mass transfer efficiency have been measured for portions of an advanced nuclear fuel cycle using 5-cm diameter annular centrifugal contactors. Advanced features, including low mix sleeves and clean-in-place rotors, have also been evaluated in 5-cm and 12.5-cm contactors.

  15. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  16. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; McCray, J.A.; Rogers, A.Z.; Simmons, R.F.; Palethrope, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  17. Design process of the nanofluid injection mechanism in nuclear power plants.

    Science.gov (United States)

    Kang, Myoung-Suk; Jee, Changhyun; Park, Sangjun; Bang, In Choel; Heo, Gyunyoung

    2011-04-27

    Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs). This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs) assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD) in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

  18. Design process of the nanofluid injection mechanism in nuclear power plants

    Directory of Open Access Journals (Sweden)

    Bang In Choel

    2011-01-01

    Full Text Available Abstract Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs. This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

  19. Oxidative dissolution of spent nuclear fuel in aqueous alkaline solutions - An alternative to the Purex process?

    Energy Technology Data Exchange (ETDEWEB)

    Runde, Wolfgang; Peper, Shane; Brodnax, Lia; Crooks, William; Zehnder, Ralph; Jarvinen, Gordon

    2004-07-01

    As an alternative to acidic reprocessing of spent nuclear, oxidative dissolution of UO{sub 2} into aqueous alkaline solutions and subsequent separation of fission products is considered. The efficacy of such a method is limited by the kinetics of the UO{sub 2} dissolution and the capacity of alkaline solutions for dissolved U(VI) species. We performed a series of dissolution studies on UO{sub 2} and U{sub 3}O{sub 8} in aqueous alkaline solutions applying various oxidants. Among the oxidative agents commonly used to transform low-valence actinides into their higher oxidation states, H{sub 2}O{sub 2} has proven to be the most effective in basic media. Consequently, we investigated the dissolution of UO{sub 2} and U{sub 3}O{sub 8} in NaOH-H{sub 2}O{sub 2} and Na{sub 2}CO{sub 3}-H{sub 2}O{sub 2} solutions and determined the dissolution kinetics as a function of peroxide and hydroxide (carbonate) concentrations. Methods to remove fission products, e.g., Cs, Sr, Ba and Zr, from alkaline solutions will be evaluated based upon their decontamination factors. We will discuss the feasibility of using chemically oxidizing alkaline solutions as an alternative spent nuclear fuel reprocessing method based on results from experimental quantitative investigations. (authors)

  20. Automatic data processing and analysis system for monitoring region around a planned nuclear power plant

    Science.gov (United States)

    Kortström, Jari; Tiira, Timo; Kaisko, Outi

    2016-03-01

    The Institute of Seismology of University of Helsinki is building a new local seismic network, called OBF network, around planned nuclear power plant in Northern Ostrobothnia, Finland. The network will consist of nine new stations and one existing station. The network should be dense enough to provide azimuthal coverage better than 180° and automatic detection capability down to ML -0.1 within a radius of 25 km from the site.The network construction work began in 2012 and the first four stations started operation at the end of May 2013. We applied an automatic seismic signal detection and event location system to a network of 13 stations consisting of the four new stations and the nearest stations of Finnish and Swedish national seismic networks. Between the end of May and December 2013 the network detected 214 events inside the predefined area of 50 km radius surrounding the planned nuclear power plant site. Of those detections, 120 were identified as spurious events. A total of 74 events were associated with known quarries and mining areas. The average location error, calculated as a difference between the announced location from environment authorities and companies and the automatic location, was 2.9 km. During the same time period eight earthquakes between magnitude range 0.1-1.0 occurred within the area. Of these seven could be automatically detected. The results from the phase 1 stations of the OBF network indicates that the planned network can achieve its goals.

  1. 78 FR 47012 - Developing Software Life Cycle Processes Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2013-08-02

    ..., ``IEEE Standard for Developing a Software Project Life Cycle Process,'' issued 2006, with the... guidance in IEEE Std. 1074- 2006, ``IEEE Standard for Developing a Software Project Life Cycle Process... to systems and related quality assurance processes, and the requirements also extend to the...

  2. Supporting Technology for Chain of Custody of Nuclear Weapons and Materials throughout the Dismantlement and Disposition Processes

    Energy Technology Data Exchange (ETDEWEB)

    Bunch, Kyle J.; Jones, Anthony M.; Ramuhalli, Pradeep; Benz, Jacob M.; Denlinger, Laura S.

    2014-05-04

    The ratification and ongoing implementation of the New START Treaty have been widely regarded as noteworthy global security achievements for both the Obama Administration and the Putin (formerly Medvedev) regime. But deeper cuts that move beyond the United States and Russia to engage the P-5 and other nuclear weapons possessor states are envisioned under future arms control regimes, and are indeed required for the P-5 in accordance with their Article VI disarmament obligations in the Nuclear Non-Proliferation Treaty. Future verification needs will include monitoring the cessation of production of new fissile material for weapons, monitoring storage of warhead components and fissile materials and verifying dismantlement of warheads, pits, secondary stages, and other materials. A fundamental challenge to implementing a nuclear disarmament regime is the ability to thwart unauthorized material diversion throughout the dismantlement and disposition process through strong chain of custody implementation. Verifying the declared presence, or absence, of nuclear materials and weapons components throughout the dismantlement and disposition lifecycle is a critical aspect of the disarmament process. From both the diplomatic and technical perspectives, verification under these future arms control regimes will require new solutions. Since any acceptable verification technology must protect sensitive design information and attributes to prevent the release of classified or other proliferation-sensitive information, non-nuclear non-sensitive modalities may provide significant new verification tools which do not require the use of additional information barriers. Alternative verification technologies based upon electromagnetic and acoustics could potentially play an important role in fulfilling the challenging requirements of future verification regimes. For example, researchers at the Pacific Northwest National Laboratory (PNNL) have demonstrated that low frequency electromagnetic

  3. Treatment process for a surface-active nitric acid solution used for rinsing nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Celles, Caroline; Bisel, Isabelle; Gastaldi, Michel; Rudloff, David [CEA Valrho/DEN/VRH/DRCP/SCPS/LPCP, BP 17171, Bagnols sur Ceze cedex, 30207 (France)

    2008-07-01

    Full text of publication follows: The rinsing of the nuclear fuel reprocessing plants before their final shut-down and their dismantling is considered by the use of surface-active compounds in nitric acid medium. Before being vitrified, this solution must be mineralized and concentrated. Two treatments ways have been studied: The first was to act a direct concentration followed up by a mineralization through the Fenton reaction. The second way proposed to begin with the mineralization followed up by a concentration with denitration. The second way was not kept since laboratory tests showed organic phase demixion in distillate which is difficult to manage in industrial plants. The study presents the performance Fenton reaction with a new surface-active solution. In order to be representative of real solution in these nuclear installations to rinse, some hypothesis have been established concerning cations and TBP concentrations, likely to be present in the installation. Influences of parameters like temperature, the nature of the Fenton catalyst, the catalyst concentration, and the concentration of surface-active species were tested. The study highlight the kinetics of the mineralization, the H{sub 2}O{sub 2} consumption, the degradation products of surface-active compounds and TBP. Finally optimized operating conditions have been proposed and a reference test has been performed showing a mineralization out put of about 90% for 167 hours of reaction. Added H{sub 2}O{sub 2} leads to dilute the medium by a factor 2. The main degradation products obtained at the end of reaction are acetic, formic and phosphoric acids. Contrary to acetic and formic acid which will probably be destroyed during following concentration, phosphoric acid behaviour has to be studied in plutonium containing solutions. This study is still in progress. (authors)

  4. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Science.gov (United States)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  5. Analytical and experimental stiffness estimation of heat pipe supporter for nuclear power plant through a homogenization process

    Directory of Open Access Journals (Sweden)

    Sang-Young Kim

    2015-07-01

    Full Text Available This article aims to study the in-plane stiffness estimation of heat pipe supporter (a large lattice structure using experimental and numerical methods. The in-plane stiffness of heat pipe supporter for nuclear power plant is very important because of the safety against natural disasters, such as seismic load or tsunami, and has to be evaluated because it greatly affects the durability of the heat exchanger. However, the modeling process of the whole lattice structure for finite element analysis increases resources needed caused by too many nodes and elements. In this study, the mechanical properties of large lattice structures are determined by a unit cell finite element analysis. The mechanical behavior of a large lattice structure has been estimated by finite element analysis through a homogenization process for reducing analysis time and efforts. The finite element analysis results have been verified and show a good agreement with the experimental results.

  6. Relation between nuclear envelope and nuclear lamina in nuclear assembly in vitro

    Institute of Scientific and Technical Information of China (English)

    蔡树涛; 翟中和

    1997-01-01

    Xenopus laevis egg extracts cell-free nuclear assembly system was used as an experimental model to study the process of nuclear lamina assembly in nuclear reconstitution in vitro. The experimental results showed that lamin was involved in the nuclear assembly in vitro. The assembly of nuclear lamina was preceded by the assembly of nuclear matrix, and probably, inner nuclear matrix assembly provided the basis for nuclear lamina assembly. Inhibition of normal assembly of nuclear lamina, by preincubating egg extracts cell-free system with anti-lamin antibodies, resulted in abnormal assembly of nuclear envelope, suggesting that nuclear envelope assembly is closely associated with nuclear lamina assembly.

  7. Final disposal of spent nuclear fuel in Sweden. Some unresolved issues and challenges in the design and implementation of the forthcoming planning and EIA processes

    Energy Technology Data Exchange (ETDEWEB)

    Bjarnadottir, H.; Hilding-Rydevik, T. [Nordregio, Stockholm (Sweden)

    2001-06-01

    The aim of the study is to highlight some unresolved and challenging issues in the forthcoming approximately six year long Environmental Impact Assessment (EIA) and planning process of the long-term disposal of spent nuclear fuel in Sweden. Different international and Nordic experiences of the processes for final disposal as well as from other development of similar scope, where experiences assumed to be of importance for final disposal of nuclear waste, have been described. Furthermore, issues relating to 'good EIA practice' as well as certain aspects of planning theory have also been presented. The current Swedish situation for the planning and EIA process of the final disposal of spent nuclear fuel was also been summarized. These different 'knowledge areas' have been compared and measured against our perception of the expectations towards the forthcoming process, put forward by different Swedish actors in the field. The result is a presentation of a number of questions and identification issues that the authors consider need special attention in the design and conduction of the planning and EIA process. The study has been realized through a literature survey and followed by reading and analysis of the written material. The main focus of the literature search was on material describing planning processes, actor perspectives and EIA. Material and literature on the technical and scientific aspects of spent nuclear fuel disposal was however deliberately avoided. There is a wealth of international and Swedish literature concerning final disposal of spent nuclear fuel - concerning both technical issues and issues concerning for example public participation and risk perception. But material of a more systematic and comparative nature (relating to both empirical and theoretical issues, and to practical experiences) in relation to EIA processes and communicative planning for final disposal of spent nuclear fuel seems to be more sparsely represented

  8. Data acquisition and management system for a nuclear processes simulator; Sistema de adquisicion y manejo de datos para un simulador de procesos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.A.; Santiago C, J. [Facultad de Ingenieria, UNAM, Campus Morelos (Mexico)]. e-mail: cchavez2@cableonline.com.mx

    2003-07-01

    In the development of this work (Data acquisition and management system for a simulator of nuclear processes (SAMAD)), is important to mention the main modules that involve the operation of the same one. At the beginning it was necessary to contemplate the possible programming languages, as well as the compatibility and handling easiness among them. The used languages to be able to land the contemplated ideas are: C{sup ++}, PHP, HTML, as well as the My SQL database manager. After this it was designed the database (DB), which contains the tables of each one of the components, this according to the enter file type of the RELAP5 code that will be use for each simulation, as well as, tables that will allow us to relate and to maintain the control of the information supplied to the DB. Once created the database is interacting with it through an application program based on PHP (Preprocessor). The application basically consists, in extracting the data from each one of the components to work in that moment, that is to say, to obtain the data of the enter file, as well as, to depurate the data, excluding comments. The preprocessor gives bigger easiness to place the data in the DB. Also, it was developed an graphic interface that allows to register variables to the DB, depending from the unfolding to visualize. Another application that has been implemented is the Data Collector that has as function, to obtain in a direct way the data of the display variables of the RELAP5 code, with the purpose of storing them in the DB, this will be carry out in real time and it was updated in a very small time period. (Author)

  9. Thermodynamic study of residual heat from a high temperature nuclear reactor to analyze its viability in cogeneration processes; Estudio termodinamico del calor residual de un reactor nuclear de alta temperatura para analizar su viabilidad en procesos de cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Santillan R, A.; Valle H, J.; Escalante, J. A., E-mail: santillanaura@gmail.com [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2015-09-15

    In this paper the thermodynamic study of a nuclear power plant of high temperature at gas turbine (GTHTR300) is presented for estimating the exploitable waste heat in a process of desalination of seawater. One of the most studied and viable sustainable energy for the production of electricity, without the emission of greenhouse gases, is the nuclear energy. The fourth generation nuclear power plants have greater advantages than those currently installed plants; these advantages have to do with security, increased efficiencies and feasibility to be coupled to electrical cogeneration processes. In this paper the thermodynamic study of a nuclear power plant type GTHTR300 is realized, which is selected by greater efficiencies and have optimal conditions for use in electrical cogeneration processes due to high operating temperatures, which are between 700 and 950 degrees Celsius. The aim of the study is to determine the heat losses and the work done at each stage of the system, determining where they are the greatest losses and analyzing in that processes can be taken advantage. Based on the study was appointed that most of the energy losses are in form of heat in the coolers and usually this is emitted into the atmosphere without being used. From the results a process of desalination of seawater as electrical cogeneration process is proposed. This paper contains a brief description of the operation of the nuclear power plant, focusing on operation conditions and thermodynamic characteristics for the implementation of electrical cogeneration process, a thermodynamic analysis based on mass and energy balance was developed. The results allow quantifying the losses of thermal energy and determining the optimal section for coupling of the reactor with the desalination process, seeking to have a great overall efficiency. (Author)

  10. Rhenium solubility in borosilicate nuclear waste glass: implications for the processing and immobilization of technetium-99.

    Science.gov (United States)

    McCloy, John S; Riley, Brian J; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J; Rodriguez, Carmen P; Hrma, Pavel; Kim, Dong-Sang; Lukens, Wayne W; Kruger, Albert A

    2012-11-20

    The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm [by mass, added as KReO(4) (Re(7+))] were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.

  11. Time-of-flight mass measurements for nuclear processes in neutron star crusts

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Alfredo [National Superconducting Cyclotron Laboratory (NSCL); Matos, M. [Louisiana State University; Schatz, Hendrik [Michigan State University, East Lansing; Amthor, A. M. [National Superconducting Cyclotron Laboratory (NSCL); Bazin, D. [National Superconducting Cyclotron Laboratory (NSCL); Beard, Mary [University of Notre Dame, IN; Becerril, A. [National Superconducting Cyclotron Laboratory (NSCL); Brown, Edward [Michigan State University, East Lansing; Elliot, T [National Superconducting Cyclotron Laboratory (NSCL); Gade, A. [National Superconducting Cyclotron Laboratory (NSCL); Galaviz, D. [National Superconducting Cyclotron Laboratory (NSCL); George, S. [National Superconducting Cyclotron Laboratory (NSCL); Gupta, Sanjib [Indian Institute of Technology, Kanpur; Hix, William Raphael [ORNL; Lau, Rita [National Superconducting Cyclotron Laboratory (NSCL); Moeller, Peter [Los Alamos National Laboratory (LANL); Pereira, J. [National Superconducting Cyclotron Laboratory (NSCL); Portillo, M. [National Superconducting Cyclotron Laboratory (NSCL); Rogers, A. M. [National Superconducting Cyclotron Laboratory (NSCL); Shapira, Dan [ORNL; Smith, E. [Ohio State University; Stolz, A. [Michigan State University, East Lansing; Wallace, M. [Los Alamos National Laboratory (LANL); Wiescher, Michael [University of Notre Dame, IN

    2011-01-01

    The location of electron capture heat sources in the crust of accreting neutron stars depends on the masses of extremely neutron-rich nuclei. We present first results from a new implementation of the time-of-flight technique to measure nuclear masses of rare isotopes at the National Supercon- ducting Cyclotron Laboratory. The masses of 16 neutron-rich nuclei in the Sc Ni element range were determined simultaneously, improving the accuracy compared to previous data in 12 cases. The masses of 61V, 63Cr, 66Mn, and 74Ni were measured for the first time with mass excesses of 30.510(890) MeV, 35.280(650) MeV, 36.900(790) MeV, and 49.210(990) MeV, respectively. With the measurement of the 66Mn mass, the location of the two dominant heat sources in the outer crust of accreting neutron stars, which exhibit so called superbursts, is now experimentally constrained. We find that the location of the 66Fe 66Mn electron capture transition occurs sig- nificantly closer to the surface than previously assumed because our new experimental Q-value is 2.1 MeV smaller than predicted by the FRDM mass model. The results also provide new insights into the structure of neutron-rich nuclei around N = 40.

  12. Time-of-flight mass measurements for nuclear processes in neutron star crusts

    CERN Document Server

    Estrade, A; Schatz, H; Amthor, A M; Bazin, D; Beard, M; Becerril, A; Brown, E F; Cyburt, R; Elliot, T; Gade, A; Galaviz, D; George, S; Gupta, S S; Hix, W R; Lau, R; Lorusso, G; Moller, P; Pereira, J; Portillo, M; Rogers, A M; Shapira, D; Smith, E; Stolz, A; Wallace, M; Wiescher, M

    2011-01-01

    The location of electron capture heat sources in the crust of accreting neutron stars depends on the masses of extremely neutron-rich nuclei. We present first results from a new implementation of the time-of-flight technique to measure nuclear masses of rare isotopes at the National Superconducting Cyclotron Laboratory. The masses of 16 neutron-rich nuclei in the scandium -- nickel range were determined simultaneously, improving the accuracy compared to previous data in 12 cases. The masses of $^{61}${V}, $^{63}${Cr}, $^{66}${Mn}, and $^{74}${Ni} were measured for the first time with mass excesses of $-30.510(890)$ MeV, $-35.280(650)$ MeV, $-36.900(790)$ MeV, and $-49.210(990)$ MeV, respectively. With the measurement of the $^{66}$Mn mass, the locations of the two dominant electron capture heat sources in the outer crust of accreting neutron stars that exhibit superbursts are now experimentally constrained. We find that the location of the $^{66}$Fe$\\rightarrow^{66}$Mn electron capture transition occurs signi...

  13. Time-of-Flight Mass Measurements for Nuclear Processes in Neutron Star Crusts

    Science.gov (United States)

    Estradé, A.; Matoš, M.; Schatz, H.; Amthor, A. M.; Bazin, D.; Beard, M.; Becerril, A.; Brown, E. F.; Cyburt, R.; Elliot, T.; Gade, A.; Galaviz, D.; George, S.; Gupta, S. S.; Hix, W. R.; Lau, R.; Lorusso, G.; Möller, P.; Pereira, J.; Portillo, M.; Rogers, A. M.; Shapira, D.; Smith, E.; Stolz, A.; Wallace, M.; Wiescher, M.

    2011-10-01

    We present results from time-of-flight nuclear mass measurements at the National Superconducting Cyclotron Laboratory that are relevant for neutron star crust models. The masses of 16 neutron-rich nuclei in the scandium-nickel range were determined simultaneously, with the masses of V61, Cr63, Mn66, and Ni74 measured for the first time with mass excesses of -30.510(890)MeV, -35.280(650)MeV, -36.900(790)MeV, and -49.210(990)MeV, respectively. With these results the locations of the dominant electron capture heat sources in the outer crust of accreting neutron stars that exhibit super bursts are now experimentally constrained. We find the experimental Q value for the Fe66→Mn66 electron capture to be 2.1 MeV (2.6σ) smaller than predicted, resulting in the transition occurring significantly closer to the neutron star surface.

  14. An ARGONAUTE4-containing nuclear processing center colocalized with Cajal bodies in Arabidopsis thaliana.

    Science.gov (United States)

    Li, Carey Fei; Pontes, Olga; El-Shami, Mahmoud; Henderson, Ian R; Bernatavichute, Yana V; Chan, Simon W-L; Lagrange, Thierry; Pikaard, Craig S; Jacobsen, Steven E

    2006-07-14

    ARGONAUTE4 (AGO4) and RNA polymerase IV (Pol IV) are required for DNA methylation guided by 24 nucleotide small interfering RNAs (siRNAs) in Arabidopsis thaliana. Here we show that AGO4 localizes to nucleolus-associated bodies along with the Pol IV subunit NRPD1b; the small nuclear RNA (snRNA) binding protein SmD3; and two markers of Cajal bodies, trimethylguanosine-capped snRNAs and the U2 snRNA binding protein U2B''. AGO4 interacts with the C-terminal domain of NRPD1b, and AGO4 protein stability depends on upstream factors that synthesize siRNAs. AGO4 is also found, along with the DNA methyltransferase DRM2, throughout the nucleus at presumed DNA methylation target sites. Cajal bodies are conserved sites for the maturation of ribonucleoprotein complexes. Our results suggest a function for Cajal bodies as a center for the assembly of an AGO4/NRPD1b/siRNA complex, facilitating its function in RNA-directed gene silencing at target loci.

  15. Priorities for modeling biological processes in climates altered by nuclear war

    Energy Technology Data Exchange (ETDEWEB)

    Detling, J.K.; Kercher, J.R.; Post, W.M.; Cowles, S.W.; Harwell, M.A.

    1987-01-01

    This document describes research that has been accomplished or currently models the effects of reduced light and temperature on terrestrial systems. We shall divide the systems to be studied into cultivated lands and uncultivated lands. The cultivated class consists of monoculture systems in which the individual plants belong to the same age and size class. The systems in the uncultivated class consist of uneven age, multi-species assemblies of interacting plants and animals. The uncultivated class ranges from minimally managed systems, e.g., rangelands and some forests, to completely unmanaged wildlands. For the cultivated case, the variable of concern is the annual yield of the crop under consideration. The models should be able to estimate percent yield loss as a function of reductions of light and temperature. The models should be accurate for the range of environments predicted for the growing season immediately following or during which the hypothetical nuclear exchange occurs. The models should be able to estimate yield loss in any subsequent year for which climatic conditions still differ significantly from normal. For the uncultivated case, the modelling program needs to be able to predict the effects on individual plants much the same as in the cultivated case; but in addition, the modelling program will have the task of estimating the effect that these changes in individual organisms will have at higher levels of organization, i.e., on populations, communities, and regional distributions of species. 25 refs., 1 tab.

  16. Removal of Radiocesium from Food by Processing: Data Collected after the Fukushima Daiichi Nuclear Power Plant Accident - 13167

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shigeo; Tagami, Keiko [Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences, Anagawa 4-9-1, Inage-ku, Chiba 263-8555 (Japan)

    2013-07-01

    Removal of radiocesium from food by processing is of great concern following the accident of TEPCO's Fukushima Daiichi Nuclear Power Plant accident. Foods in markets are monitored and recent monitoring results have shown that almost all food materials were under the standard limit concentration levels for radiocesium (Cs-134+137), that is, 100 Bq kg{sup -1} in raw foods, 50 Bq kg{sup -1} in baby foods, and 10 Bq kg{sup -1} in drinking water; those food materials above the limit cannot be sold. However, one of the most frequently asked questions from the public is how much radiocesium in food would be removed by processing. Hence, information about radioactivity removal by processing of food crops native to Japan is actively sought by consumers. In this study, the food processing retention factor, F{sub r}, which is expressed as total activity in processed food divided by total activity in raw food, is reported for various types of corps. For white rice at a typical polishing yield of 90-92% from brown rice, the F{sub r} value range was 0.42-0.47. For leafy vegetable (indirect contamination), the average F{sub r} values were 0.92 (range: 0.27-1.2) after washing and 0.55 (range: 0.22-0.93) after washing and boiling. The data for some fruits are also reported. (authors)

  17. Studies of the use of heat from high temperature nuclear sources for hydrogen production processes

    Science.gov (United States)

    Farbman, G. H.

    1976-01-01

    Future uses of hydrogen and hydrogen production processes that can meet the demand for hydrogen in the coming decades were considered. To do this, a projection was made of the market for hydrogen through the year 2000. Four hydrogen production processes were selected, from among water electrolysis, fossil based and thermochemical water decomposition systems, and evaluated, using a consistent set of ground rules, in terms of relative performance, economics, resource requirements, and technology status.

  18. Large-scale continuous process to vitrify nuclear defense waste: operating experience with nonradioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Cosper, M B; Randall, C T; Traverso, G M

    1982-01-01

    The developmental program underway at SRL has demonstrated the vitrification process proposed for the sludge processing facility of the DWPF on a large scale. DWPF design criteria for production rate, equipment lifetime, and operability have all been met. The expected authorization and construction of the DWPF will result in the safe and permanent immobilization of a major quantity of existing high level waste. 11 figures, 4 tables.

  19. Investigation of Discharge Performance of SOFC Using Biogas and Its Off-gas%生物气及其电池尾气 SOFC 放电性能研究

    Institute of Scientific and Technical Information of China (English)

    王德震; 左薇; 张军; 吴晓燕; 孔晓伟

    2015-01-01

    To investigate the discharge performance of Ni-YSZ anode solid oxide fuel cell using various biogas and evaluate the valve of reusing its off-gas, electrochemical performance and gas properties were studied when two SOFC operated in tandem using diverse ratio of CH4/CO2 at 750 ℃. Compared with the cell performance using H2, two SOFC both operated at high power density with the first stage SOFC using various CH4/CO2 biogas and the second stage using the off-gas of the first stage. Both of two SOFC worked mostly steadily at constant current density for short time with little carbon deposition. The analysis of the gas properties at 566 mA•cm-2 indicated that the dry reforming rate was the highest when the ratio of CH4/CO2 was 2. The research results show that power generation of SOFC using biogas and its off-gas is feasible. This research can be useful for designing gas circuit of SOFC piles using biogas.%为探究以不同浓度生物气为燃料的固体氧化物燃料电池(SOFC)发电性能及该类电池尾气的再发电价值,通过模拟含不同比例甲烷和二氧化碳的生物气,在750℃下对气路串联 Ni/YSZ 阳极支撑 SOFC 进行放电性能测试和气体特性分析。放电结果显示燃料气经第一级 SOFC 利用后通入第二级 SOFC,同氢气经过两级 SOFC 相比,不同浓度下生物气均获得了较高的功率密度,且短时间恒流时,两级电池均能稳定运行;两级电池均以566 mA· cm-2电流密度恒流放电时的气体分析表明,当 CO2/CH4为2时,电池内甲烷的干重整率最高。研究结果表明两级 SOFC 使用生物气及其电池尾气发电是可行的,可为以生物气为燃料 SOFC 电堆气路设计提供依据。

  20. Nuclear analytical chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Brune, D.; Forkman, B.; Persson, B.

    1984-01-01

    This book covers the general theories and techniques of nuclear chemical analysis, directed at applications in analytical chemistry, nuclear medicine, radiophysics, agriculture, environmental sciences, geological exploration, industrial process control, etc. The main principles of nuclear physics and nuclear detection on which the analysis is based are briefly outlined. An attempt is made to emphasise the fundamentals of activation analysis, detection and activation methods, as well as their applications. The book provides guidance in analytical chemistry, agriculture, environmental and biomedical sciences, etc. The contents include: the nuclear periodic system; nuclear decay; nuclear reactions; nuclear radiation sources; interaction of radiation with matter; principles of radiation detectors; nuclear electronics; statistical methods and spectral analysis; methods of radiation detection; neutron activation analysis; charged particle activation analysis; photon activation analysis; sample preparation and chemical separation; nuclear chemical analysis in biological and medical research; the use of nuclear chemical analysis in the field of criminology; nuclear chemical analysis in environmental sciences, geology and mineral exploration; and radiation protection.

  1. Absence of superoxide dismutase activity causes nuclear DNA fragmentation during the aging process

    Energy Technology Data Exchange (ETDEWEB)

    Muid, Khandaker Ashfaqul; Karakaya, Hüseyin Çaglar; Koc, Ahmet, E-mail: ahmetkoc@iyte.edu.tr

    2014-02-07

    Highlights: • Aging process increases ROS accumulation. • Aging process increases DNA damage levels. • Absence of SOD activity does not cause DNA damage in young cells. • Absence of SOD activity accelerate aging and increase oxidative DNA damages during the aging process. - Abstract: Superoxide dismutases (SOD) serve as an important antioxidant defense mechanism in aerobic organisms, and deletion of these genes shortens the replicative life span in the budding yeast Saccharomyces cerevisiae. Even though involvement of superoxide dismutase enzymes in ROS scavenging and the aging process has been studied extensively in different organisms, analyses of DNA damages has not been performed for replicatively old superoxide dismutase deficient cells. In this study, we investigated the roles of SOD1, SOD2 and CCS1 genes in preserving genomic integrity in replicatively old yeast cells using the single cell comet assay. We observed that extend of DNA damage was not significantly different among the young cells of wild type, sod1Δ and sod2Δ strains. However, ccs1Δ mutants showed a 60% higher amount of DNA damage in the young stage compared to that of the wild type cells. The aging process increased the DNA damage rates 3-fold in the wild type and more than 5-fold in sod1Δ, sod2Δ, and ccs1Δ mutant cells. Furthermore, ROS levels of these strains showed a similar pattern to their DNA damage contents. Thus, our results confirm that cells accumulate DNA damages during the aging process and reveal that superoxide dismutase enzymes play a substantial role in preserving the genomic integrity in this process.

  2. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate

  3. On the quantum information processing in nuclear magnetic resonance quantum computing experiments

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, E.R. de; Bonk, F.A.; Vidoto, E.L.G.; Bonagamba, T.J. [Universidade de Sao Paulo (IFSC/USP), Sao Carlos, SP (Brazil). Inst. de Fisica; Sarthour, R.S.; Guimaraes, A.P.; Oliveira, I.S. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Freitas, J.C.C. [Universidade Federal do Espirito Santo (UFES), Vitoria, ES (Brazil). Dept. de Fisica

    2003-07-01

    Full text: Nuclear Magnetic Resonance appeared in the late nineties to be the most promising candidate to run quantum computing algorithms. An impressive number of experiments demonstrating the implementation of all logic gates and quantum algorithms in systems with a small number of qubits stimulated the general excitement about the technique, and greatly promoted the field. Particularly important were those experiments where entanglement of particles were aimed at. Entanglement is the most fundamental (and weird !) aspect of quantum systems, and is at the basis of quantum teleportation and quantum cryptography, yet impossible to prove in NMR experiments. The hardcore of NMR quantum computing are the so-called pseudo-pure states, upon which radiofrequency (RF) pulses act to implement quantum mechanical unitary transformations, promoting changes in both, Zeeman level populations and coherences in the density matrix. Whereas pseudo-pure states are special non-equilibrium diagonal states, coherences encode information about superposition states. Now, one could safely say that the whole business of quantum computing goes about controlling relative ket phases. In spite of the impossibility to univocally associating a given quantum state to a NMR spectrum, it is possible to demonstrate the phase action of RF pulses over relative ket phases, even if no population changes take place. In this talk these issues will be addressed, and we will show experimental results of our own where this is done in the two-qubit quadrupole nuclei {sup 23}Na in C{sub 10}H{sub 21}NaO{sub 4}S liquid crystal. We demonstrate the reversibility of the Hadamard gate, and of a quantum circuit which generates pseudo-Bell states. The success of the operation reaches almost 100% in the case of the state |01+|10, 80% in the cases of |00> + |01> and |10> + |11>, and 65% for the cat-state |00> + |11>. (author)

  4. Decontamination of nuclear graphite by thermal processing; Dekontamination von Nukleargraphit durch thermische Behandlung

    Energy Technology Data Exchange (ETDEWEB)

    Florjan, Monika W.

    2010-04-15

    The main problem in view of the direct disposal of the nuclear graphite is its large volume. This waste contains long-lived and short-lived radionuclides which determine the waste strategy. The irradiated graphite possess high amount of the {sup 14}C isotope. The main object of the present work was the selective separation of {sup 14}C isotope from the isotope {sup 12}C by thermal treatment (pyrolysis, partial oxidation). A successful separation could reduce the radiotoxicity and offer a different disposal strategy. Three different graphite types were investigated. The samples originate from the reflector and from the flaking of spherical fuel elements of the high-temperature reactor (AVR) Juelich. The samples from the thermal column of the research reactor (Merlin, Juelich) were also investigated. The maximum tritium releases were obtained both in inert gas atmosphere (N{sub 2}) and under water vapour-oxidizing conditions at 1280 C and 900 C. Furthermore it could be shown that 28% of {sup 14}C could be released under inert gas conditions at a 1280 C. By additive of oxidizing agent such as water vapour and oxygen the {sup 14}C release could be increased. Under water vapour-oxidizing conditions at a temperature of 1280 C up to 93% of the {sup 14}C was separated from the graphite. The matrix corrosion of 5.4% was obtained. The selective separation of the {sup 14}C is possible, because a substantial part of the radiocarbon is bound near the grain boundary surfaces. (orig.)

  5. Sample registration software for process automation in the Neutron Activation Analysis (NAA) Facility in Malaysia nuclear agency

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Nur Aira Abd, E-mail: nur-aira@nuclearmalaysia.gov.my; Yussup, Nolida; Ibrahim, Maslina Bt. Mohd; Mokhtar, Mukhlis B.; Soh Shaari, Syirrazie Bin Che; Azman, Azraf B. [Technical Support Division, Malaysian Nuclear Agency, 43000, Kajang, Selangor (Malaysia); Salim, Nazaratul Ashifa Bt. Abdullah [Division of Waste and Environmental Technology, Malaysian Nuclear Agency, 43000, Kajang, Selangor (Malaysia); Ismail, Nadiah Binti [Fakulti Kejuruteraan Elektrik, UiTM Pulau Pinang, 13500 Permatang Pauh, Pulau Pinang (Malaysia)

    2015-04-29

    Neutron Activation Analysis (NAA) had been established in Nuclear Malaysia since 1980s. Most of the procedures established were done manually including sample registration. The samples were recorded manually in a logbook and given ID number. Then all samples, standards, SRM and blank were recorded on the irradiation vial and several forms prior to irradiation. These manual procedures carried out by the NAA laboratory personnel were time consuming and not efficient. Sample registration software is developed as part of IAEA/CRP project on ‘Development of Process Automation in the Neutron Activation Analysis (NAA) Facility in Malaysia Nuclear Agency (RC17399)’. The objective of the project is to create a pc-based data entry software during sample preparation stage. This is an effective method to replace redundant manual data entries that needs to be completed by laboratory personnel. The software developed will automatically generate sample code for each sample in one batch, create printable registration forms for administration purpose, and store selected parameters that will be passed to sample analysis program. The software is developed by using National Instruments Labview 8.6.

  6. Development of the advanced nuclear materials -A study on the polymer catalyst process technology-

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Il Hyun; Jung, Heung Suk; Lee, Han Soo; An, Doh Heui; Kang, Heui Suk; Baek, Seung Woo; Lee, Sung Hoh; Sung, Kee Woong; Kim, Kwang Lak; Kim, Jong Hoh; Koo, Je Hyoo; Park, Tae Keun; Kim, Sang Hwan; Yoo, Ryong; Song, Myung Jae; Son, Soon Hwan; Choi, Jung Kil; Lee, Jae Choon; Jung, Moon Kyoo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Heavy water is used as moderator and coolant in pressurized heavy water power plants. According to the governmental long-term plan for power supply, Korea is scheduled to construct new four pressurized heavy water power plants till the year 2006. Total heavy water make-up for these plants would be 22 Mg/a from the year 2006. Reformed hydrogen processes is considered best suited to Korea. Hydrophobic catalysts for this process were manufactured and the performance of hydrogen isotope exchange was investigated. The overall mass transfer coefficients varied between 0.004 and 2.295 m3 HD/m3 Bed.sec. and heavy water separation processes using the catalysts were optimized. 66 figs, 62 tabs, 19 refs. (Author).

  7. The importance of nuclear masses in the astrophysical rp-process

    CERN Document Server

    Schatz, H

    2006-01-01

    The importance of mass measurements for astrophysical capture processes in general, and for the rp-process in X-ray bursts in particular is discussed. A review of the current uncertainties in the effective lifetimes of the major waiting points 64Ge, 68Se, and 72Kr demonstrates that despite of recent measurements uncertainties are still significant. It is found that mass measurements with an accuracy of the order of 10 keV or better are desirable, and that reaction rate uncertainties play a critical role as well.

  8. Nuclear-waste-management. Quarterly progress report, July-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1981-12-01

    Progress reports and summaries are presented for the following: high-level waste process development, alternate waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and fuel pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

  9. Nuclear-waste-management. Quarterly progress report, July-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1981-12-01

    Progress reports and summaries are presented for the following: high-level waste process development, alternate waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and fuel pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

  10. Signal processing system design for improved shutdown system of CANDU{sup ®} nuclear reactors in large break LOCA events

    Energy Technology Data Exchange (ETDEWEB)

    Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Xia, Lingzhi; Isham, Manir U. [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Ponomarev, Vladimir [Megawatt Solutions, 1235 Radom St., unit 68, Pickering, ON, Canada L1W 1J3 (Canada)

    2016-03-15

    Highlights: • Neutronic signal processing system design to improve CANDU SDS1 performance. • Reactor modeling for CANDU LLOCA transient. • MATLAB/Simulink system implementation for the SDS1 trip logic. • Increasing the SDS1 trip response. - Abstract: For CANDU reactors, several options to improve CANDU nuclear power plant operation safety margin have been investigated in this paper. A particular attention is paid to the response time of CANDU shutdown system number 1 (SDS1) in case of large break loss of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental analysis is performed to CANDU LLOCA event, and the power transient signal is generated. In order to improve the SDS1 response time during LLOCA events, an innovative power measurement and signal processing system is particularly designed. The new signal processing system is implemented with the input of the LLOCA power transient, and the simulation results of the reactor trip time and signal are compared to those of the existing system in CANDU power plants. It is demonstrated that the new signal processing system can not only achieve a shorter reactor trip time than the existing system, but also accommodate the spurious trip immunity. This will significantly enhance the safety margin for the power plant operation, or bring extra economical benefits to the power plant units.

  11. Digital pulse processing and optimization of the front-end electronics for nuclear instrumentation.

    Science.gov (United States)

    Bobin, C; Bouchard, J; Thiam, C; Ménesguen, Y

    2014-05-01

    This article describes an algorithm developed for the digital processing of signals provided by a high-efficiency well-type NaI(Tl) detector used to apply the 4πγ technique. In order to achieve a low-energy threshold, a new front-end electronics has been specifically designed to optimize the coupling to an analog-to-digital converter (14 bit, 125 MHz) connected to a digital development kit produced by Altera(®). The digital pulse processing is based on an IIR (Infinite Impulse Response) approximation of the Gaussian filter (and its derivatives) that can be applied to the real-time processing of digitized signals. Based on measurements obtained with the photon emissions generated by an (241)Am source, the energy threshold is estimated to be equal to ~2 keV corresponding to the physical threshold of the NaI(Tl) detector. An algorithm developed for a Silicon Drift Detector used for low-energy x-ray spectrometry is also described. In that case, the digital pulse processing is specifically designed for signals provided by a reset-type preamplifier ((55)Fe source).

  12. Sycamore amyloplasts can import and process precursors of nuclear encoded chloroplast proteins.

    Science.gov (United States)

    Strzalka, K; Ngernprasirtsiri, J; Watanabe, A; Akazawa, T

    1987-12-16

    Amyloplasts isolated from white-wild suspension-cultured cells of sycamore (Acer pseudoplatanus L.) are found to import and process the precursor of the small subunit (pS) of ribulose-1,5-bisphosphate carboxylase/oxygenase of spinach, but they lack the ability to form its holoenzyme due to the absence of both the large subunit and its binding-protein. They also import the precursor of the 33-kDa extrinsic protein (p33-kDa) of the O2-evolving complex of Photosystem II from spinach, but process is only to an intermediate form (i33-kDa). Chloroplasts from green-mutant cells of sycamore process p33-kDa to its mature form in this heterologous system. These results suggest that the thylakoid-associated protease responsible for the second processing step of p33-kDa is missing in amyloplasts, possibly due to the absence of thylakoid-membranes. In contrast, the apparent import of the precursor of the light-harvesting chlorophyll a/b-binding apoprotein (pLHCP) from spinach was not detected. Sycamore amyloplasts may lack the ability to import this particular thylakoid-protein, or rapidly degrade the imported molecules in the absence of thylakoid-membranes for their proper insertion.

  13. Characterization of quantum algorithms by quantum process tomography using quadrupolar spins in solid-state nuclear magnetic resonance.

    Science.gov (United States)

    Kampermann, H; Veeman, W S

    2005-06-01

    NMR quantum computing with qubit systems represented by nuclear spins (I=12) in small molecules in liquids has led to the most successful experimental quantum information processors so far. We use the quadrupolar spin-32 sodium nuclei of a NaNO3 single crystal as a virtual two-qubit system. The large quadrupolar coupling in comparison with the environmental interactions and the usage of strongly modulating pulses allow us to manipulate the system fast enough and at the same time keeping the decoherence reasonably slow. The experimental challenge is to characterize the "calculation" behavior of the quantum processor by process tomography which is here adapted to the quadrupolar spin system. The results of a selection of quantum gates and algorithms are presented as well as a detailed analysis of experimental results.

  14. Homotypic cell cannibalism, a cell-death process regulated by the nuclear protein 1, opposes to metastasis in pancreatic cancer

    Science.gov (United States)

    Cano, Carla E; Sandí, María José; Hamidi, Tewfik; Calvo, Ezequiel L; Turrini, Olivier; Bartholin, Laurent; Loncle, Céline; Secq, Véronique; Garcia, Stéphane; Lomberk, Gwen; Kroemer, Guido; Urrutia, Raul; Iovanna, Juan L

    2012-01-01

    Pancreatic adenocarcinoma (PDAC) is an extremely deadly disease for which all treatments available have failed to improve life expectancy significantly. This may be explained by the high metastatic potential of PDAC cells, which results from their dedifferentiation towards a mesenchymal phenotype. Some PDAC present cell-in-cell structures whose origin and significance are currently unknown. We show here that cell-in-cells form after homotypic cell cannibalism (HoCC). We found PDAC patients whose tumours display HoCC develop less metastasis than those without. In vitro, HoCC was promoted by inactivation of the nuclear protein 1 (Nupr1), and was enhanced by treatment with transforming growth factor β. HoCC ends with death of PDAC cells, consistent with a metastasis suppressor role for this phenomenon. Hence, our data indicates a protective role for HoCC in PDAC and identifies Nupr1 as a molecular regulator of this process. PMID:22821859

  15. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

    2016-08-31

    Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  16. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

    2016-08-31

    This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  17. Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

    Energy Technology Data Exchange (ETDEWEB)

    Eleon, Cyrille; Passard, Christian; Hupont, Nicolas; Estre, Nicolas [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Battel, Benjamin; Doumerc, Philippe; Dupuy, Thierry; Batifol, Marc [AREVA NC, La Hague plant - Nuclear Measurement Team, F-50444 Beaumont-Hague (France); Grassi, Gabriele [AREVA NC, 1 place Jean-Millier, 92084 Paris-La-Defense cedex (France)

    2015-07-01

    Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no. 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)

  18. Spent nuclear fuel project cold vacuum drying facility process water conditioning system design description

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1998-11-30

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Process Water Conditioning (PWC) System. The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), the HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the PWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  19. EFFICACY OF FILTRATION PROCESSES TO OBTAIN WATER CLARITY AT K EAST SPENT NUCLEAR FUEL (SNF) BASIN

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN JB

    2006-09-28

    The objective is to provide water clarity to the K East Basin via filtration processes. Several activities are planned that will challenge not only the capacity of the existing ion exchange modules to perform as needed but also the current filtration system to maintain water clarity. Among the planned activities are containerization of sludge, removal of debris, and hydrolasing the basin walls to remove contamination.

  20. Innovative precipitation in emulsion process: toward a non-nuclear industrial application

    Energy Technology Data Exchange (ETDEWEB)

    Ollivier, M.; Borda, G.; Charton, S. [CEA, Centre de Marcoule, DEN,DTEC,SGCS, F-30207 Bagnols-sur-Ceze (France); Flouret, J. [OCM, ZI Quai Jean Jaures, 197 Avenue Marie Curie, 07800 La Voulte-sur-Rhone (France)

    2016-07-01

    A precipitation in emulsion process has been proposed by Borda et al. in 2008 for the continuous precipitation of lanthanides or actinides as oxalate, in order to either increase the production capacity or allow the precipitation of long-life radioactive elements under optimum safety conditions. During research/development tests, a strong correlation between the emulsion's properties and those of the particles produced have been evidenced, thus enabling the size and morphology of the powder to be tuned by varying the droplets properties, the latter being controlled by the column operating conditions. This process thus appears as an attractive alternative to conventional processes for the synthesis of high-value precipitates; as it offers interesting intensification capabilities. In this context, the feasibility of the precipitation of bismuth subnitrate (BSN), for which the emulsion route for precipitation seems to be particularly attractive, has been studied. Indeed, the division of the reacting volume into droplets may allow efficient temperature regulation of the exothermic reaction. In addition, an improvement of the product appearance is expected. This first phase of the feasibility study focused on the choice of the organic phase and the sensitivity of the droplets and solid particles properties to the operating conditions. Following the encouraging results observed in stirred-tank reactor, we successfully tested the implementation in a pulsed column, at lab-scale. (authors)

  1. Nuclear binding around the RP-process waiting points $^{68}$Se and $^{72}$Kr

    CERN Multimedia

    2002-01-01

    Encouraged by the success of mass determinations of nuclei close to the Z=N line performed at ISOLTRAP during the year 2000 and of the recent decay spectroscopy studies on neutron-deficient Kr isotopes (IS351 collaboration), we aim to measure masses and proton separation energies of the bottleneck nuclei defining the flow of the astrophysical rp-process beyond A$\\sim$70. In detail, the program includes mass measurements of the rp-process waiting point nuclei $^{68}$Se and $^{72}$Kr and determination of proton separation energies of the proton-unbound $^{69}$Br and $^{73}$Rb via $\\beta$-decays of $^{69}$Kr and $^{73}$Sr, respectively. The aim of the project is to complete the experimental database for astrophysical network calculations and for the liquid-drop type of mass models typically used in the modelling of the astrophysical rp process in the region. The first beamtime is scheduled for the August 2001 and the aim is to measure the absolute mass of the waiting-point nucleus $^{72}$Kr.

  2. Pebble heater suppresses synthesis of dioxins and furans in off-gas generated by incineration of halogen-rich fuel from WEEE

    Energy Technology Data Exchange (ETDEWEB)

    Schlummer, M.; Gruber, L.; Maeurer, A.; Wolz, G. [Fraunhofer Institute for Process Engineering and Packaging IVV, Freising (Germany); Fischer, W.; Quicker, P. [ATZ-EVUS, Development Center for Process Engineering, Sulzbach-Rosenberg (Germany)

    2004-09-15

    Changes in German and European legislation have led to altered approaches for the disposal of polymer-rich shredding residues (SR). Whereas disposal in landfills was the strategy of choice in the last decades, thermal treatment is supported now. However, when waste electric and electronic equipment (WEEE) is the source of SR, thermal treatment is complicated by a bromine and chlorine load in the lower percent range the presence of polybrominated dioxins and furans (PBDD/F) in the ppb range and by brominated flame retardants including polybrominated biphenyl ethers, which serve as dioxin precursors. Here we present data of a pilot application of the pebble heater technology for the treatment of raw gas derived from the incineration of polymeric materials from WEEE. Since the pilot experiments were performed on an existing pebble heater test plant in the small-technical scale, waste throughput and experimental design had to be adjusted to the given circumstances. As the study focussed on exhaust treatment and not on the incineration process itself, a liquid fuel was applied as a model for SR from WEEE. The incineration of a liquid fuel was preferred, since it could be implemented in the given test plant by spray injection, thus minimising technical modifications of the test plant and optimising the combustion efficiency compared to incineration of solid polymer granulates. Fuel and exhaust gases, which passed the pebble heater bed, were sampled and analysed for PCDD/F and PBDD/F. The pilot incineration was tested for the compliance with the PCDD/F emission limits given by European directive 2000/76/EC, and overall mass balances were calculated for PCDD/F and PBDD/F.

  3. Toward a Greenish Nuclear Fuel Cycle: Ionic Liquids as Solvents for Spent Nuclear Fuel Reprocessing and Other Decontamination Processes for Contaminated Metal Waste

    Science.gov (United States)

    Straka, Martin

    2016-12-01

    The final disposition of spent nuclear fuel (SNF) is an area that requires innovative solutions. The use of ionic liquids (ILs) has been examined as one means to remediate SNF in a variety of different chemical environments and with different chemical starting materials. The effectiveness of various ILs for SNF reprocessing, as well as the reaction chemistry that occurs in them, is discussed.

  4. Results From an International Simulation Study on Couples Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

    Energy Technology Data Exchange (ETDEWEB)

    J. Rutqvist; D. Barr; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q. Liu; Y. Oda; W. Wang; C. Zhang

    2006-08-02

    As part of the ongoing international DECOVALEX project, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near waste emplacement drifts of geological nuclear waste repositories. The simulations were conducted for two generic repository types, one with open and the other with back-filled repository drifts, under higher and lower postclosure temperatures, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses is currently being resolved. In particular, good agreement in the basic thermal-mechanical responses was achieved for both repository types, even though some teams used relatively simplified thermal-elastic heat-conduction models that neglected complex near-field thermal-hydrological processes. The good agreement between the complex and simplified process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level.

  5. Hydroxylamine as a potential reagent for dissolution off gas scrubbing in spent fuel reprocessing: kinetics of the iodine reduction. An example of similarity between the studies on the chemistry of iodine in reactor safety and in spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Cau Dit Coumes, C.; Devisme, F. [Commissariat a l`Energie Atomique, CE/VRH, Bagnols-sur-Ceze (France); Vargas, S.; Chopin-Dumas, J. [Laboratoire d`Electrochimie Inorganique, ENSSPICAM, Marseille (France)

    1996-12-01

    Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30{sup o}C) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs.

  6. Design features of a sulphuric acid plant based on lead and zinc sintering machine off-gas%硫酸用新型耐蚀合金的研究与开发

    Institute of Scientific and Technical Information of China (English)

    刘焕安

    2001-01-01

    Design features of a 150kt/a sulphuric acid plant based on lead and zinc sintering machine off-gases are described. The plant adopted a single absorption technology including closed dilute-acid-scrubbing gas cleaning and ammonia-acid off-gas treatment. The dilute acid settling system,cooling water circulation system, installation of electrostatic precipitator, high-temperature absorption technology and acid distributor of drying and absorption section, and preheater, hot bypass and insulation of conversion section are emphasized in detail.%论述硫酸对金属腐蚀的特殊性和合金设计的基本原理。介绍高温浓硫酸用高硅不锈钢HD-1、合金球墨铸铁HD-3以及稀硫酸用高钼含氮奥氏体不锈钢HD-7、HD-11的研究开发和应用范围。

  7. Problems in experimental and mathematical investigations of the accidental thermalhydraulic processes in RBMK nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nigmatulin, B.I.; Tikhonenko, L.K. [Engineering Centre (EREC) for Nuclear Plants Safety, Electrogorsk (Russian Federation); Blinkov, V.N. [Aviation Institute, Kharkov (Ukraine)] [and others

    1995-09-01

    In this paper the thermalhydraulic scheme and peculiarities of the boiling water graphite-moderated channel-type reactor RBMK are presented and discussed shortly. The essential for RBMK transient regimes, accidental situations and accompanying thermalhydraulic phenomena and processes are formulated. These data are presented in the form of cross reference matrix (version 1) for system computer codes verification. The paper includes qualitative analysis of the computer codes and integral facilities which have been used or can be used for RBMK transients and accidents investigations. The stability margins for RBMK-1000 and RBMK-1500 are shown.

  8. Singular value decomposition for photon-processing nuclear imaging systems and applications for reconstruction and computing null functions.

    Science.gov (United States)

    Jha, Abhinav K; Barrett, Harrison H; Frey, Eric C; Clarkson, Eric; Caucci, Luca; Kupinski, Matthew A

    2015-09-21

    Recent advances in technology are enabling a new class of nuclear imaging systems consisting of detectors that use real-time maximum-likelihood (ML) methods to estimate the interaction position, deposited energy, and other attributes of each photon-interaction event and store these attributes in a list format. This class of systems, which we refer to as photon-processing (PP) nuclear imaging systems, can be described by a fundamentally different mathematical imaging operator that allows processing of the continuous-valued photon attributes on a per-photon basis. Unlike conventional photon-counting (PC) systems that bin the data into images, PP systems do not have any binning-related information loss. Mathematically, while PC systems have an infinite-dimensional null space due to dimensionality considerations, PP systems do not necessarily suffer from this issue. Therefore, PP systems have the potential to provide improved performance in comparison to PC systems. To study these advantages, we propose a framework to perform the singular-value decomposition (SVD) of the PP imaging operator. We use this framework to perform the SVD of operators that describe a general two-dimensional (2D) planar linear shift-invariant (LSIV) PP system and a hypothetical continuously rotating 2D single-photon emission computed tomography (SPECT) PP system. We then discuss two applications of the SVD framework. The first application is to decompose the object being imaged by the PP imaging system into measurement and null components. We compare these components to the measurement and null components obtained with PC systems. In the process, we also present a procedure to compute the null functions for a PC system. The second application is designing analytical reconstruction algorithms for PP systems. The proposed analytical approach exploits the fact that PP systems acquire data in a continuous domain to estimate a continuous object function. The approach is parallelizable and

  9. Development of the GANEX process for the reprocessing of Gen IV spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Miguirditchian, M.; Chareyre, L.; Sorel, C.; Bisel, I.; Baron, P.; Masson, M. [CEA Marcoule - DEN/DRCP/SCPS - BP 17171, Bagnols-sur-Ceze, 30207 (France)

    2008-07-01

    The GANEX (group actinide extraction) process is composed of two extraction cycles following the dissolution of the spent fuel. In the first cycle, uranium(VI) is selectively extracted from the dissolution solution using a mono amide extractant DEHiBA (NN--di-(ethyl-2-hexyl)iso-butyr-amide) diluted in HTP (hydrogen tetra-propylene). Experimental data and modelling of uranium(VI) and nitric acid extractions are presented. A flowsheet was designed and was successfully tested in laboratory scale mixer-settlers on a surrogate uranium(VI)/HNO{sub 3} feed. For the group actinide separation in the second cycle, the DIAMEX-SANEX process was adjusted to separate neptunium and plutonium along with americium and curium. The data showed the possibility to extract all actinides together with good selectivities versus lanthanides. The flowsheets of the two GANEX cycles which will be tested on a high active feed at the end of 2008 in Atalante facility are presented. (authors)

  10. Evaluation of helium impurity impacts on Spent Nuclear Fuel project processes (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    SHERRELL, D.L.

    1999-09-21

    This document identifies the types and quantities of impurities that may be present within helium that is introduced into multi-canister overpacks (MCO)s by various SNF Project facilities, including, but not limited to the Cold Vacuum Drying (CVD) Facility (CVDF). It then evaluates possible impacts of worst case impurity inventories on MCO drying, transportation, and storage processes. Based on the evaluation results, this document: (1) concludes that the SNF Project helium procurement specification can be a factor-of-ten less restrictive than a typical vendor's standard offering (99.96% pure versus the vendor's 99.997% pure standard offering); (2) concludes that the CVDF's current 99.5% purity requirement is adequate to control the quality of the helium that is delivered to the MCO by the plant's helium distribution system; and (3) recommends specific impurity limits for both of the above cases.

  11. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon Consult AB, Nykoeping (Sweden)

    1995-04-01

    The report reviews the properties of spent fuel which are considered to be of most importance in determining the corrosion behaviour in groundwaters. Pellet cracking and fragment size distribution are discussed, together with the available results of specific surface area measurements on spent fuel. With respect to the importance of fuel microstructure, emphasis is placed on recent work on the so called structural rim effect, which consists of the formation of a zone of high porosity, and the polygonization of fuel grains to form many sub-grains, at the pellet rim, and appears to be initiated when the average pellet burnup exceeds a threshold of about 40 MWd/kgU. Due to neutron spectrum effects, the pellet rim is also associated with the buildup of plutonium and other actinides, which results in an enhanced local burnup and specific activity of both beta-gamma and alpha radiation, thus representing a greater potential for radiolysis effects in ingressed groundwater. The report presents and discusses the results of quantitative determination of the radial profiles of burnup and alpha activity on spent fuel with average burnups from 21.2 to 49 MWd/kgU. In addition to the radial variation of fission product and actinide inventories due to the effects mentioned above, migration, redistribution and release of some fission products can occur during reactor irradiation and the report concludes with a short review of these processes.

  12. Quasicrystalline Approach to Prediting the Spinel-Nepheline Liquidus: Application to Nuclear Waste Glass Processing

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, Carol

    2005-10-10

    The crystal-melt equilibria in complex fifteen component melts are modeled based on quasicrystalline concepts. A pseudobinary phase diagram between acmite (which melts incongruently to a transition metal ferrite spinel) and nepheline is defined. The pseudobinary lies within the Al{sub 2}O{sub 3}-Fe{sub 2}O{sub 3}-Na{sub 2}O-SiO{sub 2} quaternary system that defines the crystallization of basalt glass melts. The pseudobinary provides the partitioning of species between the melt and the primary liquidus phases. The medium range order of the melt and the melt-crystal exchange equilibria are defined based on a constrained mathematical treatment that considers the crystallochemical coordination of the elemental species in acmite and nepheline. The liquidus phases that form are shown to be governed by the melt polymerization and the octahedral site preference energies. This quasicrystalline liquidus model has been used to prevent unwanted crystallization in the world's largest high level waste (HLW) melter for the past three years while allowing >10 wt% higher waste loadings to be processed.

  13. Structural Characterization of Natural and Processed Zircons with X-Rays and Nuclear Techniques

    Directory of Open Access Journals (Sweden)

    Laura C. Damonte

    2017-01-01

    Full Text Available In ceramic industry, zircon sand is widely used in different applications because zirconia plays a role as common opacifying constituent. In particular, it is used as a basic component of glazes applied to ceramic tiles and sanitary ware as well as an opacifier in unglazed bulk porcelain stoneware. Natural zircon sands are the major source of zirconium minerals for industrial applications. In this paper, long, medium, and short range studies were conducted on zirconium minerals originated from Australia, South Africa, and United States of America using conventional and less conventional techniques (i.e., X-Ray Diffraction (XRD, Positron Annihilation Lifetime Spectroscopy (PALS, and Perturbed Angular Correlations (PAC in order to reveal the type and the extension of the regions that constitute the metamict state of zircon sands and the modifications therein produced as a consequence of the industrial milling process and the thermal treatment in the production line. Additionally, HPGe gamma-ray spectroscopy confirms the occurrence of significant levels of natural radioactivity responsible for metamictization in the investigated zircon samples. Results from XRD, PALS, and PAC analysis confirm that the metamict state of zircon is a dispersion of submicron disordered domains in a crystalline matrix of zircon.

  14. COMPACT STELLAR BINARY ASSEMBLY IN THE FIRST NUCLEAR STAR CLUSTERS AND r-PROCESS SYNTHESIS IN THE EARLY UNIVERSE

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez-Ruiz, Enrico; MacLeod, Morgan [Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States); Trenti, Michele [Kavli Institute for Cosmology and Institute of Astronomy, University of Cambridge, Madingley Road, Cambridge CB3 0HA (United Kingdom); Roberts, Luke F. [TAPIR, California Institute of Technology, Pasadena, California 91125 (United States); Lee, William H.; Saladino-Rosas, Martha I. [Instituto de Astronomía, Universidad Nacional Autónoma de México, México DF 04510, México (Mexico)

    2015-04-01

    Investigations of elemental abundances in the ancient and most metal deficient stars are extremely important because they serve as tests of variable nucleosynthesis pathways and can provide critical inferences of the type of stars that lived and died before them. The presence of r-process elements in a handful of carbon-enhanced metal-poor (CEMP-r) stars, which are assumed to be closely connected to the chemical yield from the first stars, is hard to reconcile with standard neutron star mergers. Here we show that the production rate of dynamically assembled compact binaries in high-z nuclear star clusters can attain a sufficient high value to be a potential viable source of heavy r-process material in CEMP-r stars. The predicted frequency of such events in the early Galaxy, much lower than the frequency of Type II supernovae but with significantly higher mass ejected per event, can naturally lead to a high level of scatter of Eu as observed in CEMP-r stars.

  15. NVL2, a nucleolar AAA-ATPase, is associated with the nuclear exosome and is involved in pre-rRNA processing

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikatsu, Yuki [Department of Life Systems, Institute of Technology and Science, The University of Tokushima Graduate School, Tokushima 770-8506 (Japan); Ishida, Yo-ichi; Sudo, Haruka [Department of Molecular and Cellular Biochemistry, Meiji Pharmaceutical University, Kiyose, Tokyo 204-8588 (Japan); Yuasa, Keizo; Tsuji, Akihiko [Department of Life Systems, Institute of Technology and Science, The University of Tokushima Graduate School, Tokushima 770-8506 (Japan); Nagahama, Masami, E-mail: nagahama@my-pharm.ac.jp [Department of Molecular and Cellular Biochemistry, Meiji Pharmaceutical University, Kiyose, Tokyo 204-8588 (Japan)

    2015-08-28

    Nuclear VCP-like 2 (NVL2) is a member of the chaperone-like AAA-ATPase family and is involved in the biosynthesis of 60S ribosomal subunits in mammalian cells. We previously showed the interaction of NVL2 with a DExD/H-box RNA helicase MTR4/DOB1, which is a known cofactor for an exoribonuclease complex, the exosome. This finding implicated NVL2 in RNA metabolic processes during ribosome biogenesis. In the present study, we found that a series of mutations within the ATPase domain of NVL2 causes a defect in pre-rRNA processing into mature 28S and 5.8S rRNAs. Co-immunoprecipitation analysis showed that NVL2 was associated with the nuclear exosome complex, which includes RRP6 as a nucleus-specific catalytic subunit. This interaction was prevented by depleting either MTR4 or RRP6, indicating their essential role in mediating this interaction with NVL2. Additionally, knockdown of MPP6, another cofactor for the nuclear exosome, also prevented the interaction by causing MTR4 to dissociate from the nuclear exosome. These results suggest that NVL2 is involved in pre-rRNA processing by associating with the nuclear exosome complex and that MPP6 is required for maintaining the integrity of this rRNA processing complex. - Highlights: • ATPase-deficient mutants of NVL2 have decreased pre-rRNA processing. • NVL2 associates with the nuclear exosome through interactions with MTR4 and RRP6. • MPP6 stabilizes MTR4-RRP6 interaction and allows NVL2 to interact with the complex.

  16. Children's (Pediatric) Nuclear Medicine

    Medline Plus

    Full Text Available ... its earliest stages as well as a patient’s immediate response to therapeutic interventions. Children's (pediatric) nuclear medicine ... leaving the nuclear medicine facility. Through the natural process of radioactive decay, the small amount of radiotracer ...

  17. Aqueous alteration in CR chondrites: Meteorite parent body processes as analogue for long-term corrosion processes relevant for nuclear waste disposal

    Science.gov (United States)

    Morlok, Andreas; Libourel, Guy

    2013-02-01

    Aqueous alteration of carbonaceous chondrites is one of the fundamental processes on accreting planetesimals that changes pristine materials from the formation of the Solar System. The study of mineralogical, petrological and chemical changes resulting from this alteration provides insight into the physical and chemical setting of forming planetesimals. CR chondrites provide samples for all stages of aqueous alteration, from type 3 to 1 (entirely hydrated), and are thus suited to study the alteration of pristine materials in a coherent sequence. Vitrification is a common way to store and stabilize fission products and minor actinides resulting from the reprocessing of nuclear spent fuel in a nuclear boro-silica glass in steel containers. The waste material has to be stored safely for a period of at least 105-106 years in a clay-rich geological repository. Laboratory experiments being too short to follow the long-term evolution of these materials, we analyzed the mineralogical, petrological and chemical changes in a series of CR chondrites (Renazzo CR2, Al Rais CR2, and GRO 95577 CR1) to serve as analogues. Rims of secondary materials around metal grains in contact to the fine-grained matrix serve as analogue to the interface between steel containment and the surrounding clay-rich geological layer, while chondrule glassy mesostasis is used as a proxy of the nuclear glass. With increasing degree of aqueous alteration in the sequence, Renazzo → Al Rais → GRO 95577, the size of the rims increase. Fe-rich alteration rims are ˜10 μm in thickness around metal grains in the fine-grained matrix in Renazzo. In Al Rais, multi-layered structures of interchanging Fe, S and P/Ca-rich layers appear, with a thickness of up to ˜30 μm. In the highly altered GRO 95577, extensive inner and external rims of secondary phases reach up to ˜200 μm into the surrounding matrix. In chondrules, metal in contact with the altered mesostasis shows similar trends, but with thinner

  18. Releases of radioactive substances from Swedish nuclear power plants (RAKU)

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, T.; Bergstroem, C. [ALARA Engineering AB, Skultuna (Sweden)

    1997-04-01

    Releases of radioactivity to air and water from Swedish nuclear power plants have been studied and compared with those from foreign reactors. Averaged over the years from commissioning of the reactors to the last year data are available, the release of radioactive noble gas from the Swedish BWRs has been about the same as from comparable foreign reactors. The oldest Swedish BWRs, Oskarshamn 1 and 2 (O1 and O2) and Ringhals 1 (R1), have simple off-gas systems with only one delay volume. All BWRs in US, Germany, Japan and Switzerland are equipped with more sophisticated off-gas systems. It can be expected that O1, O2 and R1 therefore will have the highest release of noble gas activity at an international comparison if they do not modernize their off-gas system. BWRs in US, Germany and Japan are today equipped with recombiners and with one exception also charcoal columns. Japanese BWRs report zero releases to air. Releases of radioactivity to water after commissioning was about the same for most of the studied reactors. Some of the newest German plants have had low annual releases already at commissioning. Improvements of the treatment systems at old German, Swiss and US reactors have significantly lowered the releases. For most of the Swedish plants the annual releases to water have remained at the initial level. Forsmark 3 has succeeded in decreasing the release of radionuclides to water by a factor of almost one hundred compared to other Swedish reactors. Also O3 has managed to decrease the liquid effluents. Japanese plants have zero release of radioactivity excluding tritium to water. The release of tritium is about the same for all reactors of the same type in the world. 35 refs, 31 figs, 24 tabs.

  19. Nuclear Medicine

    Science.gov (United States)

    ... for Parents/Teachers Resource Links for Students Glossary Nuclear Medicine What is nuclear medicine? What are radioactive ... NIBIB-funded researchers advancing nuclear medicine? What is nuclear medicine? Nuclear medicine is a medical specialty that ...

  20. Proceedings of seventh symposium on sharing of computer programs and technology in nuclear medicine, computer assisted data processing

    Energy Technology Data Exchange (ETDEWEB)

    Howard, B.Y.; McClain, W.J.; Landay, M. (comps.)

    1977-01-01

    The Council on Computers (CC) of the Society of Nuclear Medicine (SNM) annually publishes the Proceedings of its Symposium on the Sharing of Computer Programs and Technology in Nuclear Medicine. This is the seventh such volume and has been organized by topic, with the exception of the invited papers and the discussion following them. An index arranged by author and by subject is included.

  1. Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

    Science.gov (United States)

    Reza, S. M. Mohsin

    Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ˜10% of the graphite from PSR the PVT is reduced from 541°C to 421°C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ˜350°C while keeping the outlet temperature at 950°C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes

  2. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  3. Evaluation of the Cell Proliferation Process of Ovarian Follicles in Hypothyroid Rats by Proliferation Cell Nuclear Antigen Immunohistochemical Technique

    Directory of Open Access Journals (Sweden)

    M. Moghaddam Dorafshani

    2012-10-01

    Full Text Available Introduction & Objective: The normal females reproductive function , needs hypothalamus-hypophysis-ovarian extensive hormonal messages. Primary hypothyroidism is characterized by reduced production and secretion of thyroid hormones. During follicular growth PCNA (Proliferating Cell Nuclear Antigen and cycklin D complex play an important role in regulating cell proliferation .This study aimed to determine the cell proliferation index and how this process changes induced by thyroid hormone decreased in rat ovarian follicles.Materials & Methods: In this experimental study, 20 Wistar female rats were divided into experimental and control groups. Experimental group was chemically thyroidectomized by administering propylthiouracil (PTU (500 mg per liter of drinking water. The control group received normal drinking water. After three weeks rats were killed and their ovaries dissected and fixed for the histological preparation. Cell proliferation was determined by PCNA and stereological methods were used for counting cells.Results: Cell proliferation index showed a significant decrease in the frequency of follicular growth from prenatal to graafian follicles in hypothyroidism groups(P0.05 . PCNA expression determined that Primary follicle growth begins earlier. Positive PCNA cells were not observed in primordial follicles of the groups.Conclusion: According to the results of our study, this hypothesis is raised that granulosa cells in growing follicles may be increased by follicle adjacent cells in ovarian stroma . Hormonal changes following the reduction of thyroid hormones may greatly affect the cell proliferation index and lead to faster follicle degeneration.(Sci J Hamadan Univ Med Sci 2012; 19 (3:5-15

  4. Thermal Lens Spectroscopy as a 'new' analytical tool for actinide determination in nuclear reprocessing processes

    Energy Technology Data Exchange (ETDEWEB)

    Canto, Fabrice; Couston, Laurent; Magnaldo, Alastair [CEA-Valrho DEN/DRCP/SCPS/LCAM BP17171 30207 Bagnols/Ceze cedex (France); Broquin, Jean-Emmanuel [IMEP/ENSERG 23 rue des Martyrs BP257 38016 Grenoble (France); Signoret, Philippe [UM2/IES UMR 5214. Place Eugene Bataillon 34095 Montpellier cedex5 (France)

    2008-07-01

    Thermal Lens Spectroscopy (TLS) consists of measuring the effects induced by the relaxation of molecules excited by photons. Twenty years ago, the Cea already worked on TLS. Technologic reasons impeded. But, needs in sensitive analytical methods coupled with very low sample volumes (for example, traces of Np in the COEX{sup TM} process) and also the reduction of the nuclear wastes encourage us to revisit this method thanks to the improvement of optoelectronic technologies. We can also imagine coupling TLS with micro-fluidic technologies, decreasing significantly the experiments cost. Generally two laser beams are used for TLS: one for the selective excitation by molecular absorption (inducing the thermal lens) and one for probing the thermal lens. They can be coupled with different geometries, collinear or perpendicular, depending on the application and on the laser mode. Also, many possibilities of measurement have been studied to detect the thermal lens signal: interferometry, direct intensities variations, deflection etc... In this paper, one geometrical configuration and two measurements have been theoretically evaluated. For a single photodiode detection (z-scan) the limit of detection is calculated to be near 5*10{sup -6} mol*L{sup -1} for Np(IV) in dodecane. (authors)

  5. A comparative simulation study of coupled THM processes and their effect on fractured rock permeability around nuclear waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Rutqvist, Jonny; Barr, Deborah; Birkholzer, Jens T.; Fujisaki, Kiyoshi; Kolditz, Olf; Liu, Quan-Shen; Fujita, tomoo; Wang, Wenqing; Zhang, Cheng-Yuan

    2008-10-23

    This paper presents an international, multiple-code, simulation study of coupled thermal, hydrological, and mechanical (THM) processes and their effect on permeability and fluid flow in fractured rock around heated underground nuclear waste emplacement drifts. Simulations were conducted considering two types of repository settings: (a) open emplacement drifts in relatively shallow unsaturated volcanic rock, and (b) backfilled emplacement drifts in deeper saturated crystalline rock. The results showed that for the two assumed repository settings, the dominant mechanism of changes in rock permeability was thermal-mechanically-induced closure (reduced aperture) of vertical fractures, caused by thermal stress resulting from repository-wide heating of the rock mass. The magnitude of thermal-mechanically-induced changes in permeability was more substantial in the case of an emplacement drift located in a relatively shallow, low-stress environment where the rock is more compliant, allowing more substantial fracture closure during thermal stressing. However, in both of the assumed repository settings in this study, the thermal-mechanically-induced changes in permeability caused relatively small changes in the flow field, with most changes occurring in the vicinity of the emplacement drifts.

  6. A new combined nuclear magnetic resonance and Raman spectroscopic probe applied to in situ investigations of catalysts and catalytic processes

    Energy Technology Data Exchange (ETDEWEB)

    Camp, Jules C. J.; Mantle, Michael D. [Department of Chemical Engineering and Biotechnology, University of Cambridge, Pembroke Street, Cambridge CB2 3RA (United Kingdom); York, Andrew P. E. [Johnson Matthey Technology Centre, Blounts Court, Sonning Common, Reading RG4 9NH (United Kingdom); McGregor, James, E-mail: james.mcgregor@sheffield.ac.uk [Department of Chemical and Biological Engineering, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2014-06-15

    Both Raman and nuclear magnetic resonance (NMR) spectroscopies are valuable analytical techniques capable of providing mechanistic information and thereby providing insights into chemical processes, including catalytic reactions. Since both techniques are chemically sensitive, they yield not only structural information but also quantitative analysis. In this work, for the first time, the combination of the two techniques in a single experimental apparatus is reported. This entailed the design of a new experimental probe capable of recording simultaneous measurements on the same sample and/or system of interest. The individual datasets acquired by each spectroscopic method are compared to their unmodified, stand-alone equivalents on a single sample as a means to benchmark this novel piece of equipment. The application towards monitoring reaction progress is demonstrated through the evolution of the homogeneous catalysed metathesis of 1‑hexene, with both experimental techniques able to detect reactant consumption and product evolution. This is extended by inclusion of magic angle spinning (MAS) NMR capabilities with a custom made MAS 7 mm rotor capable of spinning speeds up to 1600 Hz, quantified by analysis of the spinning sidebands of a sample of KBr. The value of this is demonstrated through an application involving heterogeneous catalysis, namely the metathesis of 2-pentene and ethene. This provides the added benefit of being able to monitor bo