WorldWideScience

Sample records for source terms released

  1. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

    This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

  2. On the sequence of core-melt accidents: Fission product release, source terms and Chernobyl release

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, H

    1986-01-01

    There is a sketch of our ideas on the course of a core melt-out accident in a PWR. There is then a survey of the most important results on fission product release, which were obtained by experiments on the SASCHA melt-out plant. The 3rd part considers questions which are important for determining source terms for the environment and the last part contains some considerations on radioactivity release from the Chernobyl reactor.

  3. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    International Nuclear Information System (INIS)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2016-01-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  4. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  5. The influence of source term release parameters on health effects

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Ha, Jae Joo

    1998-08-01

    In this study, the influence of source term release parameters on the health effects was examined. This is very useful in identifying the relative importance of release parameters and can be an important factor in developing a strategy for reducing offsite risks. The release parameters investigated in this study are release height, heat content, fuel burnup, release time, release duration, and warning time. The health effects affected by the change of release parameters are early fatalities, cancer fatalities, early injuries, cancer injuries, early fatality risk, population weighted early fatality risk, population weighted cancer fatality risk, effective whole body population dose, population exceeding an early acute red bone marrow dose of 1.5 Sv, and distance at which early fatalities are expected to occur. As release height increases, the values of early health effects such as early fatalities and injuries decrease. However, the release height dose not have significant influences on late health effects. The values of both early and late health effects decrease as heat content increases. The increase fuel burnup, i.e., the increase of core inventories increases the late health effects, however, has small influence on the early health effects. But, the number of early injuries increases as the fuel burnup increases. The effects of release time increase shows very similar influence on both the early and late health effects. As the release time increases to 2 hours, the values of health effects increase and then decrease rapidly. As release duration increases, the values of late health effects increase slightly, however, the values of early health effects decrease. As warning time increases to 2 hours, the values of late health effects decrease and then shows no variation. The number of early injuries decreases rapidly as the warning time increases to 2 hours. However, the number of early fatalities and the early fatality risk increase as the warning time increases

  6. The Multimedia Environmental Pollutant Assessment System (MEPAS)reg-sign: Source-term release formulations

    International Nuclear Information System (INIS)

    Streile, G.P.; Shields, K.D.; Stroh, J.L.; Bagaasen, L.M.; Whelan, G.; McDonald, J.P.; Droppo, J.G.; Buck, J.W.

    1996-11-01

    This report is one of a series of reports that document the mathematical models in the Multimedia Environmental Pollutant Assessment System (MEPAS). Developed by Pacific Northwest National Laboratory for the US Department of Energy, MEPAS is an integrated impact assessment software implementation of physics-based fate and transport models in air, soil, and water media. Outputs are estimates of exposures and health risk assessments for radioactive and hazardous pollutants. Each of the MEPAS formulation documents covers a major MEPAS component such as source-term, atmospheric, vadose zone/groundwater, surface water, and health exposure/health impact assessment. Other MEPAS documentation reports cover the sensitivity/uncertainty formulations and the database parameter constituent property estimation methods. The pollutant source-term release component is documented in this report. MEPAS simulates the release of contaminants from a source, transport through the air, groundwater, surface water, or overland pathways, and transfer through food chains and exposure pathways to the exposed individual or population. For human health impacts, risks are computed for carcinogens and hazard quotients for noncarcinogens. MEPAS is implemented on a desktop computer with a user-friendly interface that allows the user to define the problem, input the required data, and execute the appropriate models for both deterministic and probabilistic analyses

  7. Release modes and processes relevant to source-term calculations at Yucca Mountain

    International Nuclear Information System (INIS)

    Apted, M.J.

    1994-01-01

    The feasibility of permanent disposal of radioactive high-level waste (HLW) in repositories located in deep geologic formations is being studied world-wide. The most credible release pathway is interaction between groundwater and nuclear waste forms, followed by migration of radionuclide-bearing groundwater to the accessible environment. Under hydrologically unsaturated conditions, vapor transport of volatile radionuclides is also possible. The near-field encompasses the waste packages composed of engineered barriers (e.g. man-made materials, such as vitrified waste forms, corrosion-resistant containers), while the far-field includes the natural barriers (e.g. host rock, hydrologic setting). Taken together, these two subsystems define a series of multiple, redundant barriers that act to assure the safe isolation of nuclear waste. In the U.S., the Department of energy (DOE) is investigating the feasibility of safe, long-term disposal of high-level nuclear waste at the Yucca Mountain site in Nevada. The proposed repository horizon is located in non-welded tuffs within the unsaturated zone (i.e. above the water table) at Yucca Mountain. The purpose of this paper is to describe the source-term models for radionuclide release from waste packages at Yucca Mountain site. The first section describes the conceptual release modes that are relevant for this site and waste package design, based on a consideration of the performance of currently proposed engineered barriers under expected and unexpected conditions. No attempt is made to asses the reasonableness nor probability of occurrence for any specific release mode. The following section reviews the waste-form characteristics that are required to model and constrain the release of radionuclides from the waste package. The next section present mathematical models for the conceptual release modes, selected from those that have been implemented into a probabilistic total system assessment code developed for the Electric Power

  8. Revised accident source terms for light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  9. Determination of Source Term for an Annual Stack Release of Gas Reactor G.A. Siwabessy

    International Nuclear Information System (INIS)

    Sudiyati; Syahrir; Unggul Hartoyo; Nugraha Luhur

    2008-01-01

    Releases of radionuclide from the reactor are noble gases, halogenides and particulates. The measurements were carried out directly on the air monitoring system of the stack. The results of these measurements are compared with the annual Source-Term data from the Safety Analyses report (SAR) of RSG-GAS. The measurement results are smaller than the data reported in SAR document. (author)

  10. Chernobyl source term estimation

    International Nuclear Information System (INIS)

    Gudiksen, P.H.; Harvey, T.F.; Lange, R.

    1990-09-01

    The Chernobyl source term available for long-range transport was estimated by integration of radiological measurements with atmospheric dispersion modeling and by reactor core radionuclide inventory estimation in conjunction with WASH-1400 release fractions associated with specific chemical groups. The model simulations revealed that the radioactive cloud became segmented during the first day, with the lower section heading toward Scandinavia and the upper part heading in a southeasterly direction with subsequent transport across Asia to Japan, the North Pacific, and the west coast of North America. By optimizing the agreement between the observed cloud arrival times and duration of peak concentrations measured over Europe, Japan, Kuwait, and the US with the model predicted concentrations, it was possible to derive source term estimates for those radionuclides measured in airborne radioactivity. This was extended to radionuclides that were largely unmeasured in the environment by performing a reactor core radionuclide inventory analysis to obtain release fractions for the various chemical transport groups. These analyses indicated that essentially all of the noble gases, 60% of the radioiodines, 40% of the radiocesium, 10% of the tellurium and about 1% or less of the more refractory elements were released. These estimates are in excellent agreement with those obtained on the basis of worldwide deposition measurements. The Chernobyl source term was several orders of magnitude greater than those associated with the Windscale and TMI reactor accidents. However, the 137 Cs from the Chernobyl event is about 6% of that released by the US and USSR atmospheric nuclear weapon tests, while the 131 I and 90 Sr released by the Chernobyl accident was only about 0.1% of that released by the weapon tests. 13 refs., 2 figs., 7 tabs

  11. Estimation of Source Term Behaviors in SBO Sequence in a Typical 1000MWth PWR and Comparison with Other Source Term Results

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il; Fynan, Douglas; Jung, Yong Hoon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Since the Three Mile Island (TMI) (1979), Chernobyl (1986), Fukushima Daiichi (2011) accidents, the assessment of radiological source term effects on the environment has been a key concern of nuclear safety. In the Fukushima Daiichi accident, the long-term SBO (station blackout) accident occurs. Using the worst case assumptions like in Fukushima accident on the accident sequences and on the availability of safety systems, the thermal hydraulic behaviors, core relocation and environmental source terms behaviors are estimated for long-term SBO accident for OPR-1000 reactor. MELCOR code version 1.8.6 is used in this analysis. Source term results estimated in this study is compared with other previous studies and estimated results in Fukushima accidents in UNSCEAR-2013 report. This study estimated that 11 % of iodine can be released to environment and 2% of cesium can be released to environment. UNSCEAR-2013 report estimated that 2 - 8 % of iodine have been released to environment and 1 - 3 % of cesium have been released to the environment. They have similar results in the aspect of release fractions of iodine and cesium to environment.

  12. Real time source term and dose assessment

    International Nuclear Information System (INIS)

    Breznik, B.; Kovac, A.; Mlakar, P.

    2001-01-01

    The Dose Projection Programme is a tool for decision making in case of nuclear emergency. The essential input data for quick emergency evaluation in the case of hypothetical pressurised water reactor accident are following: source term, core damage assessment, fission product radioactivity, release source term and critical exposure pathways for an early phase of the release. A reduced number of radio-nuclides and simplified calculations can be used in dose calculation algorithm. Simple expert system personal computer programme has been developed for the Krsko Nuclear Power Plant for dose projection within the radius of few kilometers from the pressurised water reactor in early phase of an accident. The input data are instantaneous data of core activity, core damage indicators, release fractions, reduction factor of the release pathways, spray operation, release timing, and dispersion coefficient. Main dose projection steps are: accurate in-core radioactivity determination using reactor power input; core damage and in-containment source term assessment based on quick indications of instrumentation or on activity analysis data; user defines release pathway for typical PWR accident scenarius; dose calculation is performed only for exposure pathway critical for decision about evacuation or sheltering in early phase of an accident.(author)

  13. Release of radionuclides following severe accident in interim storage facility. Source term determination

    International Nuclear Information System (INIS)

    Morandi, S.; Mariani, M.; Giacobbo, F.; Covini, R.

    2006-01-01

    Among the severe accidents that can cause the release of radionuclides from an interim storage facility, with a consequent relevant radiological impact on the population, there is the impact of an aircraft on the facility. In this work, a safety assessment analysis for the case of an aircraft crash into an interim storage facility is tackled. To this aim a methodology, based upon DOE, IAEA and NUREG standard procedures and upon conservative yet realistic hypothesis, has been developed in order to evaluate the total radioactivity, source term, released to the biosphere in consequence of the impact, without recurring to the use of complicated numerical codes. The procedure consists in the identification of the accidental scenarios, in the evaluation of the consequent damage to the building structures and to the waste packages and in the determination of the total release of radionuclides through the building-atmosphere interface. The methodology here developed has been applied to the case of an aircraft crash into an interim storage facility currently under design. Results show that in case of perforation followed by a fire incident the total released activity would be greater of some orders of magnitude with respect to the case of mere perforation. (author)

  14. Review of SFR In-Vessel Radiological Source Term Studies

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2008-10-01

    An effort has been made in this study to search for and review the literatures in public domain on the studies of the phenomena related to the release of radionuclides and aerosols to the reactor containment of the sodium fast reactor (SFR) plants (i.e., in-vessel source term), made in Japan and Europe including France, Germany and UK over the last few decades. Review work is focused on the experimental programs to investigate the phenomena related to determining the source terms, with a brief review on supporting analytical models and computer programs. In this report, the research programs conducted to investigate the CDA (core disruptive accident) bubble behavior in the sodium pool for determining 'primary' or 'instantaneous' source term are first introduced. The studies performed to determine 'delayed source term' are then described, including the various stages of phenomena and processes: fission product (FP) release from fuel , evaporation release from the surface of the pool, iodine mass transfer from fission gas bubble, FP deposition , and aerosol release from core-concrete interaction. The research programs to investigate the release and transport of FPs and aerosols in the reactor containment (i.e., in-containment source term) are not described in this report

  15. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. Volume 3 is a compilation of appendices giving detailed results of the study

  16. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. 6 refs

  17. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. Volume 2 is a compilation of appendices giving detailed results of the study. 5 figs

  18. Effect of source term composition on offsite doses

    International Nuclear Information System (INIS)

    Karahalios, P.; Gardner, R.

    1985-01-01

    The development of new realistic accident source terms has identified the need to establish a basis for comparing the impact of such source terms. This paper attempts to develop a generalized basis of comparison by investigating contributions to offsite acute whole body doses from each group of radionuclides being released to the atmosphere, using CRAC2. The paper also investigates the effect of important parameters such as regional meteorology, sheltering, and duration of release. Finally, the paper focuses on significant changes in the relative importance of individual radionuclide groups in PWR2, SST1, and a revision of the Stone and Webster proposed interim source term

  19. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

    Barrett, P.R.; Foadian, H.; Rashid, Y.R.; Seager, K.D.; Gianoulakis, S.E.

    1993-01-01

    Containment of cask contents by a transport cask is a function of the cask body, one or more closure lids, and various bolting hardware, and seals associated with the cavity closure and other containment penetrations. In addition, characteristics of cask contents that impede the ability of radionuclides to move from an origin to the external environment also provide containment. In essence, multiple release barriers exist in series in transport casks, and the magnitude of the releasable activity in the cask is considerably lower than the total activity of its contents. A source term approach accounts for the magnitude of the releasable activity available in the cask by assessing the degree of barrier resistance to release provided by material characteristics and inherent barriers that impede the release of radioactive contents. Standardized methodologies for defining the spent-fuel transport packages with specified regulations have recently been developed. An essential part of applying the source term methodology involves characterizing the response of the spent fuel under regulatory conditions of transport. Thermal and structural models of the cask and fuel are analyzed and used to predict fuel rod failure probabilities. Input to these analyses and failure evaluations cover a wide range of geometrical and material properties. An important issue in the development of these models is the sensitivity of the radioactive source term generated during transport to individual parameters such as temperature and fluence level. This paper provides a summary of sensitivity analyses concentrating on the structural response and failure predictions of the spent fuel assemblies

  20. INEEL Subregional Conceptual Model Report Volume 3: Summary of Existing Knowledge of Natural and Anthropogenic Influences on the Release of Contaminants to the Subsurface Environment from Waste Source Terms at the INEEL

    Energy Technology Data Exchange (ETDEWEB)

    Paul L. Wichlacz

    2003-09-01

    This source-term summary document is intended to describe the current understanding of contaminant source terms and the conceptual model for potential source-term release to the environment at the Idaho National Engineering and Environmental Laboratory (INEEL), as presented in published INEEL reports. The document presents a generalized conceptual model of the sources of contamination and describes the general categories of source terms, primary waste forms, and factors that affect the release of contaminants from the waste form into the vadose zone and Snake River Plain Aquifer. Where the information has previously been published and is readily available, summaries of the inventory of contaminants are also included. Uncertainties that affect the estimation of the source term release are also discussed where they have been identified by the Source Term Technical Advisory Group. Areas in which additional information are needed (i.e., research needs) are also identified.

  1. Spallation Neutron Source Accident Terms for Environmental Impact Statement Input

    Energy Technology Data Exchange (ETDEWEB)

    Devore, J.R.; Harrington, R.M.

    1998-08-01

    This report is about accidents with the potential to release radioactive materials into the environment surrounding the Spallation Neutron Source (SNS). As shown in Chap. 2, the inventories of radioactivity at the SNS are dominated by the target facility. Source terms for a wide range of target facility accidents, from anticipated events to worst-case beyond-design-basis events, are provided in Chaps. 3 and 4. The most important criterion applied to these accident source terms is that they should not underestimate potential release. Therefore, conservative methodology was employed for the release estimates. Although the source terms are very conservative, excessive conservatism has been avoided by basing the releases on physical principles. Since it is envisioned that the SNS facility may eventually (after about 10 years) be expanded and modified to support a 4-MW proton beam operational capability, the source terms estimated in this report are applicable to a 4-MW operating proton beam power unless otherwise specified. This is bounding with regard to the 1-MW facility that will be built and operated initially. See further discussion below in Sect. 1.2.

  2. Selection of models to calculate the LLW source term

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1991-10-01

    Performance assessment of a LLW disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). In turn, many of these physical processes are influenced by the design of the disposal facility (e.g., infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This document provides a brief overview of disposal practices and reviews existing source term models as background for selecting appropriate models for estimating the source term. The selection rationale and the mathematical details of the models are presented. Finally, guidance is presented for combining the inventory data with appropriate mechanisms describing release from the disposal facility. 44 refs., 6 figs., 1 tab

  3. A simplified approach to evaluating severe accident source term for PWR

    International Nuclear Information System (INIS)

    Huang, Gaofeng; Tong, Lili; Cao, Xuewu

    2014-01-01

    Highlights: • Traditional source term evaluation approaches have been studied. • A simplified approach of source term evaluation for 600 MW PWR is studied. • Five release categories are established. - Abstract: For early design of NPPs, no specific severe accident source term evaluation was considered. Some general source terms have been used for some NPPs. In order to implement a best estimate, a special source term evaluation should be implemented for an NPP. Traditional source term evaluation approaches (mechanism approach and parametric approach) have some difficulties associated with their implementation. The traditional approaches are not consistent with cost-benefit assessment. A simplified approach for evaluating severe accident source term for PWR is studied. For the simplified approach, a simplified containment event tree is established. According to representative cases selection, weighted coefficient evaluation, computation of representative source term cases and weighted computation, five containment release categories are established, including containment bypass, containment isolation failure, containment early failure, containment late failure and intact containment

  4. Hanford tank residual waste - Contaminant source terms and release models

    International Nuclear Information System (INIS)

    Deutsch, William J.; Cantrell, Kirk J.; Krupka, Kenneth M.; Lindberg, Michael L.; Jeffery Serne, R.

    2011-01-01

    Highlights: → Residual waste from five Hanford spent fuel process storage tanks was evaluated. → Gibbsite is a common mineral in tanks with high Al concentrations. → Non-crystalline U-Na-C-O-P ± H phases are common in the U-rich residual. → Iron oxides/hydroxides have been identified in all residual waste samples. → Uranium release is highly dependent on waste and leachant compositions. - Abstract: Residual waste is expected to be left in 177 underground storage tanks after closure at the US Department of Energy's Hanford Site in Washington State, USA. In the long term, the residual wastes may represent a potential source of contamination to the subsurface environment. Residual materials that cannot be completely removed during the tank closure process are being studied to identify and characterize the solid phases and estimate the release of contaminants from these solids to water that might enter the closed tanks in the future. As of the end of 2009, residual waste from five tanks has been evaluated. Residual wastes from adjacent tanks C-202 and C-203 have high U concentrations of 24 and 59 wt.%, respectively, while residual wastes from nearby tanks C-103 and C-106 have low U concentrations of 0.4 and 0.03 wt.%, respectively. Aluminum concentrations are high (8.2-29.1 wt.%) in some tanks (C-103, C-106, and S-112) and relatively low ( 2 -saturated solution, or a CaCO 3 -saturated water. Uranium release concentrations are highly dependent on waste and leachant compositions with dissolved U concentrations one or two orders of magnitude higher in the tests with high U residual wastes, and also higher when leached with the CaCO 3 -saturated solution than with the Ca(OH) 2 -saturated solution. Technetium leachability is not as strongly dependent on the concentration of Tc in the waste, and it appears to be slightly more leachable by the Ca(OH) 2 -saturated solution than by the CaCO 3 -saturated solution. In general, Tc is much less leachable (<10 wt.% of the

  5. NRC source term assessment for incident response dose projections

    International Nuclear Information System (INIS)

    Easley, P.; Pasedag, W.

    1984-01-01

    The NRC provides advice and assistance to licensees and State and local authorities in responding to accidents. The TACT code supports this function by providing source term projections for two situations during early (15 to 60 minutes) accident response: (1) Core/containment damage is indicated, but there are no measured releases. Quantification of a predicted release permits emergency response before people are exposed. With TACT, response personnel can estimate releases based on fuel and cladding conditions, coolant boundary and containment integrity, and mitigative systems operability. For this type of estimate, TACT is intermediate between default assumptions and time-consuming mechanistic codes. (2) A combination of plant status and limited release data are available. For this situation, iterations between predictions based on known conditions which are compared to measured releases gives reasonable confidence in supplemental source term information otherwise unavailable: nuclide mix, releases not monitored, and trending or abrupt changes. The assumptions and models used in TACT, and examples of its use, are given in this paper

  6. An investigation of the closure problem applied to reactor accident source terms

    International Nuclear Information System (INIS)

    Brearley, I.R.; Nixon, W.; Hayns, M.R.

    1987-01-01

    The closure problem, as considered here, focuses attention on the question of when in current research programmes enough has been learned about the source terms for reactor accident releases. Noting that current research is tending to reduce the estimated magnitude of the aerosol component of atmospheric, accidental releases, several possible criteria for closure are suggested. Moreover, using the reactor accident consequence model CRACUK, the effect of gradually reducing the aerosol release fractions of a pressurized water reactor (PWR2) source term (as defined in the WASH-1400 study) is investigated and the implications of applying the suggested criteria to current source term research discussed. (author)

  7. The latest results from source term research. Overview and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E. [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica (CIEMAT), Madrid (Spain); Haste, Tim [Centre d' Etudes de Cadarache, Paul-Lez-Durance (France). Institut de Radioprotection et de Surete Nucleaire (IRSN); Kaerkelae, Teemu [VTT Technical Research Centre of Finland Ltd, Espoo (Finland)

    2016-12-15

    Source term research has continued internationally for more than 30 years, increasing confidence in calculations of the potential radioactive release to the environment after a severe reactor accident. Important experimental data have been obtained, mainly under international frameworks such as OECD/NEA and EURATOM. Specifically, Phebus FP provides major insights into fission product release and transport. Results are included in severe accident analysis codes. Data from international projects are being interpreted with a view to further improvements in these codes. This paper synthesizes the recent main outcomes from source term research on these topics, and on source term mitigation. It highlights knowledge gaps remaining and discusses ways to proceed. Aside from this further knowledge-driven research, there is consensus on the need to assess the source term predictive ability of current system codes, taking account of scale-up from experiment to reactor conditions.

  8. ITER task title - source term data, modelling, and analysis. ITER subtask no. S81TT05/5 (SEP 1-1). Global tritium source term analysis basis document. Subtask 1: operational tritium effluents and releases. Final report (1995 TASK)

    International Nuclear Information System (INIS)

    Kalyanam, K.M.

    1996-06-01

    This document represents the final report for the global tritium source term analysis task initiated in 1995. The report presents a room-by-room map/table at the subsystem level for the ITER tritium systems, identifying the major equipment, secondary containments, tritium release sources, duration/frequency of tritium releases and the release pathways. The chronic tritium releases during normal operation, as well as tritium releases due to routine maintenance of the Water Distillation Unit, Isotope Separation System and Primary and Secondary Heat Transport Systems, have been estimated for most of the subsystems, based on the IDR design, the Design Description Documents (April - Jun 1995 issues) and the design updates up to December 1995. The report also outlines the methodology and the key assumptions that are adopted in preparing the tritium release estimates. The design parameters for the ITER Basic Performance Phase (BPP) have been used in estimating the tritium releases shown in the room-by-room map/table. The tritium release calculations and the room-by-room map/table have been prepared in EXCEL, so that the estimates can be refined easily as the design evolves and more detailed information becomes available. (author). 23 refs., tabs

  9. Development of a general model to predict the rate of radionuclide release (source term) from a low-level waste shallow land burial facility

    International Nuclear Information System (INIS)

    Sullivan, T.M.; Kempf, C.R.; Suen, C.J.; Mughabghab, S.M.

    1988-01-01

    Federal Code of Regulations 10 CFR 61 requires that any near surface disposal site be capable of being characterized, analyzed, and modeled. The objective of this program is to assist NRC in developing the ability to model a disposal site that conforms to these regulations. In particular, a general computer model capable of predicting the quantity and rate of radionuclide release from a shallow land burial trench, i.e., the source term, is being developed. The framework for this general model has been developed and consists of four basic compartments that represent the major processes that influence release. These compartments are: water flow, container degradation, release from the waste packages, and radionuclide transport. Models for water flow and radionuclide transport rely on the use of the computer codes FEMWATER and FEMWASTE. These codes are generally regarded as being state-of-the-art and required little modification for their application to this project. Models for container degradation and release from waste packages have been specifically developed for this project. This paper provides a brief description of the models being used in the source term project and examples of their use over a range of potential conditions. 13 refs

  10. Considerations about source term now used aiming to emergency planning

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1987-01-01

    The applicability of source terms, in parametric studies for improving external emergengy plan for Angra-I reactor is presented. The source term is defined as, the quantity and radioactive material disposable for releasing to the environment in case of austere accident in a nuclear power plant. The following hypothesis: occuring accident, 100% of the noble gases, 50% of halogens and 1% of solid fission products contained into the reactor core, are released immediately toward the containment building; the radioactivity releasing to the environment is done at a constant rate of 0.1% in mass per day; the actuation of mitigated systems of radioactivity releasing, such as, spray of container or system of air recirculation by filters, is not considered; and the releasing is done at soil level. (M.C.K.) [pt

  11. New Source Term Model for the RESRAD-OFFSITE Code Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Charley [Argonne National Lab. (ANL), Argonne, IL (United States); Gnanapragasam, Emmanuel [Argonne National Lab. (ANL), Argonne, IL (United States); Cheng, Jing-Jy [Argonne National Lab. (ANL), Argonne, IL (United States); Kamboj, Sunita [Argonne National Lab. (ANL), Argonne, IL (United States); Chen, Shih-Yew [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-06-01

    This report documents the new source term model developed and implemented in Version 3 of the RESRAD-OFFSITE code. This new source term model includes: (1) "first order release with transport" option, in which the release of the radionuclide is proportional to the inventory in the primary contamination and the user-specified leach rate is the proportionality constant, (2) "equilibrium desorption release" option, in which the user specifies the distribution coefficient which quantifies the partitioning of the radionuclide between the solid and aqueous phases, and (3) "uniform release" option, in which the radionuclides are released from a constant fraction of the initially contaminated material during each time interval and the user specifies the duration over which the radionuclides are released.

  12. STACE: Source Term Analyses for Containment Evaluations of transport casks

    International Nuclear Information System (INIS)

    Seager, K.D.; Gianoulakis, S.E.; Barrett, P.R.; Rashid, Y.R.; Reardon, P.C.

    1992-01-01

    Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source term has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volatile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking (e.g., the quantity and size distribution of fuel rod breaches) in which experimental validation is planned. The CRUD spallation fraction is the major area where no quantitative data has been found; therefore, this also requires experimental validation. In the interim, STACE conservatively assumes a 100% spallation fraction for computing the releasable activity. The source term methodology also conservatively assumes that there is 1 Ci of residual contamination available for release in the transport cask. However, residual contamination is still by far the smallest contributor to the source term activity

  13. Source term estimation during incident response to severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    McKenna, T.J.; Glitter, J.G.

    1988-10-01

    This document presents a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. 39 refs., 48 figs., 19 tabs

  14. Source term estimation and the isotopic ratio of radioactive material released from the WIPP repository in New Mexico, USA

    International Nuclear Information System (INIS)

    Thakur, P.

    2016-01-01

    After almost 15 years of operations, the Waste Isolation Pilot Plant (WIPP) had one of its waste drums breach underground as a result of a runaway chemical reaction in the waste it contained. This incident occurred on February 14, 2014. Moderate levels of radioactivity were released into the underground air. A small portion of the contaminated underground air also escaped to the surface through the ventilation system and was detected approximately 1 km away from the facility. According to the source term estimation, the actual amount of radioactivity released from the WIPP site was less than 1.5 mCi. The highest activity detected on the surface was 115.2 μBq/m 3 for 241 Am and 10.2 μBq/m 3 for 239+240 Pu at a sampling station located 91 m away from the underground air exhaust point and 81.4 μBq/m 3 of 241 Am and 5.8 μBq/m 3 of 239+240 Pu at a monitoring station located approximately 1 km northwest of the WIPP facility. The dominant radionuclides released were americium and plutonium, in a ratio that matches the content of the breached drum. Air monitoring across the WIPP site intensified following the first reports of radiation detection underground to determine the extent of impact to WIPP personnel, the public, and the environment. In this paper, the early stage monitoring data collected by an independent monitoring program conducted by the Carlsbad Environmental Monitoring & Research Center (CEMRC) and an oversight monitoring program conducted by the WIPP's management and operating contractor, the Nuclear Waste Partnership (NWP) LLC were utilized to estimate the actual amount of radioactivity released from the WIPP underground. The Am and Pu isotope ratios were measured and used to support the hypothesis that the release came from one drum identified as having breached that represents a specific waste stream with this radionuclide ratio in its inventory. This failed drum underwent a heat and gas producing reaction that overpowered its vent and

  15. Radiological and chemical source terms for Solid Waste Operations Complex

    International Nuclear Information System (INIS)

    Boothe, G.F.

    1994-01-01

    The purpose of this document is to describe the radiological and chemical source terms for the major projects of the Solid Waste Operations Complex (SWOC), including Project W-112, Project W-133 and Project W-100 (WRAP 2A). For purposes of this document, the term ''source term'' means the design basis inventory. All of the SWOC source terms involve the estimation of the radiological and chemical contents of various waste packages from different waste streams, and the inventories of these packages within facilities or within a scope of operations. The composition of some of the waste is not known precisely; consequently, conservative assumptions were made to ensure that the source term represents a bounding case (i.e., it is expected that the source term would not be exceeded). As better information is obtained on the radiological and chemical contents of waste packages and more accurate facility specific models are developed, this document should be revised as appropriate. Radiological source terms are needed to perform shielding and external dose calculations, to estimate routine airborne releases, to perform release calculations and dose estimates for safety documentation, to calculate the maximum possible fire loss and specific source terms for individual fire areas, etc. Chemical source terms (i.e., inventories of combustible, flammable, explosive or hazardous chemicals) are used to determine combustible loading, fire protection requirements, personnel exposures to hazardous chemicals from routine and accident conditions, and a wide variety of other safety and environmental requirements

  16. The Fukushima releases: an inverse modelling approach to assess the source term by using gamma dose rate observations

    Science.gov (United States)

    Saunier, Olivier; Mathieu, Anne; Didier, Damien; Tombette, Marilyne; Quélo, Denis; Winiarek, Victor; Bocquet, Marc

    2013-04-01

    The Chernobyl nuclear accident and more recently the Fukushima accident highlighted that the largest source of error on consequences assessment is the source term estimation including the time evolution of the release rate and its distribution between radioisotopes. Inverse modelling methods have proved to be efficient to assess the source term due to accidental situation (Gudiksen, 1989, Krysta and Bocquet, 2007, Stohl et al 2011, Winiarek et al 2012). These methods combine environmental measurements and atmospheric dispersion models. They have been recently applied to the Fukushima accident. Most existing approaches are designed to use air sampling measurements (Winiarek et al, 2012) and some of them use also deposition measurements (Stohl et al, 2012, Winiarek et al, 2013). During the Fukushima accident, such measurements are far less numerous and not as well distributed within Japan than the dose rate measurements. To efficiently document the evolution of the contamination, gamma dose rate measurements were numerous, well distributed within Japan and they offered a high temporal frequency. However, dose rate data are not as easy to use as air sampling measurements and until now they were not used in inverse modelling approach. Indeed, dose rate data results from all the gamma emitters present in the ground and in the atmosphere in the vicinity of the receptor. They do not allow one to determine the isotopic composition or to distinguish the plume contribution from wet deposition. The presented approach proposes a way to use dose rate measurement in inverse modeling approach without the need of a-priori information on emissions. The method proved to be efficient and reliable when applied on the Fukushima accident. The emissions for the 8 main isotopes Xe-133, Cs-134, Cs-136, Cs-137, Ba-137m, I-131, I-132 and Te-132 have been assessed. The Daiichi power plant events (such as ventings, explosions…) known to have caused atmospheric releases are well identified in

  17. Accident source terms for Light-Water Nuclear Power Plants. Final report

    International Nuclear Information System (INIS)

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ''Calculation of Distance Factors for Power and Test Reactors'' which specified a release of fission products from the core to the reactor containment for a postulated accident involving ''substantial meltdown of the core''. This ''source term'', tile basis for tile NRC's Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC's reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ''source term'' release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ''source term'' is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it

  18. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    Energy Technology Data Exchange (ETDEWEB)

    McKenna, T J; Giitter, J

    1987-07-01

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

  19. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    International Nuclear Information System (INIS)

    McKenna, T.J.; Giitter, J.

    1987-01-01

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

  20. Source terms in relation to air cleaning

    International Nuclear Information System (INIS)

    Bernero, R.M.

    1985-01-01

    There are two sets of source terms for consideration in air cleaning, those for routine releases and those for accident releases. With about 1000 reactor years of commercial operating experience in the US done, there is an excellent data base for routine and expected transient releases. Specifications for air cleaning can be based on this body of experience with confidence. Specifications for air cleaning in accident situations is another matter. Recent investigations of severe accident behavior are offering a new basis for source terms and air cleaning specifications. Reports by many experts in the field describe an accident environment notably different from previous models. It is an atmosphere heavy with aerosols, both radioactive and inert. Temperatures are sometimes very high; radioiodine is typically in the form of cesium iodide aerosol particles; other nuclides, such as tellurium, are also important aerosols. Some of the present air cleaning requirements may be very important in light of these new accident behavior models. Others may be wasteful or even counterproductive. The use of the new data on accident behavior models to reevaluate requirements promptly is discussed

  1. Evolution of source term definition and analysis

    International Nuclear Information System (INIS)

    Lutz, R.J. Jr.

    2004-01-01

    The objective of this presentation was to provide an overview of the evolution of accident fission product release analysis methodology and the obtained results; and to provide an overview of the source term implementation analysis in regulatory decisions

  2. Determination of source term for Krsko NPP extended fuel cycle

    International Nuclear Information System (INIS)

    Nemec, T.; Persic, A.; Zagar, T.; Zefran, B.

    2004-01-01

    The activity and composition of the potential radioactive releases (source term) is important in the decision making about off-site emergency measures in case of a release into environment. Power uprate of Krsko NPP during modernization in 2000 as well as changing of the fuel type and the core design have influenced the source term value. In 2003 a project of 'Jozef Stefan' Institute and Slovenian nuclear safety administration determined a plantspecific source term for new conditions of fuel type and burnup for extended fuel cycle. Calculations of activity and isotopic composition of the core have been performed with ORIGEN-ARP program. Results showed that the core activity for extended 15 months fuel cycle is slightly lower than for the 12 months cycles, mainly due to larger share of fresh fuel. (author)

  3. Haw-glass dissolution and radionuclide release: mechanism - modelling - source term

    Energy Technology Data Exchange (ETDEWEB)

    Grambow, B [Forschungszentrum Karlsruhe, Institut fur Nukleare, Karlsruhe (Germany)

    1997-07-01

    Important release controlling processes are: 1) kinetics of glass matrix dissolution (leaching), 2) formation of secondary alteration products (controlling thermodynamic solubility), 3) sorption on surfaces in the engineered barrier system and 4) formation of mobile species. Quantification of these processes requires assessment of the energetics and dynamics of the various reversible and irreversible processes within an overall open non-equilibrium system. Corrosion/dissolution of the waste matrices is not necessarily associated with a proportional release of radionuclides. The formation of new secondary phases, such as silicates, molybdates, uranates, carbonates... establishes a new geochemical barrier for re-immobilization of radionuclides dissolved from the waste matrices. The presence of iron (corroding canisters during glass alteration) reduces the solution concentration of redox sensitive radionuclides. Consequently, the container, after being corroded, constitutes an important geochemical barrier for radionuclide re-immobilization. Geochemical modelling of the long-term behaviour of glasses must be performed in an integrated way, considering simultaneous reactions of the glass, of container corrosion, of repository rock and of backfill material. Until now, only few attempts were made for integral systems modelling. (A.C.)

  4. Source term model evaluations for the low-level waste facility performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Yim, M.S.; Su, S.I. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The estimation of release of radionuclides from various waste forms to the bottom boundary of the waste disposal facility (source term) is one of the most important aspects of LLW facility performance assessment. In this work, several currently used source term models are comparatively evaluated for the release of carbon-14 based on a test case problem. The models compared include PRESTO-EPA-CPG, IMPACTS, DUST and NEFTRAN-II. Major differences in assumptions and approaches between the models are described and key parameters are identified through sensitivity analysis. The source term results from different models are compared and other concerns or suggestions are discussed.

  5. Bayesian source term determination with unknown covariance of measurements

    Science.gov (United States)

    Belal, Alkomiet; Tichý, Ondřej; Šmídl, Václav

    2017-04-01

    Determination of a source term of release of a hazardous material into the atmosphere is a very important task for emergency response. We are concerned with the problem of estimation of the source term in the conventional linear inverse problem, y = Mx, where the relationship between the vector of observations y is described using the source-receptor-sensitivity (SRS) matrix M and the unknown source term x. Since the system is typically ill-conditioned, the problem is recast as an optimization problem minR,B(y - Mx)TR-1(y - Mx) + xTB-1x. The first term minimizes the error of the measurements with covariance matrix R, and the second term is a regularization of the source term. There are different types of regularization arising for different choices of matrices R and B, for example, Tikhonov regularization assumes covariance matrix B as the identity matrix multiplied by scalar parameter. In this contribution, we adopt a Bayesian approach to make inference on the unknown source term x as well as unknown R and B. We assume prior on x to be a Gaussian with zero mean and unknown diagonal covariance matrix B. The covariance matrix of the likelihood R is also unknown. We consider two potential choices of the structure of the matrix R. First is the diagonal matrix and the second is a locally correlated structure using information on topology of the measuring network. Since the inference of the model is intractable, iterative variational Bayes algorithm is used for simultaneous estimation of all model parameters. The practical usefulness of our contribution is demonstrated on an application of the resulting algorithm to real data from the European Tracer Experiment (ETEX). This research is supported by EEA/Norwegian Financial Mechanism under project MSMT-28477/2014 Source-Term Determination of Radionuclide Releases by Inverse Atmospheric Dispersion Modelling (STRADI).

  6. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  7. Centrifugal Filtration System for Severe Accident Source Term Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Shu Chang; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this paper is to present the conceptual design of a filtration system that can be used to process airborne severe accident source term. Reactor containment may lose its structural integrity due to over-pressurization during a severe accident. This can lead to uncontrolled radioactive releases to the environment. For preventing the dispersion of these uncontrolled radioactive releases to the environment, several ways to capture or mitigate these radioactive source term releases are under investigation at KAIST. Such technologies are based on concepts like a vortex-like air curtain, a chemical spray, and a suction arm. Treatment of the radioactive material captured by these systems would be required, before releasing to environment. For current filtration systems in the nuclear industry, IAEA lists sand, multi-venturi scrubber, high efficiency particulate arresting (HEPA), charcoal and combinations of the above in NS-G-1-10, 4.143. Most if not all of the requirements of the scenario for applying this technology near the containment of an NPP site and the environmental constraints were analyzed for use in the design of the centrifuge filtration system.

  8. Source term and radiological consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Mourad, R.

    1987-09-01

    This report presents the results of a study of the source term and radiological consequences of the Chernobyl accident. The results two parts. The first part was performed during the first 2 months following the accident and dealt with the evaluation of the source term and an estimate of individual doses in the European countries outside the Soviet Union. The second part was performed after August 25-29, 1986 when the Soviets presented in a IAEA Conference in Vienna detailed information about the accident, including source term and radiological consequences in the Soviet Union. The second part of the study reconfirms the source term evaluated in the first part and in addition deals with the radiological consequences in the Soviet Union. Source term and individual doses are calculated from measured post-accident data, reported by the Soviet Union and European countries, microcomputer program PEAR (Public Exposure from Accident Releases). 22 refs

  9. Review of specific effects in atmospheric dispersion calculations. The impact of source-term characteristics -and the processes that modify them post release- on dry and wet deposition rates

    International Nuclear Information System (INIS)

    Cooper, P.J.; Underwood, B.Y.; Brearley, I.

    1985-01-01

    In the first half of the work the source-term characteristics potentially influencing behaviour were identified and examined. It was concluded that a number of source characteristics, in addition to those conventionally provided for consequence assessment, could significantly influence deposition behaviour. Linking with this, a review was undertaken of past reactor-accident risk assessment and more recent source-term studies to pick out information, if any, on the parameters of interest. The second half of the study resulted in a list of processes capable of transforming the released material vis-a-vis deposition characteristics, including processes occurring in the near field associated with the initial release transient and also those occurring over a longer time span as the plume travels downwind. Scoping calculations were performed for some of the processes in the context of idealized accident scenarios, leading to the conclusions that in some circumstances post-release mechanisms could have an important impact on the deposition behaviour of released material. Statistical theory was used to describe the behaviour of a plume both before and after detachment, and the limitations of the theory were discussed. A review of the lateral wind velocity spectra was undertaken so that simplified spectra could be constructed and used to predict the plume behaviour as a function of travel time, stability category and release duration. It was found that commonly used methods of allowing for release duration overpredicted the dependence, in general, upon release duration. For example the adoption of a stability-independent meandering term would lead to the underprediction of threshold effects such as early death and land/crop interdiction. In addition, theory indicated that the 'Y' curves for different stability categories would converge gradually with increasing travel time

  10. The Chernobyl reactor accident source term: development of a consensus view

    International Nuclear Information System (INIS)

    Devell, L.; Guntay, S.; Powers, D.A.

    1995-11-01

    Ten years after the reactor accident at Chernobyl, a great deal more data is available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident, a task that is substantially more difficult than it might first appear to be. The Chernobyl station, like other nuclear power plants, was not instrumented to characterize a disastrous accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for several days. Characterization of the contamination caused by the releases of radioactivity has had a much lower priority than remediation of the contamination. Consequently, an assessment of the Chernobyl accident source term must rely to a significant extent on inferential evidence. The assessment presented here begins with an examination of the core inventories of radioactive materials. In subsequent sections of the report, the magnitude and timing of the releases of radioactivity are described. Then, the composition, chemical forms, and physical forms of the releases are discussed. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved the understanding of the Chernobyl source term. Because of the special features of the reactor design and the peculiarities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability to the safety analysis of other types of reactors

  11. A nuclear source term analysis for spacecraft power systems

    International Nuclear Information System (INIS)

    McCulloch, W.H.

    1998-01-01

    All US space missions involving on board nuclear material must be approved by the Office of the President. To be approved the mission and the hardware systems must undergo evaluations of the associated nuclear health and safety risk. One part of these evaluations is the characterization of the source terms, i.e., the estimate of the amount, physical form, and location of nuclear material, which might be released into the environment in the event of credible accidents. This paper presents a brief overview of the source term analysis by the Interagency Nuclear Safety Review Panel for the NASA Cassini Space Mission launched in October 1997. Included is a description of the Energy Interaction Model, an innovative approach to the analysis of potential releases from high velocity impacts resulting from launch aborts and reentries

  12. New Open-Source Version of FLORIS Released | News | NREL

    Science.gov (United States)

    New Open-Source Version of FLORIS Released New Open-Source Version of FLORIS Released January 26 , 2018 National Renewable Energy Laboratory (NREL) researchers recently released an updated open-source simplified and documented. Because of the living, open-source nature of the newly updated utility, NREL

  13. Source term identification in atmospheric modelling via sparse optimization

    Science.gov (United States)

    Adam, Lukas; Branda, Martin; Hamburger, Thomas

    2015-04-01

    Inverse modelling plays an important role in identifying the amount of harmful substances released into atmosphere during major incidents such as power plant accidents or volcano eruptions. Another possible application of inverse modelling lies in the monitoring the CO2 emission limits where only observations at certain places are available and the task is to estimate the total releases at given locations. This gives rise to minimizing the discrepancy between the observations and the model predictions. There are two standard ways of solving such problems. In the first one, this discrepancy is regularized by adding additional terms. Such terms may include Tikhonov regularization, distance from a priori information or a smoothing term. The resulting, usually quadratic, problem is then solved via standard optimization solvers. The second approach assumes that the error term has a (normal) distribution and makes use of Bayesian modelling to identify the source term. Instead of following the above-mentioned approaches, we utilize techniques from the field of compressive sensing. Such techniques look for a sparsest solution (solution with the smallest number of nonzeros) of a linear system, where a maximal allowed error term may be added to this system. Even though this field is a developed one with many possible solution techniques, most of them do not consider even the simplest constraints which are naturally present in atmospheric modelling. One of such examples is the nonnegativity of release amounts. We believe that the concept of a sparse solution is natural in both problems of identification of the source location and of the time process of the source release. In the first case, it is usually assumed that there are only few release points and the task is to find them. In the second case, the time window is usually much longer than the duration of the actual release. In both cases, the optimal solution should contain a large amount of zeros, giving rise to the

  14. Quantification of severe accident source terms of a Westinghouse 3-loop plant

    International Nuclear Information System (INIS)

    Lee Min; Ko, Y.-C.

    2008-01-01

    Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment

  15. Examination of Conservatism in Ground-level Source Release Assumption when Performing Consequence Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-yeop; Lim, Ho-Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    One of these assumptions frequently assumed is the assumption of ground-level source release. The user manual of a consequence analysis software HotSpot is mentioning like below: 'If you cannot estimate or calculate the effective release height, the actual physical release height (height of the stack) or zero for ground-level release should be used. This will usually yield a conservative estimate, (i.e., larger radiation doses for all downwind receptors, etc).' This recommendation could be agreed in aspect of conservatism but quantitative examination of the effect of this assumption to the result of consequence analysis is necessary. The source terms of Fukushima Dai-ichi NPP accident have been estimated by several studies using inverse modeling and one of the biggest sources of the difference between the results of these studies was different effective source release height assumed by each studies. It supports the importance of the quantitative examination of the influence by release height. Sensitivity analysis of the effective release height of radioactive sources was performed and the influence to the total effective dose was quantitatively examined in this study. Above 20% difference is maintained even at longer distances, when we compare the dose between the result assuming ground-level release and the results assuming other effective plume height. It means that we cannot ignore the influence of ground-level source assumption to the latent cancer fatality estimations. In addition, the assumption of ground-level release fundamentally prevents detailed analysis including diffusion of plume from effective plume height to the ground even though the influence of it is relatively lower in longer distance. When we additionally consider the influence of surface roughness, situations could be more serious. The ground level dose could be highly over-estimated in short downwind distance at the NPP sites which have low surface roughness such as Barakah site in

  16. Bioaerosol releases from compost facilities: Evaluating passive and active source terms at a green waste facility for improved risk assessments

    Science.gov (United States)

    Taha, M. P. M.; Drew, G. H.; Longhurst, P. J.; Smith, R.; Pollard, S. J. T.

    The passive and active release of bioaerosols during green waste composting, measured at source is reported for a commercial composting facility in South East (SE) England as part of a research programme focused on improving risk assessments at composting facilities. Aspergillus fumigatus and actinomycetes concentrations of 9.8-36.8×10 6 and 18.9-36.0×10 6 cfu m -3, respectively, measured during the active turning of green waste compost, were typically 3-log higher than previously reported concentrations from static compost windrows. Source depletion curves constructed for A. fumigatus during compost turning and modelled using SCREEN3 suggest that bioaerosol concentrations could reduce to background concentrations of 10 3 cfu m -3 within 100 m of this site. Authentic source term data produced from this study will help to refine the risk assessment methodologies that support improved permitting of compost facilities.

  17. The Analytical Repository Source-Term (AREST) model: Description and documentation

    International Nuclear Information System (INIS)

    Liebetrau, A.M.; Apted, M.J.; Engel, D.W.; Altenhofen, M.K.; Strachan, D.M.; Reid, C.R.; Windisch, C.F.; Erikson, R.L.; Johnson, K.I.

    1987-10-01

    The geologic repository system consists of several components, one of which is the engineered barrier system. The engineered barrier system interfaces with natural barriers that constitute the setting of the repository. A model that simulates the releases from the engineered barrier system into the natural barriers of the geosphere, called a source-term model, is an important component of any model for assessing the overall performance of the geologic repository system. The Analytical Repository Source-Term (AREST) model being developed is one such model. This report describes the current state of development of the AREST model and the code in which the model is implemented. The AREST model consists of three component models and five process models that describe the post-emplacement environment of a waste package. All of these components are combined within a probabilistic framework. The component models are a waste package containment (WPC) model that simulates the corrosion and degradation processes which eventually result in waste package containment failure; a waste package release (WPR) model that calculates the rates of radionuclide release from the failed waste package; and an engineered system release (ESR) model that controls the flow of information among all AREST components and process models and combines release output from the WPR model with failure times from the WPC model to produce estimates of total release. 167 refs., 40 figs., 12 tabs

  18. Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David S. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, Acacia Joann [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew D. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-02-28

    Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and

  19. Data assimilation and source term estimation during the early phase of a nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Golubenkov, A.; Borodin, R. [SPA Typhoon, Emergency Centre (Russian Federation); Sohier, A.; Rojas Palma, C. [Centre de l`Etude de l`Energie Nucleaire, Mol (Belgium)

    1996-02-01

    The mathematical/physical base of possible methods to model the source term during an accidental release of radionuclides is discussed. Knowledge of the source term is important in view of optimizing urgent countermeasures to the population. In most cases however, it will be impossible to assess directly the release dynamics. Therefore methods are under development in which the source term is modelled, based on the comparison of off-site monitoring data and model predictions using an atmospheric dispersion model. The degree of agreement between the measured and calculated characteristics of the radioactive contamination of the air and the ground surface is an important criterion in this process. Due to the inherent complexity, some geometrical transformations taking space-time discrepancies between observed and modelled contamination fields are defined before the source term is adapted. This work describes the developed algorithms which are also tested against data from some tracer experiments performed in the past. This method is also used to reconstruct the dynamics of the Chernobyl source term. Finally this report presents a concept of software to reconstruct a multi-isotopic source term in real-time.

  20. Data assimilation and source term estimation during the early phase of a nuclear accident

    International Nuclear Information System (INIS)

    Golubenkov, A.; Borodin, R.; Sohier, A.; Rojas Palma, C.

    1996-02-01

    The mathematical/physical base of possible methods to model the source term during an accidental release of radionuclides is discussed. Knowledge of the source term is important in view of optimizing urgent countermeasures to the population. In most cases however, it will be impossible to assess directly the release dynamics. Therefore methods are under development in which the source term is modelled, based on the comparison of off-site monitoring data and model predictions using an atmospheric dispersion model. The degree of agreement between the measured and calculated characteristics of the radioactive contamination of the air and the ground surface is an important criterion in this process. Due to the inherent complexity, some geometrical transformations taking space-time discrepancies between observed and modelled contamination fields are defined before the source term is adapted. This work describes the developed algorithms which are also tested against data from some tracer experiments performed in the past. This method is also used to reconstruct the dynamics of the Chernobyl source term. Finally this report presents a concept of software to reconstruct a multi-isotopic source term in real-time

  1. Severe accident source term reassessment

    International Nuclear Information System (INIS)

    Hazzan, M.J.; Gardner, R.; Warman, E.A.; Jacobs, S.B.

    1987-01-01

    This paper summarizes the status of the reassessment of severe reactor accident source terms, which are defined as the quantity, type, and timing of fission product releases from such accidents. Concentration is on the major results and conclusions of analyses with modern methods for both pressurized water reactors (PWRs) and boiling water reactors (BWRs), and the special case of containment bypass. Some distinctions are drawn between analyses for PWRs and BWRs. In general, the more the matter is examined, the consequences, or probability of serious consequences, seem to be less. (author)

  2. Evaluation of short- and long-term fission product sources at the Fukushima Daiichi NPP

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Naitoh, Masanori; Suzuki, Hiroaki; Okada, Hidetoshi; Pellegrini, Marco; Achilli, Andrea; Hanamoto, Yukio; Sasaki, Hiroaki

    2014-01-01

    Research on fission product (FP) behaviors used to be one of the most important subjects in water chemistry but it is not done nowadays as a consequence of the increased integrity of nuclear fuels and the minimization of FP release into the environment. Evaluation of FP release into the environment is still one of the key issues for severe accident analysis, though. Although there have been a long quiet period in nuclear safety research, how to detect initiation of severe accidents, how to prevent them and how to mitigate them are still important subjects for nuclear engineering, and how to control the severe accidents after their occurrence, especially how to control FP release into the environment, has seldom been discussed in the water chemistry group recently. The paper is intended to address the issue of fewer activities for FP studies. FP sources are divided into two categories, short- and long-term FP sources. Short-term FP source can be evaluated based on the measured data obtained from monitoring posts (MPs), which give us clear evidence on the importance of radioactive iodine and cesium releases into the environment. It used to be considered that during primary containment vessel (PCV) venting, release of each element, e.g., iodine and cesium, was determined by the suppression pool scrubbing efficiency and most of the cesium would likely be removed in the pool due to its large scrubbing efficiency. But as a result of analyzing the MP data at early stage of the Fukushima Daiichi nuclear power plant (NPP) accident, it was confirmed that the releases of both elements were in proportion to their inventories in the reactors and their scrubbing efficiencies were almost the same. The scrubbing efficiency which increased with the pool water temperature became almost the same for iodine and cesium around the pool water boiling temperature. As a result of the mass balance analysis for FPs in the contaminated water accumulated in the Fukushima Daiichi plant site, it

  3. Reevaluation of HFIR source term: Supplement 2

    International Nuclear Information System (INIS)

    Thomas, W.E.

    1986-11-01

    The HFIR source term has been reevaluated to assess the impact of the increase in core lifetime from 15 to 24 days. Calculations were made to determine the nuclide activities of the iodines, noble gases, and other fission products. The results show that there is no significant change in off-site dose due to the increased fuel cycle for the release scenario postulated in ORNL-3573

  4. SOURCE TERM ESTIMATION BASED on PLANT STATUS and on GAMMA DOSE RATES Measured by an ON-line environmental Monitoring Network

    International Nuclear Information System (INIS)

    Stubna, M.; Bujan, A.; Duranova, T.

    1997-01-01

    A number of severe accident analyses for reactor unit with WWER-440 (213) has been performed in order to evaluate the source term and radiological consequences. As a tool for these analyses the WWER modified version of Source Term Code Package and Real Time Accident Release Consequences codes have been used. A set of emergency procedures - manuals for quick estimation of the source term and countermeasures introduction during early -pre-release phase of severe accident progression has been developed at Nuclear Power Plants Research Institute Trnava, Inc. These manuals are subdivided into three groups: 1.) evaluation of the barriers integrity, 2.) source term estimation and 3.) estimation of the distances for the countermeasures introduction. A methodology and computer module for interpretation of environmental data - source term assessment during post-release phase from on-line environmental network has been developed at Nuclear Power Plants Research Institute Trnava, Inc. The method is based on the conversion of measured dose rates to the source term,i.e. airborne radioactivity release rate, taking into account real meteorological data and location of the measure points. The bootstrap method for the estimation of the mean value of source term Q as integral value of the release and confidence interval of Q has been selected. The methodology of Q distribution into fission product groups according to code Real Time Accident Release Consequences needs is based on known plant status, i.e. on the results of pre calculated accident sequences. The paper describes the methodologies introduced above and the way of their application

  5. Source term estimation via monitoring data and its implementation to the RODOS system

    International Nuclear Information System (INIS)

    Bohunova, J.; Duranova, T.

    2000-01-01

    A methodology and computer code for interpretation of environmental data, i.e. source term assessment, from on-line environmental monitoring network was developed. The method is based on the conversion of measured dose rates to the source term, i.e. airborne radioactivity release rate, taking into account real meteorological data and location of the monitoring points. The bootstrap estimation methodology and bipivot method to estimate the source term from on-site gamma dose rate monitors is used. The mentioned methods provide an estimate of the mean value of the source term and a confidence interval for it. (author)

  6. Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, J.J.; Kamboj, S.; Gnanapragasam, E.; Yu, C. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2013-07-01

    RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short

  7. Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

    International Nuclear Information System (INIS)

    Cheng, J.J.; Kamboj, S.; Gnanapragasam, E.; Yu, C.

    2013-01-01

    RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short

  8. Application for approval of derived authorized limits for the release of the 190-C trenches and 105-C process water tunnels at the Hanford Site: Volume 2 - source term development

    International Nuclear Information System (INIS)

    Denham, D.H.; Winslow, S.L.; Moeller, M.P.; Kennedy, W.E. Jr.

    1997-03-01

    As part of environmental restoration activities at the Hanford Site, Bechtel Hanford, Inc. is conducting a series of evaluations to determine appropriate release conditions for specific facilities following the completion of decontamination and decommissioning projects. The release conditions, with respect to the residual volumetric radioactive contamination, are termed authorized limits. This report presents the summary of the supporting information and the final application for approval of derived authorized limits for the release of the 190-C trenches and the 105-C process water tunnels. This document contains two volumes; this volume (Vol. 2) contains the radiological characterization data, spreadsheet analyses, and radiological source terms

  9. Evaluation of the LMFBR cover gas source term and synthesis of the associated R and D

    International Nuclear Information System (INIS)

    Balard, F.; Carluec, B.

    1996-01-01

    At the end of the seventies and the beginning of the eighties, there appeared a pressing need of experimental results to assess the LMFBR's safety level. Because of the urgency, analytical studies were not systematically undertaken and maximum credible cover gas instantaneous source terms (radionuclides core release fraction) were got directly from crude out-of-pile experiment interpretations. Two types of studies and mock-ups were undertaken depending on the timescale of the phenomena: instantaneous source terms (corresponding to an unlikely energetic core disruptive accident CDA), and delayed ones (tens of minutes to some hours). The experiments performed in this frame are reviewed in this presentation: 1) instantaneous source term: - FAUST experiments: I, Cs, UO2 source terms (FzK, Germany), - FAST experiments : pool depth influence on non volatile source term (USA), - CARAVELLE experiments: nonvolatile source term in SPX1 geometry (CEA, France); 2) delayed source term: - NALA experiments: I, Cs, Sr, UO2 source term (FzK, Germany), - PAVE experiments: I source term (CEA, France), - NACOWA experiments: cover gas aerosols enrichment in I and Cs (FzK, Germany) - other French experiments in COPACABANA and GULLIVER facilities. The volatile fission products release is tightly bound to sodium evaporation and a large part of the fission products is dissolved in the liquid sodium aerosols present in the cover gas. Thus the knowledge of the amount of aerosol release to the cover gas is important for the evaluation of the source term. The maximum credible cover gas instantaneous source terms deduced from the experiments have led to conservative source terms to be taken into account in safety analysis. Nevertheless modelling attempts of the observed (in-pile or out-of-pile) physico-chemical phenomena have been undertaken for extrapolation to the reactor case. The main topics of this theoretical research are as follows: fission products evaporation in the cover gas (Fz

  10. Chernobyl source term, atmospheric dispersion, and dose estimation

    International Nuclear Information System (INIS)

    Gudiksen, P.H.; Harvey, T.F.; Lange, R.

    1988-02-01

    The Chernobyl source term available for long-range transport was estimated by integration of radiological measurements with atmospheric dispersion modeling, and by reactor core radionuclide inventory estimation in conjunction with WASH-1400 release fractions associated with specific chemical groups. These analyses indicated that essentially all of the noble gases, 80% of the radioiodines, 40% of the radiocesium, 10% of the tellurium, and about 1% or less of the more refractory elements were released. Atmospheric dispersion modeling of the radioactive cloud over the Northern Hemisphere revealed that the cloud became segmented during the first day, with the lower section heading toward Scandinavia and the uppper part heading in a southeasterly direction with subsequent transport across Asia to Japan, the North Pacific, and the west coast of North America. The inhalation doses due to direct cloud exposure were estimated to exceed 10 mGy near the Chernobyl area, to range between 0.1 and 0.001 mGy within most of Europe, and to be generally less than 0.00001 mGy within the US. The Chernobyl source term was several orders of magnitude greater than those associated with the Windscale and TMI reactor accidents, while the 137 Cs from the Chernobyl event is about 6% of that released by the US and USSR atmospheric nuclear weapon tests. 9 refs., 3 figs., 6 tabs

  11. LEAK: A source term generator for evaluating release rates from leaking vessels

    International Nuclear Information System (INIS)

    Clinton, J.H.

    1994-01-01

    An interactive computer code for estimating the rate of release of any one of several materials from a leaking tank or broken pipe leading from a tank is presented. It is generally assumed that the material in the tank is liquid. Materials included in the data base are acetonitrile, ammonia, carbon tetrachloride, chlorine, chlorine trifluoride, fluorine, hydrogen fluoride, nitric acid, nitrogen tetroxide, sodium hydroxide, sulfur hexafluoride, sulfuric acid, and uranium hexafluoride. Materials that exist only as liquid and/or vapor over expected ranges of temperature and pressure can easily be added to the data base file. The Fortran source code for LEAK and the data file are included with this report

  12. Source-term reevaluation for US commercial nuclear power reactors: a status report

    International Nuclear Information System (INIS)

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date

  13. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  14. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    Porfirio, Rogilson Nazare da Silva

    1996-01-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  15. Literature study of source term research for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sponton, L.L.; NiIsson, Lars

    2001-04-01

    A literature survey has been carried out in support of ongoing source term calculations with the MELCOR code of some severe accident scenarios for the Swedish Ringhals 2 pressurised water reactor (PWR). The research in the field of severe accidents in power reactors and the source term for subsequent release of radioisotopes was intensified after the Harrisburg accident and has produced a large amount of reports and papers. This survey was therefore limited to research concerning PWR type of reactors and with emphasis on papers related to MELCOR code development. A background is given, relating to some historic documents, and then more recent research after 1990 is reviewed. Of special interest is the ongoing PMbus-programme which is creating new and important results of benefit to the code development and validation of, among others, the MELCOR code. It is concluded that source term calculations involve simulation of many interacting complex physical phenomena, which result in large uncertainties The research has, however, over the years led to considerable improvements Thus has the uncertainty in source term predictions been reduced one to two orders of magnitude from the simpler codes in the early 1980-s to the more realistic codes of today, like MELCOR.

  16. Literature study of source term research for PWRs

    International Nuclear Information System (INIS)

    Sponton, L.L.; NiIsson, Lars

    2001-04-01

    A literature survey has been carried out in support of ongoing source term calculations with the MELCOR code of some severe accident scenarios for the Swedish Ringhals 2 pressurised water reactor (PWR). The research in the field of severe accidents in power reactors and the source term for subsequent release of radioisotopes was intensified after the Harrisburg accident and has produced a large amount of reports and papers. This survey was therefore limited to research concerning PWR type of reactors and with emphasis on papers related to MELCOR code development. A background is given, relating to some historic documents, and then more recent research after 1990 is reviewed. Of special interest is the ongoing PMbus-programme which is creating new and important results of benefit to the code development and validation of, among others, the MELCOR code. It is concluded that source term calculations involve simulation of many interacting complex physical phenomena, which result in large uncertainties The research has, however, over the years led to considerable improvements Thus has the uncertainty in source term predictions been reduced one to two orders of magnitude from the simpler codes in the early 1980-s to the more realistic codes of today, like MELCOR

  17. EPICS application source/release control

    International Nuclear Information System (INIS)

    Zieman, B.; Anderson, J.; Kraimer, M.

    1995-01-01

    This manual describes a set of Application Source/Release Control tools (appSR) that can be used to develop software for EPICS based control systems. The Application Source/Release Control System (appSR) has been unbundled from base EPICS and is now available as an EPICS extension. Due to this unbundling, two new directories must be added to a user's path (see section ''Environment'' on page 3 for more information) and a new command getapp must be issued after the getrel command to get a specific version of appSR (see section ''Creating The Initial Application System Area'' on page 7 for more information). It is now required that GNU make version 3.71 or later be used for makes instead of SUN make. Users should now type gmake instead of make

  18. Bayesian inverse modeling and source location of an unintended 131I release in Europe in the fall of 2011

    Science.gov (United States)

    Tichý, Ondřej; Šmídl, Václav; Hofman, Radek; Šindelářová, Kateřina; Hýža, Miroslav; Stohl, Andreas

    2017-10-01

    In the fall of 2011, iodine-131 (131I) was detected at several radionuclide monitoring stations in central Europe. After investigation, the International Atomic Energy Agency (IAEA) was informed by Hungarian authorities that 131I was released from the Institute of Isotopes Ltd. in Budapest, Hungary. It was reported that a total activity of 342 GBq of 131I was emitted between 8 September and 16 November 2011. In this study, we use the ambient concentration measurements of 131I to determine the location of the release as well as its magnitude and temporal variation. As the location of the release and an estimate of the source strength became eventually known, this accident represents a realistic test case for inversion models. For our source reconstruction, we use no prior knowledge. Instead, we estimate the source location and emission variation using only the available 131I measurements. Subsequently, we use the partial information about the source term available from the Hungarian authorities for validation of our results. For the source determination, we first perform backward runs of atmospheric transport models and obtain source-receptor sensitivity (SRS) matrices for each grid cell of our study domain. We use two dispersion models, FLEXPART and Hysplit, driven with meteorological analysis data from the global forecast system (GFS) and from European Centre for Medium-range Weather Forecasts (ECMWF) weather forecast models. Second, we use a recently developed inverse method, least-squares with adaptive prior covariance (LS-APC), to determine the 131I emissions and their temporal variation from the measurements and computed SRS matrices. For each grid cell of our simulation domain, we evaluate the probability that the release was generated in that cell using Bayesian model selection. The model selection procedure also provides information about the most suitable dispersion model for the source term reconstruction. Third, we select the most probable location of

  19. Bayesian inverse modeling and source location of an unintended 131I release in Europe in the fall of 2011

    Directory of Open Access Journals (Sweden)

    O. Tichý

    2017-10-01

    Full Text Available In the fall of 2011, iodine-131 (131I was detected at several radionuclide monitoring stations in central Europe. After investigation, the International Atomic Energy Agency (IAEA was informed by Hungarian authorities that 131I was released from the Institute of Isotopes Ltd. in Budapest, Hungary. It was reported that a total activity of 342 GBq of 131I was emitted between 8 September and 16 November 2011. In this study, we use the ambient concentration measurements of 131I to determine the location of the release as well as its magnitude and temporal variation. As the location of the release and an estimate of the source strength became eventually known, this accident represents a realistic test case for inversion models. For our source reconstruction, we use no prior knowledge. Instead, we estimate the source location and emission variation using only the available 131I measurements. Subsequently, we use the partial information about the source term available from the Hungarian authorities for validation of our results. For the source determination, we first perform backward runs of atmospheric transport models and obtain source-receptor sensitivity (SRS matrices for each grid cell of our study domain. We use two dispersion models, FLEXPART and Hysplit, driven with meteorological analysis data from the global forecast system (GFS and from European Centre for Medium-range Weather Forecasts (ECMWF weather forecast models. Second, we use a recently developed inverse method, least-squares with adaptive prior covariance (LS-APC, to determine the 131I emissions and their temporal variation from the measurements and computed SRS matrices. For each grid cell of our simulation domain, we evaluate the probability that the release was generated in that cell using Bayesian model selection. The model selection procedure also provides information about the most suitable dispersion model for the source term reconstruction. Third, we select the most

  20. Review of radionuclide source terms used for performance-assessment analyses

    International Nuclear Information System (INIS)

    Barnard, R.W.

    1993-06-01

    Two aspects of the radionuclide source terms used for total-system performance assessment (TSPA) analyses have been reviewed. First, a detailed radionuclide inventory (i.e., one in which the reactor type, decay, and burnup are specified) is compared with the standard source-term inventory used in prior analyses. The latter assumes a fixed ratio of pressurized-water reactor (PWR) to boiling-water reactor (BWR) spent fuel, at specific amounts of burnup and at 10-year decay. TSPA analyses have been used to compare the simplified source term with the detailed one. The TSPA-91 analyses did not show a significant difference between the source terms. Second, the radionuclides used in source terms for TSPA aqueous-transport analyses have been reviewed to select ones that are representative of the entire inventory. It is recommended that two actinide decay chains be included (the 4n+2 ''uranium'' and 4n+3 ''actinium'' decay series), since these include several radionuclides that have potentially important release and dose characteristics. In addition, several fission products are recommended for the same reason. The choice of radionuclides should be influenced by other parameter assumptions, such as the solubility and retardation of the radionuclides

  1. Source terms for airborne radioactivity arising from uranium mill wastes

    International Nuclear Information System (INIS)

    O'Riordan, M.C.; Downing, A.L.

    1978-01-01

    One of the problems in assessing the radiological impact of uranium milling is to determine the rates of release to the air of material from the various sources of radioactivity. Such source terms are required for modelling the transport of radioactive material in the atmosphere. Activity arises from various point and area sources in the mill itself and from the mill tailings. The state of the tailings changes in time from slurry to solid. A layer of water may be maintained over the solids during the life of the mine, and the tailings may be covered with inert material on abandonment. Releases may be both gaseous and particulate. This paper indicates ways in which radon emanation and the suspension of long-lived particulate activity might be quantified, and areas requiring further exploration are identified

  2. Using Bayesian Belief Network (BBN) modelling for rapid source term prediction. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Knochenhauer, M.; Swaling, V.H.; Dedda, F.D.; Hansson, F.; Sjoekvist, S.; Sunnegaerd, K. [Lloyd' s Register Consulting AB, Sundbyberg (Sweden)

    2013-10-15

    The project presented in this report deals with a number of complex issues related to the development of a tool for rapid source term prediction (RASTEP), based on a plant model represented as a Bayesian belief network (BBN) and a source term module which is used for assigning relevant source terms to BBN end states. Thus, RASTEP uses a BBN to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, composition, timing, and release path of released radio-nuclides). The output is a set of possible source terms with associated probabilities. One major issue has been associated with the integration of probabilistic and deterministic analyses are addressed, dealing with the challenge of making the source term determination flexible enough to give reliable and valid output throughout the accident scenario. The potential for connecting RASTEP to a fast running source term prediction code has been explored, as well as alternative ways of improving the deterministic connections of the tool. As part of the investigation, a comparison of two deterministic severe accident analysis codes has been performed. A second important task has been to develop a general method where experts' beliefs can be included in a systematic way when defining the conditional probability tables (CPTs) in the BBN. The proposed method includes expert judgement in a systematic way when defining the CPTs of a BBN. Using this iterative method results in a reliable BBN even though expert judgements, with their associated uncertainties, have been used. It also simplifies verification and validation of the considerable amounts of quantitative data included in a BBN. (Author)

  3. Using Bayesian Belief Network (BBN) modelling for rapid source term prediction. Final report

    International Nuclear Information System (INIS)

    Knochenhauer, M.; Swaling, V.H.; Dedda, F.D.; Hansson, F.; Sjoekvist, S.; Sunnegaerd, K.

    2013-10-01

    The project presented in this report deals with a number of complex issues related to the development of a tool for rapid source term prediction (RASTEP), based on a plant model represented as a Bayesian belief network (BBN) and a source term module which is used for assigning relevant source terms to BBN end states. Thus, RASTEP uses a BBN to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, composition, timing, and release path of released radio-nuclides). The output is a set of possible source terms with associated probabilities. One major issue has been associated with the integration of probabilistic and deterministic analyses are addressed, dealing with the challenge of making the source term determination flexible enough to give reliable and valid output throughout the accident scenario. The potential for connecting RASTEP to a fast running source term prediction code has been explored, as well as alternative ways of improving the deterministic connections of the tool. As part of the investigation, a comparison of two deterministic severe accident analysis codes has been performed. A second important task has been to develop a general method where experts' beliefs can be included in a systematic way when defining the conditional probability tables (CPTs) in the BBN. The proposed method includes expert judgement in a systematic way when defining the CPTs of a BBN. Using this iterative method results in a reliable BBN even though expert judgements, with their associated uncertainties, have been used. It also simplifies verification and validation of the considerable amounts of quantitative data included in a BBN. (Author)

  4. Effluent release limits, sources and control

    International Nuclear Information System (INIS)

    Swindell, G.E.

    1977-01-01

    Objectives of radiation protection in relation to releases. Environmental transfer models for radionuclides. Relationship between releases, environmental levels and doses to persons. Establishment of release limits: Limits based on critical population group concept critical pathway analysis and identification of critical group. Limits based on optimization of radiation protection individual dose limits, collective doses and dose commitments 1) differential cost benefit analysis 2) authorized and operational limits taking account of future exposures. Monitoring of releases to the environment: Objectives of effluent monitoring. Typical sources and composition of effluents; design and operation of monitoring programmes; recording and reporting of monitoring results; complementary environmental monitoring. (orig.) [de

  5. Source term assessment using inverse modeling of radiation dose measured with environmental radiation monitors located at different positions

    International Nuclear Information System (INIS)

    Srinivas, C.V.; Rakesh, P.T.; Baskaran, R.; Venkatraman, B.

    2018-01-01

    Source term is an important input for consequence analysis using Decision Support Systems (DSS) to project radiological impact in the event of nuclear emergencies. A source term model called 'ASTER' is incorporated in the Online Nuclear Emergency Response System (ONERS) operational at Kalpakkam site for decision making during nuclear emergencies. This computes release rates using inverse method by employing an atmospheric dispersion model and gamma dose rates measured by environmental radiation monitors (ERM) deployed around the nuclear plant. The estimates may depend on the distribution of ERMs around the release location. In this work, data from various gamma monitors located at different radii 0.75 km and 1.5 km is used to assess the accuracy in the source term estimation for stack releases of MAPS-PHWR at Kalpakkam

  6. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term - Trial Calculation

    International Nuclear Information System (INIS)

    Grabaskas, David

    2016-01-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  7. Protracted releases: inferring source terms and predicting dispersal

    International Nuclear Information System (INIS)

    Vamanu, D.V.

    1988-02-01

    Analytical solutions are given to the transport-diffusion equation for archetype, atmospheric protracted releases featuring fronts of initiation, culminations, and tails of extinction. The interplay of the fitting parameters ensures that the model accommodates a wide typology of events, nearing in the extremes the instantaneous puff of the Lagrangian models, and the continuous stack emission of the Gaussian models, respectively. (author)

  8. Projected Source Terms for Potential Sabotage Events Related to Spent Fuel Shipments

    International Nuclear Information System (INIS)

    Luna, R.E.; Neuhauser, K.S.; Vigil, M.G.

    1999-01-01

    Two major studies, one sponsored by the U.S. Department of Energy and the other by the U.S. Nuclear Regulatory Commission, were conducted in the late 1970s and early 1980s to provide information and source terms for an optimally successful act of sabotage on spent fuel casks typical of those available for use. This report applies the results of those studies and additional analysis to derive potential source terms for certain classes of sabotage events on spent fuel casks and spent fuel typical of those which could be shipped in the early decades of the 21st century. In addition to updating the cask and spent fuel characteristics used in the analysis, two release mechanisms not included in the earlier works were identified and evaluated. As would be expected, inclusion of these additional release mechanisms resulted in a somewhat higher total release from the postulated sabotage events. Although health effects from estimated releases were addressed in the earlier study conducted for U.S. Department of Energy, they have not been addressed in this report. The results from this report maybe used to estimate health effects

  9. Dose assessments for Greifswald and Cadarache with updated source terms from ITER NSSR-2

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1998-08-01

    The International Thermonuclear Experimental Reactor ITER is in its late engineering phase. One of the most important safety aspects - in particular for achieving public acceptance - is to assure that the releases of harzardous material are minimal during normal operation and for accidental events, even if very unlikely. To this purpose probabilistic dose assessments for accidental atmospheric releases of various ITER source terms which contain tritium and/or activation products were performed for the sites of Greifswald, Germany, and Cadarache, France. In addition, routine releases into the atmosphere and hydrosphere have been evaluated. No country specific rules were applied and the input parameters were adapted as far as possible to those used within former studies to achieve a better comparability with site independent dose assessments performed in the frame of ITER. The calculations were based on source terms which, for the first time, contain a combination of tritium and activation products. This allowed a better judgment of the contribution to the total dose of the individual fusion relevant materials. The results were compared to site independent dose limits defined in the frame of ITER. Annual doses from routine releases (CAT-I) are below 0.1 μSv for the aquatic scenarios and are close to 1 μSv for the atmospheric source terms. Source terms for two different categories of accidental releases, representing 'extremely unlikely events' (CAT-IV) and 'hypothetical sequences' (CAT-V), were investigated. In none of these cases, the release scenarios of category CAT-IV exceed the ITER limits. In addition, relevant characteristic quantities of the early dose distribution from the hypothetical scenarios of type CAT-V are still below 50 mSv or 100 mSv, values which are commonly used as lower reference values for evacuation in many potential home countries of ITER. These site specific assessments confirmed that the proposed release limits and thus the derived dose

  10. Source term and behavioural parameters for a postulated HIFAR loss-of-coolant accident

    International Nuclear Information System (INIS)

    May, F.G.

    1987-01-01

    The fraction of the fission product inventory which might be released into the atmosphere of the HIFAR reactor containment building (RCB) during a postulated loss-of-coolant accident (LOCA) has been evaluated as a function of time, for each classification of airborne radioactivity. This appraisal will be used as the source term for a computer program, which uses realistic attenuation of the fission product aerosol in a single compartment model with a defined leakrate to predict possible radioactive releases into the environment in a hypothetical bounding case reactor accident which is rather more severe in all major aspects than any single LOCA. Also given are the parameters governing the attenuation of the aerosol and vapours in the atmosphere of the RCB so that their behaviour may be accurately modelled. The source terms for several other types of accident involving the meltdown of fuel elements have also been considered but in less detail than the LOCA case. In some of the cases, the fission products are released directly to atmosphere, so there is no attenuation of the release by deposition within the RCB

  11. Field study of the long-term release of block copolymers from fouling-release coatings

    DEFF Research Database (Denmark)

    Noguer, Albert Camós; Olsen, A.; Hvilsted, Søren

    2017-01-01

    The addition of block copolymers (i.e. oils) is a common technique to enhance the biofouling-resistance properties of poly(dimethylsiloxane) (PDMS)-based fouling-release coatings. These copolymers diffuse from the bulk to the surface of the coating, thus modifying the properties of the surface an...... fouling-release coatings. Finally, the potential of long-term field-studies is discussed, as compared to short-term laboratory experiments usually performed within fouling-release coatings studies....

  12. Computer programs for developing source terms for a UF{sub 6} dispersion model to simulate postulated UF{sub 6} releases from buildings

    Energy Technology Data Exchange (ETDEWEB)

    Williams, W.R.

    1985-03-01

    Calculational methods and computer programs for the analysis of source terms for postulated releases of UF{sub 6} are presented. Required thermophysical properties of UF{sub 6}, HF, and H{sub 2}O are described in detail. UF{sub 6} reacts with moisture in the ambient environment to form HF and H{sub 2}O. The coexistence of HF and H{sub 2}O significantly alters their pure component properties, and HF vapor polymerizes. Transient compartment models for simulating UF{sub 6} releases inside gaseous diffusion plant feed and withdrawl buildings and cascade buildings are also described. The basic compartment model mass and energy balances are supported by simple heat transfer, ventilation system, and deposition models. A model that can simulate either a closed compartment or a steady-state ventilation system is also discussed. The transient compartment models provide input to an atmospheric dispersion model as output.

  13. COMPASS: A source term code for investigating capillary barrier performance

    International Nuclear Information System (INIS)

    Zhou, Wei; Apted, J.J.

    1996-01-01

    A computer code COMPASS based on compartment model approach is developed to calculate the near-field source term of the High-Level-Waste repository under unsaturated conditions. COMPASS is applied to evaluate the expected performance of Richard's (capillary) barriers as backfills to divert infiltrating groundwater at Yucca Mountain. Comparing the release rates of four typical nuclides with and without the Richard's barrier, it is shown that the Richard's barrier significantly decreases the peak release rates from the Engineered-Barrier-System (EBS) into the host rock

  14. Application of the source term code package to obtain a specific source term for the Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Souto, F.J.

    1991-06-01

    The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNLV). The following programme has been carried out to meet this objective: (a) implementation of the STCP, (b) acquisition of specific data for CNLV to execute the STCP, and (c) calculations of specific source terms for accident sequences at CNLV. The STCP has been implemented and validated on CDC 170/815 and CDC 180/860 main frames as well as on a Micro VAX 3800 system. In order to get a plant-specific source term, data on the CNLV including initial core inventory, burn-up, primary containment structures, and materials used for the calculations have been obtained. Because STCP does not explicitly model containment failure, dry well failure in the form of a catastrophic rupture has been assumed. One of the most significant sequences from the point of view of possible off-site risk is the loss of off-site power with failure of the diesel generators and simultaneous loss of high pressure core spray and reactor core isolation cooling systems. The probability for that event is approximately 4.5 x 10 -6 . This sequence has been analysed in detail and the release fractions of radioisotope groups are given in the full report. 18 refs, 4 figs, 3 tabs

  15. Long term safety assessment of geological waste disposal systems: issues on release scenarios

    International Nuclear Information System (INIS)

    Khan, S.A.; Qureshi, A.A.

    1995-01-01

    Geological insolation of high level nuclear waste is an attractive waste disposal concept. However, long term safety demonstration of this concept is a major challenge to the operators, regulators and the scientific community. Identification of the factors responsible for the release of radionuclides from geosphere to biosphere,is first step in this regard. Current understanding of the release scenarios indicates that faulting, ground after percolation, seismicity, volcanism and human intrusion are the dominating release factors for most of the candidate rock formations. The major source of uncertainties is the probability values of various release events due to random nature of catastrophic geological events and past poor historical records of the frequencies of such events. There is consensus among the experts that the waste release via human intrusion is the most unpredictable scenario at present state of the knowledge. (author)

  16. The SSI TOOLBOX Source Term Model SOSIM - Screening for important radionuclides and parameter sensitivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Avila Moreno, R.; Barrdahl, R.; Haegg, C.

    1995-05-01

    The main objective of the present study was to carry out a screening and a sensitivity analysis of the SSI TOOLBOX source term model SOSIM. This model is a part of the SSI TOOLBOX for radiological impact assessment of the Swedish disposal concept for high-level waste KBS-3. The outputs of interest for this purpose were: the total released fraction, the time of total release, the time and value of maximum release rate, the dose rates after direct releases of the biosphere. The source term equations were derived and simple equations and methods were proposed for calculation of these. A literature survey has been performed in order to determine a characteristic variation range and a nominal value for each model parameter. In order to reduce the model uncertainties the authors recommend a change in the initial boundary condition for solution of the diffusion equation for highly soluble nuclides. 13 refs.

  17. Low-level radioactive waste performance assessments: Source term modeling

    International Nuclear Information System (INIS)

    Icenhour, A.S.; Godbee, H.W.; Miller, L.F.

    1995-01-01

    Low-level radioactive wastes (LLW) generated by government and commercial operations need to be isolated from the environment for at least 300 to 500 yr. Most existing sites for the storage or disposal of LLW employ the shallow-land burial approach. However, the U.S. Department of Energy currently emphasizes the use of engineered systems (e.g., packaging, concrete and metal barriers, and water collection systems). Future commercial LLW disposal sites may include such systems to mitigate radionuclide transport through the biosphere. Performance assessments must be conducted for LUW disposal facilities. These studies include comprehensive evaluations of radionuclide migration from the waste package, through the vadose zone, and within the water table. Atmospheric transport mechanisms are also studied. Figure I illustrates the performance assessment process. Estimates of the release of radionuclides from the waste packages (i.e., source terms) are used for subsequent hydrogeologic calculations required by a performance assessment. Computer models are typically used to describe the complex interactions of water with LLW and to determine the transport of radionuclides. Several commonly used computer programs for evaluating source terms include GWSCREEN, BLT (Breach-Leach-Transport), DUST (Disposal Unit Source Term), BARRIER (Ref. 5), as well as SOURCE1 and SOURCE2 (which are used in this study). The SOURCE1 and SOURCE2 codes were prepared by Rogers and Associates Engineering Corporation for the Oak Ridge National Laboratory (ORNL). SOURCE1 is designed for tumulus-type facilities, and SOURCE2 is tailored for silo, well-in-silo, and trench-type disposal facilities. This paper focuses on the source term for ORNL disposal facilities, and it describes improved computational methods for determining radionuclide transport from waste packages

  18. Development of Reference Source Terms for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Kim, ByungIl; Lee, Chonghui; Lee, Dongsu; Ko, Heejin; Kang, Sangho [KEPCO Engineering and Construction Co. Inc., Yongin (Korea, Republic of)

    2014-05-15

    These source terms are developed for the typical U. S. NPP and do not reflect the design characteristics of EU-APR1400 (1,400 MWe PWR) which will be applied for the EUR certification in European countries. The process of developing the RST for EU-APR1400 is to undergo a similar process that NUREG-1465 had gone through when it came out with its proposed source terms. The purpose of this study is to develop the EU-APR1400 design-specific RST complied with the EUR. The Large LOCA is the reference equence used in the NUREG-1465 evaluation, whereas the EUAPR1400 risk-significant sequences are dominated by small LOCA and non-LOCA sequences. Moreover, when considering the EU-APR1400 has many design features to mitigate the consequences of severe accident phenomena, it is not surprising that the aspects of both release fractions and durations are distinctly different from NUREG-1465. This RST will be continuously updated to reflect to the design features of EU-APR1400, and then, be used as the reference for design purposes such as criteria satisfaction of radioactivity releases, equipment survivability, control room habitability for severe accident, and so on.

  19. DUSTMS-D: DISPOSAL UNIT SOURCE TERM - MULTIPLE SPECIES - DISTRIBUTED FAILURE DATA INPUT GUIDE.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.M.

    2006-01-01

    Performance assessment of a low-level waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). Many of these physical processes are influenced by the design of the disposal facility (e.g., how the engineered barriers control infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This has been done and the resulting models have been incorporated into the computer code DUST-MS (Disposal Unit Source Term-Multiple Species). The DUST-MS computer code is designed to model water flow, container degradation, release of contaminants from the wasteform to the contacting solution and transport through the subsurface media. Water flow through the facility over time is modeled using tabular input. Container degradation models include three types of failure rates: (a) instantaneous (all containers in a control volume fail at once), (b) uniformly distributed failures (containers fail at a linear rate between a specified starting and ending time), and (c) gaussian failure rates (containers fail at a rate determined by a mean failure time, standard deviation and gaussian distribution). Wasteform release models include four release mechanisms: (a) rinse with partitioning (inventory is released instantly upon container failure subject to equilibrium partitioning (sorption) with

  20. The Chernobyl reactor accident source term: Development of a consensus view

    International Nuclear Information System (INIS)

    Guntay, S.; Powers, D.A.; Devell, L.

    1997-01-01

    In August 1986, scientists from the former Soviet Union provided the nuclear safety community with an impressively detailed account of what was then known about the Chernobyl accident. This included assessments of the magnitudes, rates, and compositions of radionuclide releases during the ten days following initiation of the accident. A summary report based on the Soviet report, the oral presentations, and the discussions with scientists from various countries was issued by the International Atomic Energy Agency shortly thereafter. Ten years have elapsed since the reactor accident at Chernobyl. A great deal more data is now available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for about ten days. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved our understanding of the Chernobyl source term. Because of the special features of the reactor design and the pecularities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability of the safety analysis of other types of reactors

  1. Using Bayesian Belief Network (BBN) modelling for Rapid Source Term Prediction. RASTEP Phase 1

    International Nuclear Information System (INIS)

    Knochenhauer, M.; Swaling, V.H.; Alfheim, P.

    2012-09-01

    The project is connected to the development of RASTEP, a computerized source term prediction tool aimed at providing a basis for improving off-site emergency management. RASTEP uses Bayesian belief networks (BBN) to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, timing, and pathway of released radio-nuclides). The output is a set of possible source terms with associated probabilities. In the NKS project, a number of complex issues associated with the integration of probabilistic and deterministic analyses are addressed. This includes issues related to the method for estimating source terms, signal validation, and sensitivity analysis. One major task within Phase 1 of the project addressed the problem of how to make the source term module flexible enough to give reliable and valid output throughout the accident scenario. Of the alternatives evaluated, it is recommended that RASTEP is connected to a fast running source term prediction code, e.g., MARS, with a possibility of updating source terms based on real-time observations. (Author)

  2. Using Bayesian Belief Network (BBN) modelling for Rapid Source Term Prediction. RASTEP Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Knochenhauer, M.; Swaling, V.H.; Alfheim, P. [Scandpower AB, Sundbyberg (Sweden)

    2012-09-15

    The project is connected to the development of RASTEP, a computerized source term prediction tool aimed at providing a basis for improving off-site emergency management. RASTEP uses Bayesian belief networks (BBN) to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, timing, and pathway of released radio-nuclides). The output is a set of possible source terms with associated probabilities. In the NKS project, a number of complex issues associated with the integration of probabilistic and deterministic analyses are addressed. This includes issues related to the method for estimating source terms, signal validation, and sensitivity analysis. One major task within Phase 1 of the project addressed the problem of how to make the source term module flexible enough to give reliable and valid output throughout the accident scenario. Of the alternatives evaluated, it is recommended that RASTEP is connected to a fast running source term prediction code, e.g., MARS, with a possibility of updating source terms based on real-time observations. (Author)

  3. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Denman, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Clark, Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Denning, Richard S. [Consultant, Columbus, OH (United States)

    2016-10-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  4. Probabilistic source term predictions for use with decision support systems

    International Nuclear Information System (INIS)

    Grindon, E.; Kinniburgh, C.G.

    2003-01-01

    Full text: Decision Support Systems for use in off-site emergency management, following an incident at a Nuclear Power Plant (NPP) within Europe, are becoming accepted as a useful and appropriate tool to aid decision makers. An area which is not so well developed is the 'upstream' prediction of the source term released into the environment. Rapid prediction of this source term is crucial to the appropriate early management of a nuclear emergency. The initial source term prediction would today be typically based on simple tabulations taking little, or no, account of plant status. It is the interface between the inward looking plant control room team and the outward looking off-site emergency management team that needs to be addressed. This is not an easy proposition as these two distinct disciplines have little common basis from which to communicate their immediate findings and concerns. Within the Euratom Fifth Framework Programme (FP5), complementary approaches are being developed to the pre-release stage; each based on software tools to help bridge this gap. Traditionally source terms (or releases into the environment) provided for use with Decision Support Systems are estimated on a deterministic basis. These approaches use a single, deterministic assumption about plant status. The associated source term represents the 'best estimate' based an available information. No information is provided an the potential for uncertainty in the source term estimate. Using probabilistic methods the outcome is typically a number of possible plant states each with an associated source term and probability. These represent both the best estimate and the spread of the likely source term. However, this is a novel approach and the usefulness of such source term prediction tools is yet to be tested on a wide scale. The benefits of probabilistic source term estimation are presented here; using, as an example, the SPRINT tool developed within the FP5 STERPS project. System for the

  5. The EC CAST project (carbon-14 source term)

    International Nuclear Information System (INIS)

    Williams, S. J.

    2015-01-01

    Carbon-14 is a key radionuclide in the assessment of the safety of underground geological disposal facilities for radioactive wastes. It is possible for carbon-14 to be released from waste packages in a variety of chemical forms, both organic and inorganic, and as dissolved or gaseous species The EC CAST (CArbon-14 Source Term) project aims to develop understanding of the generation and release of carbon-14 from radioactive waste materials under conditions relevant to packaging and disposal. It focuses on the release of carbon-14 from irradiated metals (steels and zirconium alloys), from irradiated graphite and from spent ion-exchange resins. The CAST consortium brings together 33 partners. CAST commenced in October 2013 and this paper describes progress to March 2015. The main activities during this period were reviews of the current status of knowledge, the identification and acquisition of suitable samples and the design of experiments and analytical procedures. (authors)

  6. Fission product source terms and engineered safety features

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1984-01-01

    The author states that new, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents will soon be available. Although these methodologies will undoubtedly find widespread use in the development of accident response procedures, the author states that it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to strategies for the mitigation of fission product releases. Questions concerning the performance of existing engineered safety systems are reviewed

  7. Atucha-I source terms for sequences initiated by transients

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The present work is part of an expected source terms study in the Atucha I nuclear power plant during severe accidents. From the accident sequences with a significant probability to produce core damage, those initiated by operational transients have been identified as the most relevant. These sequences have some common characteristics, in the sense that all of them resume in the opening of the primary system safety valves, and leave this path open for the coolant loss. In the case these sequences continue as severe accidents, the same path will be used for the release of the radionuclides, from the core, through the primary system and to the containment. Later in the severe accident sequence, the failure of the pressure vessel will occur, and the corium will fall inside the reactor cavity, interacting with the concrete. During these processes, more radioactive products will be released inside the containment. In the present work the severe accident simulation initiated by a blackout is performed, from the point of view of the phenomenology of the behavior of the radioactive products, as they are transported in the piping, during the core-concrete interactions, and inside the containment buildings until it failure. The final result is the source term into the atmosphere. (author) [es

  8. Recent advances in the source term area within the SARNET European severe accident research network

    International Nuclear Information System (INIS)

    Herranz, L.E.; Haste, T.; Kärkelä, T.

    2015-01-01

    Highlights: • Main achievements of source term research in SARNET are given. • Emphasis on the radiologically important iodine and ruthenium fission products. • Conclusions on FP release, transport in the RCS and containment behaviour. • Significance of large-scale integral experiments to validate the analyses used. • A thorough list of the most recent references on source term research results. - Abstract: Source Term has been one of the main research areas addressed within the SARNET network during the 7th EC Framework Programme of EURATOM. The entire source term domain was split into three major areas: oxidising impact on source term, iodine chemistry in the reactor coolant system and containment and data and code assessment. The present paper synthesises the main technical outcome stemming from the SARNET FWP7 project in the area of source term and includes an extensive list of references in which deeper insights on specific issues may be found. Besides, based on the analysis of the current state of the art, an outlook of future source term research is outlined, where major changes in research environment are discussed (i.e., the end of the Phébus FP project; the end of the SARNET projects; and the launch of HORIZON 2020). Most probably research projects will be streamlined towards: release and transport under oxidising conditions, containment chemistry, existing and innovative filtered venting systems and others. These will be in addition to a number of projects that have been completed or are ongoing under different national and international frameworks, like VERDON, CHIP and EPICUR started under the International Source Term Programme (ISTP), the OECD/CSNI programmes BIP, BIP2, STEM, THAI and THAI2, and the French national programme MIRE. The experimental PASSAM project under the 7th EC Framework programme, focused on source term mitigation systems, is highlighted as a good example of a project addressing potential enhancement of safety systems

  9. Recent advances in the source term area within the SARNET European severe accident research network

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E., E-mail: luisen.herranz@ciemat.es [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 40, E-28040 Madrid (Spain); Haste, T. [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Kärkelä, T. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT Espoo (Finland)

    2015-07-15

    Highlights: • Main achievements of source term research in SARNET are given. • Emphasis on the radiologically important iodine and ruthenium fission products. • Conclusions on FP release, transport in the RCS and containment behaviour. • Significance of large-scale integral experiments to validate the analyses used. • A thorough list of the most recent references on source term research results. - Abstract: Source Term has been one of the main research areas addressed within the SARNET network during the 7th EC Framework Programme of EURATOM. The entire source term domain was split into three major areas: oxidising impact on source term, iodine chemistry in the reactor coolant system and containment and data and code assessment. The present paper synthesises the main technical outcome stemming from the SARNET FWP7 project in the area of source term and includes an extensive list of references in which deeper insights on specific issues may be found. Besides, based on the analysis of the current state of the art, an outlook of future source term research is outlined, where major changes in research environment are discussed (i.e., the end of the Phébus FP project; the end of the SARNET projects; and the launch of HORIZON 2020). Most probably research projects will be streamlined towards: release and transport under oxidising conditions, containment chemistry, existing and innovative filtered venting systems and others. These will be in addition to a number of projects that have been completed or are ongoing under different national and international frameworks, like VERDON, CHIP and EPICUR started under the International Source Term Programme (ISTP), the OECD/CSNI programmes BIP, BIP2, STEM, THAI and THAI2, and the French national programme MIRE. The experimental PASSAM project under the 7th EC Framework programme, focused on source term mitigation systems, is highlighted as a good example of a project addressing potential enhancement of safety systems

  10. Evaluation Plan on In-vessel Source Term in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials.

  11. Evaluation Plan on In-vessel Source Term in PGSFR

    International Nuclear Information System (INIS)

    Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong

    2016-01-01

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials

  12. Quantification of source-term profiles from near-field geochemical models

    International Nuclear Information System (INIS)

    McKinley, I.G.

    1985-01-01

    A geochemical model of the near-field is described which quantitatively treats the processes of engineered barrier degradation, buffering of aqueous chemistry by solid phases, nuclide solubilization and transport through the near-field and release to the far-field. The radionuclide source-terms derived from this model are compared with those from a simpler model used for repository safety analysis. 10 refs., 2 figs., 2 tabs

  13. Inverse modeling of the Chernobyl source term using atmospheric concentration and deposition measurements

    Science.gov (United States)

    Evangeliou, Nikolaos; Hamburger, Thomas; Cozic, Anne; Balkanski, Yves; Stohl, Andreas

    2017-07-01

    This paper describes the results of an inverse modeling study for the determination of the source term of the radionuclides 134Cs, 137Cs and 131I released after the Chernobyl accident. The accident occurred on 26 April 1986 in the Former Soviet Union and released about 1019 Bq of radioactive materials that were transported as far away as the USA and Japan. Thereafter, several attempts to assess the magnitude of the emissions were made that were based on the knowledge of the core inventory and the levels of the spent fuel. More recently, when modeling tools were further developed, inverse modeling techniques were applied to the Chernobyl case for source term quantification. However, because radioactivity is a sensitive topic for the public and attracts a lot of attention, high-quality measurements, which are essential for inverse modeling, were not made available except for a few sparse activity concentration measurements far from the source and far from the main direction of the radioactive fallout. For the first time, we apply Bayesian inversion of the Chernobyl source term using not only activity concentrations but also deposition measurements from the most recent public data set. These observations refer to a data rescue attempt that started more than 10 years ago, with a final goal to provide available measurements to anyone interested. In regards to our inverse modeling results, emissions of 134Cs were estimated to be 80 PBq or 30-50 % higher than what was previously published. From the released amount of 134Cs, about 70 PBq were deposited all over Europe. Similar to 134Cs, emissions of 137Cs were estimated as 86 PBq, on the same order as previously reported results. Finally, 131I emissions of 1365 PBq were found, which are about 10 % less than the prior total releases. The inversion pushes the injection heights of the three radionuclides to higher altitudes (up to about 3 km) than previously assumed (≈ 2.2 km) in order to better match both concentration

  14. Inverse modeling of the Chernobyl source term using atmospheric concentration and deposition measurements

    Directory of Open Access Journals (Sweden)

    N. Evangeliou

    2017-07-01

    Full Text Available This paper describes the results of an inverse modeling study for the determination of the source term of the radionuclides 134Cs, 137Cs and 131I released after the Chernobyl accident. The accident occurred on 26 April 1986 in the Former Soviet Union and released about 1019 Bq of radioactive materials that were transported as far away as the USA and Japan. Thereafter, several attempts to assess the magnitude of the emissions were made that were based on the knowledge of the core inventory and the levels of the spent fuel. More recently, when modeling tools were further developed, inverse modeling techniques were applied to the Chernobyl case for source term quantification. However, because radioactivity is a sensitive topic for the public and attracts a lot of attention, high-quality measurements, which are essential for inverse modeling, were not made available except for a few sparse activity concentration measurements far from the source and far from the main direction of the radioactive fallout. For the first time, we apply Bayesian inversion of the Chernobyl source term using not only activity concentrations but also deposition measurements from the most recent public data set. These observations refer to a data rescue attempt that started more than 10 years ago, with a final goal to provide available measurements to anyone interested. In regards to our inverse modeling results, emissions of 134Cs were estimated to be 80 PBq or 30–50 % higher than what was previously published. From the released amount of 134Cs, about 70 PBq were deposited all over Europe. Similar to 134Cs, emissions of 137Cs were estimated as 86 PBq, on the same order as previously reported results. Finally, 131I emissions of 1365 PBq were found, which are about 10 % less than the prior total releases. The inversion pushes the injection heights of the three radionuclides to higher altitudes (up to about 3 km than previously assumed (≈ 2.2 km in order

  15. 26 CFR 513.7 - Release of excess tax withheld at source.

    Science.gov (United States)

    2010-04-01

    ... source from life annuities and pensions, other than pensions paid by the Government of the United States... 26 Internal Revenue 19 2010-04-01 2010-04-01 false Release of excess tax withheld at source. 513.7...) REGULATIONS UNDER TAX CONVENTIONS IRELAND Withholding of Tax § 513.7 Release of excess tax withheld at source...

  16. Estimation of Source terms for Emergency Planning and Preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Chul Un; Chung, Bag Soon; Ahn, Jae Hyun; Yoon, Duk Ho; Jeong, Chul Young; Lim, Jong Dae [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Kang, Sun Gu; Suk, Ho; Park, Sung Kyu; Lim, Hac Kyu; Lee, Kwang Nam [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of)

    1997-12-31

    In this study the severe accident sequences for each plant of concern, which represent accident sequences with a high core damage frequency and significant accident consequences, were selected based on the results of probabilistic safety assessments and source term and time-histories of various safety parameters under severe accidents. Accidents progression analysis for each selected accident sequence was performed by MAAP code. It was determined that the measured values, dose rate and radioisotope concentration, could provide information to the operators on occurrence and timing of core damage, reactor vessel failure, and containment failure during severe accidents. Radioactive concentration in the containment atmosphere, which may be measured by PASS, was estimated. Radioisotope concentration in emergency planning, evaluation of source term behavior in the containment, estimation of core damage degree, analysis of severe accident phenomena, core damage timing, and the amount of radioisotope released to the environment. (author). 50 refs., 60 figs.

  17. Development, description and validation of a Tritium Environmental Release Model (TERM).

    Science.gov (United States)

    Jeffers, Rebecca S; Parker, Geoffrey T

    2014-01-01

    Tritium is a radioisotope of hydrogen that exists naturally in the environment and may also be released through anthropogenic activities. It bonds readily with hydrogen and oxygen atoms to form tritiated water, which then cycles through the hydrosphere. This paper seeks to model the migration of tritiated species throughout the environment - including atmospheric, river and coastal systems - more comprehensively and more consistently across release scenarios than is currently in the literature. A review of the features and underlying conceptual models of some existing tritium release models was conducted, and an underlying aggregated conceptual process model defined, which is presented. The new model, dubbed 'Tritium Environmental Release Model' (TERM), was then tested against multiple validation sets from literature, including experimental data and reference tests for tritium models. TERM has been shown to be capable of providing reasonable results which are broadly comparable with atmospheric HTO release models from the literature, spanning both continuous and discrete release conditions. TERM also performed well when compared with atmospheric data. TERM is believed to be a useful tool for examining discrete and continuous atmospheric releases or combinations thereof. TERM also includes further capabilities (e.g. river and coastal release scenarios) that may be applicable to certain scenarios that atmospheric models alone may not handle well. Copyright © 2013 Elsevier Ltd. All rights reserved.

  18. Perspectives on source terms based on early research and development

    International Nuclear Information System (INIS)

    Pressesky, A.J.

    1985-07-01

    This report presents an overview of the key documentation of the research and development programs relevant to the source term issue which were undertaken by the Atomic Energy Commission between 1950 and 1970. The source term is taken to be the amount, composition (physical and chemical), and timing of the projected release of radioactivity to the environment in the hypothetical event of a severe reactor accident in a light water reactor of the type currently being licensed, built and operated. The objective is to illuminate and provide perspectives on (a) the maturity of the technical data base and the analytical methodology, (b) the extent to which remaining conservatisms can be applied to compensate for uncertainties, (c) the purpose for which the technology and methodology will be used, and (d) the need to keep problems and uncertainties in proper perspective. Comments that can provide some context for the difficult programmatic choices to be made are included, and technical considerations that may be inadequately applied or neglected in some current source term calculations were studied. This review has not uncovered any significant technical considerations that have been omitted or are being inadequately treated in current source term analyses, except perhaps the contribution made to in-containment aerosols by coolant comminution upon escape at pressure from the reactor coolant system. 11 refs

  19. Design parameters and source terms: Volume 3, Source terms

    International Nuclear Information System (INIS)

    1987-10-01

    The Design Parameters and Source Terms Document was prepared in accordance with DOE request and to provide data for the environmental impact study to be performed in the future for the Deaf Smith County, Texas site for a nuclear waste repository in salt. This document updates a previous unpublished report by Stearns Catalytic Corporation (SCC), entitled ''Design Parameters and Source Terms for a Two-Phase Repository in Salt,'' 1985, to the level of the Site Characterization Plan - Conceptual Design Report. The previous unpublished SCC Study identifies the data needs for the Environmental Assessment effort for seven possible Salt Repository sites. 11 refs., 9 tabs

  20. Source term for the bounding assessment of the Canadian nuclear fuel waste disposal concept

    International Nuclear Information System (INIS)

    Flavelle, P.

    1996-02-01

    This is the second in a series to derive the bounds of the post-closure hazard of the Canadian nuclear fuel waste disposal concept, based on the premise that it is unnecessary to predict accurately the real hazard if the bounding hazard can be shown to be acceptable. In this report a reference used (Bruce A fuel, 865 GJ/kgU average burnup) is used to derive the source term for contaminant releases from the emplacement canisters. This requires development of a container failure function which defines the age of the fuel when the canister is perforated and flooded. The source term is expressed as the time-dependent fractional release rate from the used fuel or as the time-dependent contaminant concentrations in the canister porewater. It is derived as the superposition of an instant release, comprising the upper bound of the gap and grain boundary inventory in the used fuel, and the long-term dissolution of the used fuel matrix. Several dissolution models (stoichiometric dissolution/preferential leaching) under different conditions (matrix solubility limited/ unlimited; oxidizing/ reducing solubility limits; groundwater flow/ no flow) are evaluated and the one resulting in the highest release rate/ highest porewater concentration is adopted as the bounding case. Comparisons between the models are made on the basis of the potential ingestion hazard of the canister porewater, to account for differences in the hazard of different radionuclides. (author) 20 refs., 4 tabs., 9 figs

  1. Emergency preparedness source term development for the Office of Nuclear Material Safety and Safeguards-Licensed Facilities

    International Nuclear Information System (INIS)

    Sutter, S.L.; Mishima, J.; Ballinger, M.Y.; Lindsey, C.G.

    1984-08-01

    In order to establish requirements for emergency preparedness plans at facilities licensed by the Office of Nuclear Materials Safety and Safeguards, the Nuclear Regulatory Commission (NRC) needs to develop source terms (the amount of material made airborne) in accidents. These source terms are used to estimate the potential public doses from the events, which, in turn, will be used to judge whether emergency preparedness plans are needed for a particular type of facility. Pacific Northwest Laboratory is providing the NRC with source terms by developing several accident scenarios for eleven types of fuel cycle and by-product operations. Several scenarios are developed for each operation, leading to the identification of the maximum release considered for emergency preparedness planning (MREPP) scenario. The MREPP scenarios postulated were of three types: fire, tornado, and criticality. Fire was significant at oxide fuel fabrication, UF 6 production, radiopharmaceutical manufacturing, radiopharmacy, sealed source manufacturing, waste warehousing, and university research and development facilities. Tornadoes were MREPP events for uranium mills and plutonium contaminated facilities, and criticalities were significant at nonoxide fuel fabrication and nuclear research and development facilities. Techniques for adjusting the MREPP release to different facilities are also described

  2. Study of the Release Process of Open Source Software: Case Study

    OpenAIRE

    Eide, Tor Erik

    2007-01-01

    This report presents the results of a case study focusing on the release process of open source projects initiated with commercial motives. The purpose of the study is to gain an increased understanding of the release process, how a community can be attracted to the project, and how the interaction with the community evolves in commercial open source initiatives. Data has been gathered from four distinct sources to form the basis of this thesis. A thorough review of the open source literatu...

  3. Calculations in support of a potential definition of large release

    International Nuclear Information System (INIS)

    Hanson, A.L.; Davis, R.E.; Mubayi, V.

    1994-05-01

    The Nuclear Regulatory Commission has stated a hierarchy of safety goals with the qualitative safety goals as Level I of the hierarchy, backed up by the quantitative health objectives as Level II and the large release guideline as Level III. The large release guideline has been stated in qualitative terms as a magnitude of release of the core inventory whose frequency should not exceed 10 -6 per reactor year. However, the Commission did not provide a quantitative specification of a large release. This report describes various specifications of a large release and focuses, in particular, on an examination of releases which have a potential to lead to one prompt fatality in the mean. The basic information required to set up the calculations was derived from the simplified source terms which were obtained from approximations of the NUREG-1150 source terms. Since the calculation of consequences is affected by a large number of assumptions, a generic site with a (conservatively determined) population density and meteorology was specified. At this site, various emergency responses (including no response) were assumed based on information derived from earlier studies. For each of the emergency response assumptions, a set of calculations were performed with the simplified source terms; these included adjustments to the source terms, such as the timing of the release, the core inventory, and the release fractions of different radionuclides, to arrive at a result of one mean prompt fatality in each case. Each of the source terms, so defined, has the potential to be a candidate for a large release. The calculations show that there are many possible candidate source terms for a large release depending on the characteristics which are felt to be important

  4. Description of apparatus for determining radiological source terms of nuclear fuels

    International Nuclear Information System (INIS)

    Baldwin, D.L.; Woodley, R.E.; Holt, F.E.; Archer, D.V.; Steele, R.T.; Whitkop, P.G.

    1985-01-01

    New apparatus have been designed, built and are currently being employed to measure the release of volatile fission products from irradiated nuclear fuel. The system is capable of measuring radiological source terms, particularly for cesium-137, cesium-134, iodine-129 and krypton-85, in various atmospheres at temperatures up to 1200 0 C. The design allows a rapid transient heatup from ambient to full temperature, a hold at maximum temperature for a specified period, and rapid cooldown. Released fission products are measured as deposition on a platinum thermal gradient tube or in a filter/charcoal trap. Noble gases pass through to a multi-channel gamma analyzer. 1 ref., 4 figs

  5. Dose assessments for Greifswald and Cadarache with new source terms from ITER NSSR-1

    International Nuclear Information System (INIS)

    Raskob, W.; Forschungszentrum Karlsruhe GmbH Technik und Umwelt; Hasemann, I.

    1997-08-01

    Probabilistic dose assessments for accidental atmospheric releases of various ITER source terms which contain tritium and/or activation products were performed for the sites of Greifswald, Germany, and Cadarache, France. No country specific rules were applied and the input parameters were adapted as far as possible to those used within former ITER studies to achieve a better comparability with site independent dose assessments performed in the frame of ITER. The calculations were based on source terms which, at the first time, contain a combination of tritium and activation products. This allowed a better judgement of the contribution of the individual fusion relevant materials to the total dose. The results were compared to site independent dose limits defined in the frame of ITER. Source terms for two different categories, representing 'extremely unlikely events' (CAT-IV) and 'hypothetical sequences' (CAT-V), were investigated. In no cases, the release scenarios of category CAT-IV exceeded the ITER limits. In addition, early doses from the hypothetical scenarios of type CAT-V were still below 50 mSv or 100 mSv, values which are commonly used as lower reference values for evacuation in many potential home countries of ITER. Only the banning of food products was found to be a potential countermeasure which may affect larger areas. (orig.) [de

  6. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  7. Revised reactor accident source terms in the U.S. and implementation for light water reactors

    International Nuclear Information System (INIS)

    Soffer, L.; Lee, J.Y.

    1992-01-01

    Current NRC reactor accident source terms used for licensing are contained in Regulatory Guides 1.3 and 1.4 and specify that 100 % of the core inventory of noble gases and 25 % of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental (I 2 ) iodine. These assumptions have strongly affected present nuclear plant designs. Severe accident research results have confirmed that although the current source term is very substantial and has resulted in a very high level of plant capability, the present source term is no longer compatible with a realistic understanding of severe accidents. The NRC has issued a proposed revision of the reactor accident source terms as part of several regulatory activities to incorporate severe accident insights for future plants. A revision to 10 CFR 100 is also being proposed to specify site criteria directly and to eliminate source terms and doses for site evaluation. Reactor source terms will continue to be important in evaluating plant designs. Although intended primarily for future plants, existing and evolutionary power plants may voluntarily apply revised accident source term insights as well in licensing. The proposed revised accident source terms are presented in terms of fission product composition, magnitude, timing and iodine chemical form. Some implications for light water reactors are discussed. (author)

  8. Source term estimation based on in-situ gamma spectrometry using a high purity germanium detector

    International Nuclear Information System (INIS)

    Pauly, J.; Rojas-Palma, C.; Sohier, A.

    1997-06-01

    An alternative method to reconstruct the source term of a nuclear accident is proposed. The technique discussed here involves the use of in-situ gamma spectrometry. The validation of the applied methodology has been possible through the monitoring of routine releases of Ar-41 originating at a Belgian site from an air cooled graphite research reactor. This technique provides a quick nuclide specific decomposition of the source term and therefore will be have an enormous potential if implemented in nuclear emergency preparedness and radiological assessments of nuclear accidents during the early phase

  9. Source terms for analysis of accidents at a high level waste repository

    International Nuclear Information System (INIS)

    Mubayi, V.; Davis, R.E.; Youngblood, R.

    1989-01-01

    This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs

  10. Design parameters and source terms: Volume 2, Source terms: Revision 0

    International Nuclear Information System (INIS)

    1987-09-01

    The Design Parameters and Source Terms Document was prepared in accordance with DOE request and to provide data for the environmental impact study to be performed in the future for the Deaf Smith County, Texas site for a nuclear waste repository in salt. This document updates a previous unpublished report to the level of the Site Characterization Plan---Conceptual Design Report SCP-CDR. The previous study identifies the data needs for the Environmental Assessment effort for seven possible salt repository sites. Volume 2 contains tables of source terms

  11. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  12. Source-term model for the SYVAC3-NSURE performance assessment code

    International Nuclear Information System (INIS)

    Rowat, J.H.; Rattan, D.S.; Dolinar, G.M.

    1996-11-01

    Radionuclide contaminants in wastes emplaced in disposal facilities will not remain in those facilities indefinitely. Engineered barriers will eventually degrade, allowing radioactivity to escape from the vault. The radionuclide release rate from a low-level radioactive waste (LLRW) disposal facility, the source term, is a key component in the performance assessment of the disposal system. This report describes the source-term model that has been implemented in Ver. 1.03 of the SYVAC3-NSURE (Systems Variability Analysis Code generation 3-Near Surface Repository) code. NSURE is a performance assessment code that evaluates the impact of near-surface disposal of LLRW through the groundwater pathway. The source-term model described here was developed for the Intrusion Resistant Underground Structure (IRUS) disposal facility, which is a vault that is to be located in the unsaturated overburden at AECL's Chalk River Laboratories. The processes included in the vault model are roof and waste package performance, and diffusion, advection and sorption of radionuclides in the vault backfill. The model presented here was developed for the IRUS vault; however, it is applicable to other near-surface disposal facilities. (author). 40 refs., 6 figs

  13. Model description for calculating the source term of the Angra 1 environmental control system

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Amaral Neto, J.D.; Salles, M.R.

    1988-01-01

    This work presents the model used for evaluation of source term released from Angra 1 Nuclear Power Plant in case of an accident. After that, an application of the model for the case of a Fuel Assembly Drop Accident Inside the Fuel Handling Building during reactor refueling is presented. (author) [pt

  14. Design parameters and source terms: Volume 2, Source terms: Revision 0

    International Nuclear Information System (INIS)

    1987-10-01

    The Design Parameters and Source Terms Document was prepared in accordance with DOE request and to provide data for the environmental impact study to be performed in the future for the Deaf Smith County, Texas site for a nuclear waste repository in salt. This document updates a previous unpublished report by Stearns Catalytic Corporation (SCC), entitled ''Design Parameters and Source Terms for a Two-Phase Repository Salt,'' 1985, to the level of the Site Characterization Plan - Conceptual Design Report. The previous unpublished SCC Study identifies the data needs for the Environmental Assessment effort for seven possible Salt Repository sites. 2 tabs

  15. Preliminary investigation of processes that affect source term identification

    International Nuclear Information System (INIS)

    Wickliff, D.S.; Solomon, D.K.; Farrow, N.D.

    1991-09-01

    Solid Waste Storage Area (SWSA) 5 is known to be a significant source of contaminants, especially tritium ( 3 H), to the White Oak Creek (WOC) watershed. For example, Solomon et al. (1991) estimated the total 3 H discharge in Melton Branch (most of which originates in SWSA 5) for the 1988 water year to be 1210 Ci. A critical issue for making decisions concerning remedial actions at SWSA 5 is knowing whether the annual contaminant discharge is increasing or decreasing. Because (1) the magnitude of the annual contaminant discharge is highly correlated to the amount of annual precipitation (Solomon et al., 1991) and (2) a significant lag may exist between the time of peak contaminant release from primary sources (i.e., waste trenches) and the time of peak discharge into streams, short-term stream monitoring by itself is not sufficient for predicting future contaminant discharges. In this study we use 3 H to examine the link between contaminant release from primary waste sources and contaminant discharge into streams. By understanding and quantifying subsurface transport processes, realistic predictions of future contaminant discharge, along with an evaluation of the effectiveness of remedial action alternatives, will be possible. The objectives of this study are (1) to characterize the subsurface movement of contaminants (primarily 3 H) with an emphasis on the effects of matrix diffusion; (2) to determine the relative strength of primary vs secondary sources; and (3) to establish a methodology capable of determining whether the 3 H discharge from SWSA 5 to streams is increasing or decreasing

  16. Source term analysis for a criticality accident in metal production line glove boxes

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1991-06-01

    A recent development in criticality accident analysis is the deterministic calculations of the transport of fission products and actinides through the barriers of the physical facility. The knowledge of the redistribution of the materials inside the facility will help determine the reentry and clean-up procedures. The amount of radioactive materials released to the environment is the source term for dispersion calculations. We have used an integrated computer model to determine the release of fission products to the environment from a hypothetical criticality event in a glove box of the metal production line (MPL) at the Lawrence Livermore National Laboratory (LLNL)

  17. An improvement of estimation method of source term to the environment for interfacing system LOCA for typical PWR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Kim, Tae Woon; Ahn, Kwang Il [Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

  18. An artificial neural network approach to reconstruct the source term of a nuclear accident

    International Nuclear Information System (INIS)

    Giles, J.; Palma, C. R.; Weller, P.

    1997-01-01

    This work makes use of one of the main features of artificial neural networks, which is their ability to 'learn' from sets of known input and output data. Indeed, a trained artificial neural network can be used to make predictions on the input data when the output is known, and this feedback process enables one to reconstruct the source term from field observations. With this aim, an artificial neural networks has been trained, using the projections of a segmented plume atmospheric dispersion model at fixed points, simulating a set of gamma detectors located outside the perimeter of a nuclear facility. The resulting set of artificial neural networks was used to determine the release fraction and rate for each of the noble gases, iodines and particulate fission products that could originate from a nuclear accident. Model projections were made using a large data set consisting of effective release height, release fraction of noble gases, iodines and particulate fission products, atmospheric stability, wind speed and wind direction. The model computed nuclide-specific gamma dose rates. The locations of the detectors were chosen taking into account both building shine and wake effects, and varied in distance between 800 and 1200 m from the reactor.The inputs to the artificial neural networks consisted of the measurements from the detector array, atmospheric stability, wind speed and wind direction; the outputs comprised a set of release fractions and heights. Once trained, the artificial neural networks was used to reconstruct the source term from the detector responses for data sets not used in training. The preliminary results are encouraging and show that the noble gases and particulate fission product release fractions are well determined

  19. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    Science.gov (United States)

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2014-06-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of

  20. Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

    2001-08-01

    Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

  1. A study on the safety of spent fuel management. Radioactive source term modelling

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Kwan Sik; Lee, Hoo Keun; Park, Keun Il; Hwoang, Jung Ki; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1992-02-01

    The types and probabilities of events which may occur during the process of reception, transfer and storage of spent fuels in an away-from-reactor (AFR) spent fuel storage facility were analyzed in order to calculate the amount of radioactive material released to operation area and atmosphere, and the basic model for predicting the radioactive source-term under normal and abnormal operations were developed. Also, oxidation and dissolution of U0{sub 2} pellet was investigated to estimate the amount of radioactive materials released from spent fuel and the release characteristics of radionuclides from defected spent fuel rods was analyzed. Basic information using FIRAC code to analyze the ventilation system during fire accident was prepared and FIRIN was detached from FIRAC modified to simulate the compartment fire by personal computer. (Author).

  2. Using Reactive Transport Modeling to Evaluate the Source Term at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Y. Chen

    2001-12-19

    The conventional approach of source-term evaluation for performance assessment of nuclear waste repositories uses speciation-solubility modeling tools and assumes pure phases of radioelements control their solubility. This assumption may not reflect reality, as most radioelements (except for U) may not form their own pure phases. As a result, solubility limits predicted using the conventional approach are several orders of magnitude higher then the concentrations of radioelements measured in spent fuel dissolution experiments. This paper presents the author's attempt of using a non-conventional approach to evaluate source term of radionuclide release for Yucca Mountain. Based on the general reactive-transport code AREST-CT, a model for spent fuel dissolution and secondary phase precipitation has been constructed. The model accounts for both equilibrium and kinetic reactions. Its predictions have been compared against laboratory experiments and natural analogues. It is found that without calibrations, the simulated results match laboratory and field observations very well in many aspects. More important is the fact that no contradictions between them have been found. This provides confidence in the predictive power of the model. Based on the concept of Np incorporated into uranyl minerals, the model not only predicts a lower Np source-term than that given by conventional Np solubility models, but also produces results which are consistent with laboratory measurements and observations. Moreover, two hypotheses, whether Np enters tertiary uranyl minerals or not, have been tested by comparing model predictions against laboratory observations, the results favor the former. It is concluded that this non-conventional approach of source term evaluation not only eliminates over-conservatism in conventional solubility approach to some extent, but also gives a realistic representation of the system of interest, which is a prerequisite for truly understanding the long-term

  3. Using Reactive Transport Modeling to Evaluate the Source Term at Yucca Mountain

    International Nuclear Information System (INIS)

    Y. Chen

    2001-01-01

    The conventional approach of source-term evaluation for performance assessment of nuclear waste repositories uses speciation-solubility modeling tools and assumes pure phases of radioelements control their solubility. This assumption may not reflect reality, as most radioelements (except for U) may not form their own pure phases. As a result, solubility limits predicted using the conventional approach are several orders of magnitude higher then the concentrations of radioelements measured in spent fuel dissolution experiments. This paper presents the author's attempt of using a non-conventional approach to evaluate source term of radionuclide release for Yucca Mountain. Based on the general reactive-transport code AREST-CT, a model for spent fuel dissolution and secondary phase precipitation has been constructed. The model accounts for both equilibrium and kinetic reactions. Its predictions have been compared against laboratory experiments and natural analogues. It is found that without calibrations, the simulated results match laboratory and field observations very well in many aspects. More important is the fact that no contradictions between them have been found. This provides confidence in the predictive power of the model. Based on the concept of Np incorporated into uranyl minerals, the model not only predicts a lower Np source-term than that given by conventional Np solubility models, but also produces results which are consistent with laboratory measurements and observations. Moreover, two hypotheses, whether Np enters tertiary uranyl minerals or not, have been tested by comparing model predictions against laboratory observations, the results favor the former. It is concluded that this non-conventional approach of source term evaluation not only eliminates over-conservatism in conventional solubility approach to some extent, but also gives a realistic representation of the system of interest, which is a prerequisite for truly understanding the long-term

  4. Development of test apparatus for fission product release from overheated fuel element

    International Nuclear Information System (INIS)

    Takai, T.; Hirosawa, T.; Funabashi, H.; Miyahara, S.

    1996-01-01

    Evaluation of the source term released to environment under the accident conditions is important to the safety evaluation and design of reactor containment vessel. However, available data related to FBR source term are very limited, especially for the FPs release data from overheated FBR fuel. The present, source term evaluation of FBR is based on assumption from that of LWR. Though, this evaluation is very conservative. Evaluation large scale FBR source term using this method is result in extremely conservative and lead construction of large scale plant becomes doubtful from the viewpoints of cost and safety system. Though, it is necessary to evaluate source term from the realistic and rational scenario considering a characteristic of FBR. Preparation of FPs release experiment from irradiated fuel is going on to investigate the FPs release and transport and to develop the analysis code for in-vessel source term evaluation. Fabrication of this apparatus was started in 1992, and the installation was completed in 1994. This apparatus passed the facility inspection by Science and Technology Agency in March 1995. This apparatus consists of a high frequency induction furnace, thermal gradient tube (TGT), sintered metal filters, cold traps, gas-analyzer, γ-ray spectrometry system and so on. In the experiment, FPs release rate and behavior will be investigated using gamma-ray spectrogram and FP gas analysis. Physical and chemical composition of released FP would be investigated from FPs deposited profiles on TGT. Now, cold experiment using simulant FP materials are conducted. (author)

  5. The Analytical Repository Source-Term (AREST) model: Analysis of spent fuel as a nuclear waste form

    International Nuclear Information System (INIS)

    Apted, M.J.; Liebetrau, A.M.; Engel, D.W.

    1989-02-01

    The purpose of this report is to assess the performance of spent fuel as a final waste form. The release of radionuclides from spent nuclear fuel has been simulated for the three repository sites that were nominated for site characterization in accordance with the Nuclear Waste Policy Act of 1982. The simulation is based on waste package designs that were presented in the environmental assessments prepared for each site. Five distinct distributions for containment failure have been considered, and the release for nuclides from the UO 2 matrix, gap (including grain boundary), crud/surface layer, and cladding has been calculated with the Analytic Repository Source-Term (AREST) code. Separate scenarios involving incongruent and congruent release from the UO 2 matrix have also been examined using the AREST code. Congruent release is defined here as the condition in which the relative mass release rates of a given nuclide and uranium from the UO 2 matrix are equal to their mass ratios in the matrix. Incongruent release refers to release of a given nuclide from the UO 2 matrix controlled by its own solubility-limiting solid phase. Release of nuclides from other sources within the spent fuel (e.g., cladding, fuel/cladding gap) is evaluated separately from either incongruent or congruent matrix release. 51 refs., 200 figs., 9 tabs

  6. Development of computer-based function to estimate radioactive source term by coupling atmospheric model with monitoring data

    International Nuclear Information System (INIS)

    Akiko, Furuno; Hideyuki, Kitabata

    2003-01-01

    Full text: The importance of computer-based decision support systems for local and regional scale accidents has been recognized by many countries with the experiences of accidental atmospheric releases of radionuclides at Chernobyl in 1986 in the former Soviet Union. The recent increase of nuclear power plants in the Asian region also necessitates an emergency response system for Japan to predict the long-range atmospheric dispersion of radionuclides due to overseas accident. On the basis of these backgrounds, WSPEEDI (Worldwide version of System for Prediction of Environmental Emergency Dose Information) at Japan Atomic Energy Research Institute is developed to forecast long-range atmospheric dispersions of radionuclides during nuclear emergency. Although the source condition is critical parameter for accurate prediction, it is rarely that the condition can be acquired in the early stage of overseas accident. Thus, we have been developing a computer-based function to estimate radioactive source term, e.g. the release point, time and amount, as a part of WSPEEDI. This function consists of atmospheric transport simulations and statistical analysis for the prediction and monitoring of air dose rates. Atmospheric transport simulations are carried out for the matrix of possible release points in Eastern Asia and possible release times. The simulation results of air dose rates are compared with monitoring data and the best fitted release condition is defined as source term. This paper describes the source term estimation method and the application to Eastern Asia. The latest version of WSPEEDI accommodates following two models: an atmospheric meteorological model MM5 and a particle random walk model GEARN. MM5 is a non-hydrostatic meteorological model developed by the Pennsylvania State University and the National Center for Atmospheric Research (NCAR). MM5 physically calculates more than 40 meteorological parameters with high resolution in time and space based an

  7. Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

    2016-12-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  8. The use of nonlinear regression analysis for integrating pollutant concentration measurements with atmospheric dispersion modeling for source term estimation

    International Nuclear Information System (INIS)

    Edwards, L.L.; Freis, R.P.; Peters, L.G.; Gudiksen, P.H.; Pitovranov, S.E.

    1993-01-01

    The accuracy associated with assessing the environmental consequences of an accidental release of radioactivity is highly dependent on the knowledge of the source term characteristics, which are generally poorly known. The development of an automated numerical technique that integrates the radiological measurements with atmospheric dispersion modeling for more accurate source term estimation is reported. Often, this process of parameter estimation is performed by an emergency response assessor, who takes an intelligent first guess at the model parameters, then, comparing the model results with whatever measurements are available, makes an intuitive, informed next guess of the model parameters. This process may be repeated any number of times until the assessor feels that the model results are reasonable in terms of the measured observations. A new approach, based on a nonlinear least-squares regression scheme coupled with the existing Atmospheric Release Advisory Capability three-dimensional atmospheric dispersion models, is to supplement the assessor's intuition with automated mathematical methods that do not significantly increase the response time of the existing predictive models. The viability of the approach is evaluated by estimation of the known SF 6 tracer release rates associated with the Mesoscale Atmospheric Transport Studies tracer experiments conducted at the Savannah River Laboratory during 1983. These 19 experiments resulted in 14 successful, separate tracer releases with sampling of the tracer plumes along the cross-plume arc situated ∼30 km from the release site

  9. Evaluation on In-vessel Source Term in PGSFR (2015 Results)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. Since the TMI accident in 1979, extensive experimental and analytical information has been accumulated on the accident ST for LWRs. Such mechanistic models and computer codes as the MELCOR and MAAP have been developed. The results of extensive calculations and experiments have been used to formulate an alternative to the simple TID-14844 ST for regulatory purpose. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the peak burnup conditions. The nominal value of the radiological inventory is multiplied by a factor of 1.1 as an uncertainty margin to give the radiological inventory. ST in the release from the core to primary sodium is calculated by using the assumption of 4S methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using the assumption of 4S methodology.

  10. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  11. Source term experiments project (STEP): aerosol characterization system

    International Nuclear Information System (INIS)

    Schlenger, B.J.; Dunn, P.F.

    1985-01-01

    A series of four experiments has been conducted at Argonne National Laboratory's TREAT Reactor. These experiments, which are sponsored by an international consortium organized by the Electric Power Research Institute, are designed to investigate the source term, i.e., the type, quantity and timing of release of radioactive fission products from a light water reactor to the environment in the event of a severe accident in which the core is insufficiently cooled. The STEP tests have been designed to provide some of the necessary data regarding the magnitude and release rates of volatile fission products from degraded fuel pins, their physical and chemical characteristics, and aerosol formation and transport phenomena of those fission products that condense to form particles in the cooler regions of the reactor beyond the core. These are inpile experiments, whereby the test fuels are heated in a nuclear test reactor by neutron induced fission and subsequent cladding oxidation in steam environments that simulate as closely as practical predicted severe reactor accident conditions. The test sequences cover a range of pressure and fuel heatup rate, and include the effect of Ag/In/Cd control rod material. 1 ref., 8 figs., 1 tab

  12. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    International Nuclear Information System (INIS)

    Reistad, O.; Oelgaard, P.L.

    2006-04-01

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  13. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    Energy Technology Data Exchange (ETDEWEB)

    Reistad, O. [Norwegian Radiation Protection Authority (Norway); Oelgaard, P.L. [Risoe National Lab. (Denmark)

    2006-04-15

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  14. Evaluation of the Non-Transient Hydrologic Source Term from the CAMBRIC Underground Nuclear Test in Frenchman Flat, Nevada Test Site

    International Nuclear Information System (INIS)

    Tompson, A B; Maxwell, R M; Carle, S F; Zavarin, M; Pawloski, G A.; Shumaker, D E

    2005-01-01

    Hydrologic Source Term (HST) calculations completed in 1998 at the CAMBRIC underground nuclear test site were LLNL's first attempt to simulate a hydrologic source term at the NTS by linking groundwater flow and transport modeling with geochemical modeling (Tompson et al., 1999). Significant effort was applied to develop a framework that modeled in detail the flow regime and captured all appropriate chemical processes that occurred over time. However, portions of the calculations were simplified because of data limitations and a perceived need for generalization of the results. For example: (1) Transient effects arising from a 16 years of pumping at the site for a radionuclide migration study were not incorporated. (2) Radionuclide fluxes across the water table, as derived from infiltration from a ditch to which pumping effluent was discharged, were not addressed. (3) Hydrothermal effects arising from residual heat of the test were not considered. (4) Background data on the ambient groundwater flow direction were uncertain and not represented. (5) Unclassified information on the Radiologic Source Term (RST) inventory, as tabulated recently by Bowen et al. (2001), was unavailable; instead, only a limited set of derived data were available (see Tompson et al., 1999). (6) Only a small number of radionuclides and geochemical reactions were incorporated in the work. (7) Data and interpretation of the RNM-2S multiple well aquifer test (MWAT) were not available. As a result, the current Transient CAMBRIC Hydrologic Source Term project was initiated as part of a broader Phase 2 Frenchman Flat CAU flow and transport modeling effort. The source term will be calculated under two scenarios: (1) A more specific representation of the transient flow and radionuclide release behavior at the site, reflecting the influence of the background hydraulic gradient, residual test heat, pumping experiment, and ditch recharge, and taking into account improved data sources and modeling

  15. 10 CFR 50.67 - Accident source term.

    Science.gov (United States)

    2010-01-01

    ... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... to January 10, 1997, who seek to revise the current accident source term used in their design basis...

  16. Evaluation of source term parameters for spent fuel disposal in foreign countries. (1) Instant release fraction from spent fuel matrices and composition materials for fuel assemblies

    International Nuclear Information System (INIS)

    Nagata, Masanobu; Chikazawa, Takahiro; Kitamura, Akira; Tachi, Yukio; Akahori, Kuniaki

    2016-01-01

    Although spent nuclear fuel is planned to be disposed after reprocessing and vitrification of high-level radioactive waste (HLW), feasibility study on direct disposal of spent nuclear fuel (SF) has been started as one of the alternative disposal options to flexibly apply change of future energy situation in Japan. Radionuclide inventories and their release behavior after breaching spent fuel container should be assessed to confirm safety of the SF disposal. Especially, instant release fractions (IRFs), which are fractions of radionuclide released relatively faster than those released with congruent dissolution with SF and construction materials after breaching spent fuel container, may have an impact on safety assessment of the direct disposal of SF. However, detailed studies on evaluation / estimation of IRF have not been performed in Japan. Therefore, we investigated some foreign safety assessment reports on direct disposal of SF by focusing on IRF for the safety assessment of Japanese SF disposal system. As a result of comparison between the safety assessment reports in foreign countries, although some fundamental data have been referred to the reports in common, the final source term dataset was seen differences between countries in the result of taking into account the national circumstances (reactor types and burnups, etc.). We also found the difference of assignment of uncertainties among the investigated reports; a report selected pessimistic values and another report selected mean values and their deviations. It is expected that these findings are useful as fundamental information for the safety assessment of Japanese SF disposal system. (author)

  17. Integral migration and source term experiments on cement and bitumen waste forms

    International Nuclear Information System (INIS)

    Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

    1986-01-01

    This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate level wasteforms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of wasteform, at a controlled redox potential, and then through backfill and geological material packed in columns. The two wasteforms used here were cemented waste from the WAK plant at Karlsruhe, W. Germany and bitumenised intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium wire rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the wasteform. The release of all nuclides from the bitumenised waste was very low. (author)

  18. Integral migration and source-term experiments on cement and bitumen waste forms

    International Nuclear Information System (INIS)

    Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

    1986-01-01

    This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate-level waste-forms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of waste-form, at a controlled redox potential, and then through backfill and geological material packed in columns. The two waste forms used here were cemented waste from the WAK plant at Karlsruhe in the Federal Republic of Germany and bitumenized intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium were rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the waste-form. The released of all nuclides from the bitumenized waste was very low

  19. Evaluating vertical concentration profile of carbon source released from slow-releasing carbon source tablets and in situ biological nitrate denitrification activity

    Science.gov (United States)

    Yeum, Y.; HAN, K.; Yoon, J.; Lee, J. H.; Song, K.; Kang, J. H.; Park, C. W.; Kwon, S.; Kim, Y.

    2017-12-01

    Slow-releasing carbon source tablets were manufactured during the design of a small-scale in situ biological denitrification system to reduce high-strength nitrate (> 30 mg N/L) from a point source such as livestock complexes. Two types of slow-releasing tablets, precipitating tablet (PT, apparent density of 2.0 g/mL) and floating tablet (FT), were prepared to achieve a vertically even distribution of carbon source (CS) in a well and an aquifer. Hydroxypropyl methylcellulose (HPMC) was used to control the release rate, and microcrystalline cellulose pH 101 (MCC 101) was added as a binder. The #8 sand was used as a precipitation agent for the PTs, and the floating agents for the FTs were calcium carbonate and citric acid. FTs floated within 30 min. and remained in water because of the buoyance from carbon dioxide, which formed during the acid-base reaction between citric acid and calcium carbonate. The longevities of PTs with 300 mg of HPMC and FTs with 400 mg of HPMC were 25.4 days and 37.3 days, respectively. We assessed vertical CS profile in a continuous flowing physical aquifer model (release test, RT) and its efficiency on biological nitrate denitrification (denitrification test, DT). During the RT, PTs, FTs and a tracer (as 1 mg rhodamine B/L) were initially injected into a well of physical aquifer model (PAM). Concentrations of CS and the tracer were monitored along the streamline in the PAM to evaluate vertical profile of CS. During the DT, the same experiment was performed as RT, except continuous injection of solution containing 30 mg N/L into the PAM to evaluate biological denitrification activity. As a result of RT, temporal profiles of CS were similar at 3 different depths of monitoring wells. These results suggest that simultaneous addition of PT and FT be suitable for achieving a vertically even distribution of the CS in the injection well and an aquifer. In DT, similar profile of CS was detected in the injection well, and nitrate was biologically

  20. Calculation of source terms for NUREG-1150

    International Nuclear Information System (INIS)

    Breeding, R.J.; Williams, D.C.; Murfin, W.B.; Amos, C.N.; Helton, J.C.

    1987-10-01

    The source terms estimated for NUREG-1150 are generally based on the Source Term Code Package (STCP), but the actual source term calculations used in computing risk are performed by much smaller codes which are specific to each plant. This was done because the method of estimating the uncertainty in risk for NUREG-1150 requires hundreds of source term calculations for each accident sequence. This is clearly impossible with a large, detailed code like the STCP. The small plant-specific codes are based on simple algorithms and utilize adjustable parameters. The values of the parameters appearing in these codes are derived from the available STCP results. To determine the uncertainty in the estimation of the source terms, these parameters were varied as specified by an expert review group. This method was used to account for the uncertainties in the STCP results and the uncertainties in phenomena not considered by the STCP

  1. Unclassified Source Term and Radionuclide Data for Corrective Action Unit 97: Yucca Flat/Climax Mine, Nevada Test Site, Nevada, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Peter Martian

    2009-05-01

    This report documents the evaluation of the information and data available on the unclassified source term and radionuclide contamination for CAU 97: Yucca Flat/Climax Mine. The total residual inventory of radionuclides associated with one or more tests is known as the radiologic source term (RST). The RST is comprised of radionuclides in water, glass, or other phases or mineralogic forms. The hydrologic source term (HST) of an underground nuclear test is the portion of the total RST that is released into the groundwater over time following the test. In this report, the HST represents radionuclide release some time after the explosion and does not include the rapidly evolving mechanical, thermal, and chemical processes during the explosion. The CAU 97: Yucca Flat/Climax Mine has many more detonations and a wider variety of settings to consider compared to other CAUs. For instance, the source term analysis and evaluation performed for CAUs 101 and 102: Central and Western Pahute Mesa and CAU 98: Frenchman Flat did not consider vadose zone attenuation because many detonations were located near or below the water table. However, the large number of Yucca Flat/Climax Mine tests and the location of many tests above the water table warrant a more robust analysis of the unsaturated zone.

  2. Development of the methodology for application of revised source term to operating nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Kang, M.S.; Kang, P.; Kang, C.S.; Moon, J.H.

    2004-01-01

    Considering the current trend in applying the revised source term proposed by NUREG-1465 to the nuclear power plants in the U.S., it is expected that the revised source term will be applied to the Korean operating nuclear power plants in the near future, even though the exact time can not be estimated. To meet the future technical demands, it is necessary to prepare the technical system including the related regulatory requirements in advance. In this research, therefore, it is intended to develop the methodology to apply the revised source term to operating nuclear power plants in Korea. Several principles were established to develop the application methodologies. First, it is not necessary to modify the existing regulations about source term (i.e., any back-fitting to operating nuclear plants is not necessary). Second, if the pertinent margin of safety is guaranteed, the revised source term suggested by NUREG-1465 may be useful to full application. Finally, a part of revised source term could be selected to application based on the technical feasibility. As the results of this research, several methodologies to apply the revised source term to the Korean operating nuclear power plants have been developed, which include: 1) the selective (or limited) application to use only some of all the characteristics of the revised source term, such as release timing of fission products and chemical form of radio-iodine and 2) the full application to use all the characteristics of the revised source term. The developed methodologies are actually applied to Ulchin 9 and 4 units and their application feasibilities are reviewed. The results of this research are used as either a manual in establishing the plan and the procedure for applying the revised source term to the domestic nuclear plant from the utility's viewpoint; or a technical basis of revising the related regulations from the regulatory body's viewpoint. The application of revised source term to operating nuclear

  3. Subsurface Shielding Source Term Specification Calculation

    International Nuclear Information System (INIS)

    S.Su

    2001-01-01

    The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations

  4. Environmental radiation safety: source term modification by soil aerosols. Interim report

    International Nuclear Information System (INIS)

    Moss, O.R.; Allen, M.D.; Rossignol, E.J.; Cannon, W.C.

    1980-08-01

    The goal of this project is to provide information useful in estimating hazards related to the use of a pure refractory oxide of 238 Pu as a power source in some of the space vehicles to be launched during the next few years. Although the sources are designed and built to withstand re-entry into the earth's atmosphere, and to impact with the earth's surface without releasing any plutonium, the possibility that such an event might produce aerosols composed of soil and 238 PuO 2 cannot be absolutely excluded. This report presents the results of our most recent efforts to measure the degree to which the plutonium aerosol source term might be modified in a terrestrial environment. The five experiments described represent our best effort to use the original experimental design to study the change in the size distribution and concentration of a 238 PuO 2 aerosol due to coagulation with an aerosol of clay or sandy loam soil

  5. Determination of the long-term release of metal(loid)s from construction materials using DGTs.

    Science.gov (United States)

    Schmukat, A; Duester, L; Ecker, D; Heininger, P; Ternes, T A

    2013-09-15

    Long-term leaching experiments are crucial to estimate the potential release of dangerous substances from construction materials. The application of Diffuse Gradients in Thin film (DGT) in static-batch experiments was tested to study the long-term release of metal(loid)s from construction materials for hydraulic engineering, for half a year. Long-term release experiments are essential to improve calculations of the life-time release for this materials. DGTs in batch experiments were found to be a space and labour efficient application, which enabled (i) to study, in a non-invasive manner, the total release of nine metal(loid)s for half a year, (ii) to differentiate between release mechanisms and (iii) to study mechanisms which were contrary to the release or caused experimental artefacts in the batch experiments. For copper slag (test material) it was found that eight metal(loid)s were released over the whole time period of 184 d. Cu, Ni and Pb were found to be released, predominantly caused by (the) weathering of sulphide minerals. Only for Zn a surface depletion mechanism was identified. The results from the long-term batch experiments deliver new information on the release of metal(loid)s during the life cycle of construction materials with regard to river basin management objectives. Copyright © 2013 Elsevier B.V. All rights reserved.

  6. A source term estimation method for a nuclear accident using atmospheric dispersion models

    DEFF Research Database (Denmark)

    Kim, Minsik; Ohba, Ryohji; Oura, Masamichi

    2015-01-01

    The objective of this study is to develop an operational source term estimation (STE) method applicable for a nuclear accident like the incident that occurred at the Fukushima Dai-ichi nuclear power station in 2011. The new STE method presented here is based on data from atmospheric dispersion...... models and short-range observational data around the nuclear power plants.The accuracy of this method is validated with data from a wind tunnel study that involved a tracer gas release from a scaled model experiment at Tokai Daini nuclear power station in Japan. We then use the methodology developed...... and validated through the effort described in this manuscript to estimate the release rate of radioactive material from the Fukushima Dai-ichi nuclear power station....

  7. Tracking the release of IPCC AR5 on Twitter: Users, comments, and sources following the release of the Working Group I Summary for Policymakers.

    Science.gov (United States)

    Newman, Todd P

    2017-10-01

    Using the immediate release of the Working Group 1 Summary for Policymakers of the Intergovernmental Panel on Climate Change Fifth Assessment Report as a case study, this article seeks to describe what type of actors were most active during the summary release, the substance of the most propagated tweets during the summary release, and the media sources that attracted the most attention during the summary release. The results from the study suggest that non-elite actors, such as individual bloggers and concerned citizens, accounted for the majority of the most propagated tweets in the sample. This study also finds that the majority of the most propagated tweets in the sample focused on public understanding of the report. Finally, while mainstream media sources were the most frequently discussed media sources, a number of new media and science news and information sources compete for audience attention.

  8. Predicted radionuclide release from marine reactors dumped in the Kara Sea. Report of the source term working group of the international arctic seas assessment project (IASAP)

    International Nuclear Information System (INIS)

    1997-04-01

    The present report summarizes the work carried out by the Source Term Working Group of IASAP during 1994-1996. The report is based on the studies concerning the initial and current radionuclide inventories, operational history and construction of the reactors carried out by Y. Sivintsev of the Russian Research Center ''Kurchatov Institute'', Moscow and E. Yefimov of the Institute of Physics and Power Engineering, Obninsk, Russian Federation. The working group convened five times and evaluated the results of the studies and developed models for prediction of potential releases to the environment. The calculations were carried out at the Royal Naval College, Greenwich, UK, by N. Lynn, J. Warden and S. Timms and at the Lawrence Livermore National Laboratory, California, USA, by M. Mount. 31 refs, 36 figs, 18 tabs

  9. Calculation of the source term for a S1B-sequence at a VVER-1000 type reactor. Part 1

    International Nuclear Information System (INIS)

    Sdouz, G.

    1990-10-01

    The behaviour of the source term in a VVER-1000 type reactor is calculated using the 'Source Term Code Package' (STCP). The input data are based on the russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA in the hot leg followed by loss offsite and onsite electric power (S 1 B-sequence). According to the course of the calculation the results are presented and analyzed for each program. Except for the noble gases all release fractions are lower than 10 -4 . 18 refs., 10 tabs. (Author)

  10. 78 FR 48156 - Update to An Inventory of Sources and Environmental Releases of Dioxin-Like Compounds in the...

    Science.gov (United States)

    2013-08-07

    ... Inventory of Sources and Environmental Releases of Dioxin-Like Compounds in the United States for the Years... document entitled, Update to An Inventory of Sources and Environmental Releases of Dioxin-Like Compounds in... the report: An Inventory of Sources and Environmental Releases of Dioxin-Like Compounds in the United...

  11. The role of a detailed aqueous phase source release model in the LANL area G performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Vold, E.L.; Shuman, R.; Hollis, D.K. [Los Alamos National Lab., NM (United States)] [and others

    1995-12-31

    A preliminary draft of the Performance Assessment for the Los Alamos National Laboratory (LANL) low-level radioactive waste disposal facility at Area G is currently being completed as required by Department of Energy orders. A detailed review of the inventory data base records and the existing models for source release led to the development of a new modeling capability to describe the liquid phase transport from the waste package volumes. Nuclide quantities are sorted down to four waste package release categories for modeling: rapid release, soil, concrete/sludge, and corrosion. Geochemistry for the waste packages was evaluated in terms of the equilibrium coefficients, Kds, and elemental solubility limits, Csl, interpolated from the literature. Percolation calculations for the base case closure cover show a highly skewed distribution with an average of 4 mm/yr percolation from the disposal unit bottom. The waste release model is based on a compartment representation of the package efflux, and depends on package size, percolation rate or Darcy flux, retardation coefficient, and moisture content.

  12. Lysimeter data as input to performance assessment source term codes

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Sullivan, T.

    1992-01-01

    The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II c prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first seven years of sampling are presented and discussed. Application of lysimeter data to be used in performance assessment source term models is presented. Initial results from use of data in two models are discussed

  13. Impairment of Release Site Clearance within the Active Zone by Reduced SCAMP5 Expression Causes Short-Term Depression of Synaptic Release

    Directory of Open Access Journals (Sweden)

    Daehun Park

    2018-03-01

    Full Text Available Summary: Despite being a highly enriched synaptic vesicle (SV protein and a candidate gene for autism, the physiological function of SCAMP5 remains mostly enigmatic. Here, using optical imaging and electrophysiological experiments, we demonstrate that SCAMP5 plays a critical role in release site clearance at the active zone. Truncation analysis revealed that the 2/3 loop domain of SCAMP5 directly interacts with adaptor protein 2, and this interaction is critical for its role in release site clearance. Knockdown (KD of SCAMP5 exhibited pronounced synaptic depression accompanied by a slower recovery of the SV pool. Moreover, it induced a strong frequency-dependent short-term depression of synaptic release, even under the condition of sufficient release-ready SVs. Super-resolution microscopy further proved the defects in SV protein clearance induced by KD. Thus, reduced expression of SCAMP5 may impair the efficiency of SV clearance at the active zone, and this might relate to the synaptic dysfunction observed in autism. : Park et al. show that SCAMP5 plays an important role in release site clearance during intense neuronal activity. Loss of SCAMP5 results in a traffic jam at release sites, causing aberrant short-term synaptic depression that might be associated with the synaptic dysfunction observed in autism. Keywords: secretory carrier membrane protein, SCAMP5, autism spectrum disorder, adaptor protein 2, release site clearance, presynaptic active zone, short-term depression, endocytosis, super-resolution microscopy

  14. VizieR Online Data Catalog: The Chandra Source Catalog, Release 1.1 (Evans+ 2012)

    Science.gov (United States)

    Evans, I. N.; Primini, F. A.; Glotfelty, C. S.; Anderson, C. S.; Bonaventura, N. R.; Chen, J. C.; Davis, J. E.; Doe, S. M.; Evans, J. D.; Fabbiano, G.; Galle, E. C.; Gibbs, D. G.; Grier, J. D.; Hain, R. M.; Hall, D. M.; Harbo, P. N.; He, X.; Houck, J. C.; Karovska, M.; Kashyap, V. L.; Lauer, J.; McCollough, M. L.; McDowell, J. C.; Miller, J. B.; Mitschang, A. W.; Morgan, D. L.; Mossman, A. E.; Nichols, J. S.; Nowak, M. A.; Plummer, D. A.; Refsdal, B. L.; Rots, A. H.; Siemiginowska, A.; Sundheim, B. A.; Tibbetts, M. S.; van Stone, D. W.; Winkelman, S. L.; Zografou, P.

    2014-01-01

    This version of the catalog is release 1.1. It includes the information contained in release 1.0.1, plus point and compact source data extracted from HRC imaging observations, and catch-up ACIS observations released publicly prior to the end of 2009. (1 data file).

  15. Aerosol behavior and light water reactor source terms

    International Nuclear Information System (INIS)

    Abbey, F.; Schikarski, W.O.

    1988-01-01

    The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

  16. Evaluation of PCB sources and releases for identifying priorities to reduce PCBs in Washington State (USA).

    Science.gov (United States)

    Davies, Holly; Delistraty, Damon

    2016-02-01

    Polychlorinated biphenyls (PCBs) are ubiquitously distributed in the environment and produce multiple adverse effects in humans and wildlife. As a result, the purpose of our study was to characterize PCB sources in anthropogenic materials and releases to the environment in Washington State (USA) in order to formulate recommendations to reduce PCB exposures. Methods included review of relevant publications (e.g., open literature, industry studies and reports, federal and state government databases), scaling of PCB sources from national or county estimates to state estimates, and communication with industry associations and private and public utilities. Recognizing high associated uncertainty due to incomplete data, we strived to provide central tendency estimates for PCB sources. In terms of mass (high to low), PCB sources include lamp ballasts, caulk, small capacitors, large capacitors, and transformers. For perspective, these sources (200,000-500,000 kg) overwhelm PCBs estimated to reside in the Puget Sound ecosystem (1500 kg). Annual releases of PCBs to the environment (high to low) are attributed to lamp ballasts (400-1500 kg), inadvertent generation by industrial processes (900 kg), caulk (160 kg), small capacitors (3-150 kg), large capacitors (10-80 kg), pigments and dyes (0.02-31 kg), and transformers (PCB distribution and decrease exposures include assessment of PCBs in buildings (e.g., schools) and replacement of these materials, development of Best Management Practices (BMPs) to contain PCBs, reduction of inadvertent generation of PCBs in consumer products, expansion of environmental monitoring and public education, and research to identify specific PCB congener profiles in human tissues.

  17. Source Release Modeling for the Idaho National Engineering and Environmental Laboratory's Subsurface Disposal Area

    International Nuclear Information System (INIS)

    Becker, B.H.

    2002-01-01

    A source release model was developed to determine the release of contaminants into the shallow subsurface, as part of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) evaluation at the Idaho National Engineering and Environmental Laboratory's (INEEL) Subsurface Disposal Area (SDA). The output of the source release model is used as input to the subsurface transport and biotic uptake models. The model allowed separating the waste into areas that match the actual disposal units. This allows quantitative evaluation of the relative contribution to the total risk and allows evaluation of selective remediation of the disposal units within the SDA

  18. Estimate of radionuclide release characteristics into containment under severe accident conditions

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented

  19. An initial assessment of the Chernobyl-4 reactor accident release source

    International Nuclear Information System (INIS)

    Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

    1986-07-01

    The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

  20. Mechanistic facility safety and source term analysis

    International Nuclear Information System (INIS)

    PLYS, M.G.

    1999-01-01

    A PC-based computer program was created for facility safety and source term analysis at Hanford The program has been successfully applied to mechanistic prediction of source terms from chemical reactions in underground storage tanks, hydrogen combustion in double contained receiver tanks, and proccss evaluation including the potential for runaway reactions in spent nuclear fuel processing. Model features include user-defined facility room, flow path geometry, and heat conductors, user-defined non-ideal vapor and aerosol species, pressure- and density-driven gas flows, aerosol transport and deposition, and structure to accommodate facility-specific source terms. Example applications are presented here

  1. Some practical implications of source term reassessment

    International Nuclear Information System (INIS)

    1988-03-01

    This report provides a brief summary of the current knowledge of severe accident source terms and suggests how this knowledge might be applied to a number of specific aspects of reactor safety. In preparing the report, consideration has been restricted to source term issues relating to light water reactors (LWRs). Consideration has also generally been restricted to the consequences of hypothetical severe accidents rather than their probability of occurrence, although it is recognized that, in the practical application of source term research, it is necessary to take account of probability as well as consequences. The specific areas identified were as follows: Exploration of the new insights that are available into the management of severe accidents; Investigating the impact of source term research on emergency planning and response; Assessing the possibilities which exist in present reactor designs for preventing or mitigating the consequences of severe accidents and how these might be used effectively; Exploring the need for backfitting and assessing the implications of source term research for future designs; and Improving the quantification of the radiological consequences of hypothetical severe accidents for probabilistic safety assessments (PSAs) and informing the public about the realistic risks associated with nuclear power plants. 7 refs

  2. Consequences of severe radioactive releases to Nordic Marine environment

    DEFF Research Database (Denmark)

    Iosjpe, M.; Isaksson, M.; Joensen, H.P.

    - or minor – radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MWth reactor size were used. Based on source term analyses, the chosen release fractions in the study were: iodine 20% (of the total core inventory), caesium 10%, tellurium 10%, strontium 0...

  3. Revised accident source terms and control room habitability

    International Nuclear Information System (INIS)

    Lahti, G.P.; Hubner, R.S.; Johnson, W.J.; Schwartz, B.C.

    1993-01-01

    In April 1992, the NRC staff presented to the Commissioners the draft NUREG open-quotes Revised Accident Source Terms for Light-Water Nuclear Power Plants.close quotes This document is the culmination of more than ten years of NRC-sponsored research and represents the first change in the NRC's position on source terms since TID-14844 was issued in 1962. The purpose of this paper is to investigate the impact of the revised source terms on the current approach to analyzing control room habitability as required by 10 CFR 50. Sample calculations are presented that identify aspects of the model requiring clarification before the implementation of the revised source terms. 6 refs., 4 tabs

  4. Planck early results. XIII. Statistical properties of extragalactic radio sources in the Planck Early Release Compact Source Catalogue

    DEFF Research Database (Denmark)

    Lähteenmäki, A.; Poutanen, T.; Natoli, P.

    2011-01-01

    The data reported in Planck's Early Release Compact Source Catalogue (ERCSC) are exploited to measure the number counts (dN/dS) of extragalactic radio sources at 30, 44, 70, 100, 143 and 217 GHz. Due to the full-sky nature of the catalogue, this measurement extends to the rarest and brightest sou...

  5. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  6. Dilaton field released under collision of dilatonic black holes with Gauss-Bonnet term

    Energy Technology Data Exchange (ETDEWEB)

    Gwak, Bogeun [Sejong University, Department of Physics and Astronomy, Seoul (Korea, Republic of); Ro, Daeho [POSTECH, Asia Pacific Center for Theoretical Physics, Pohang, Gyeongbuk (Korea, Republic of)

    2017-08-15

    We investigate the upper limit of the gravitational radiation released upon the collision of two dilatonic black holes by analyzing the Gauss-Bonnet term. Dilatonic black holes have a dilaton hair coupled with this term. Using the laws of thermodynamics, the upper limit of the radiation is obtained, which reflected the effects of the dilaton hair. The amount of radiation released is greater than that emitted by a Schwarzschild black hole due to the contribution from the dilaton hair. In the collision, most of the dilaton hair can be released through radiation, where the energy radiated by the dilaton hair is maximized when the horizon of one black hole is minimized for a fixed second black hole. (orig.)

  7. Understanding long-term silver release from surface modified porous titanium implants.

    Science.gov (United States)

    Shivaram, Anish; Bose, Susmita; Bandyopadhyay, Amit

    2017-08-01

    Prevention of orthopedic device related infection (ODRI) using antibiotics has met with limited amount of success and is still a big concern during post-surgery. As an alternative, use of silver as an antibiotic treatment to prevent surgical infections is being used due to the well-established antimicrobial properties of silver. However, in most cases silver is used in particulate form with wound dressings or with short-term devices such as catheters but not with load-bearing implants. We hypothesize that strongly adherent silver to load-bearing implants can offer longer term solution to infection in vivo. Keeping that in mind, the focus of this study was to understand the long term release study of silver ions for a period of minimum 6months from silver coated surface modified porous titanium implants. Implants were fabricated using a LENS™ system, a powder based additive manufacturing technique, with at least 25% volume porosity, with and without TiO 2 nanotubes in phosphate buffer saline (pH 7.4) to see if the total release of silver ions is within the toxic limit for human cells. Considering the fact that infection sites may reduce the local pH, silver release was also studied in acetate buffer (pH 5.0) for a period of 4weeks. Along with that, the osseointegrative properties as well as cytotoxicity of porous titanium implants were assessed in vivo for a period of 12weeks using a rat distal femur model. In vivo results indicate that porous titanium implants with silver coating show comparable, if not better, biocompatibility and bonding at the bone-implant interface negating any concerns related to toxicity related to silver to normal cells. The current research is based on our recently patented technology, however focused on understanding longer-term silver release to mitigate infection related problems in load-bearing implants that can even arise several months after the surgery. Prevention of orthopedic device related infection using antibiotics has met

  8. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    Energy Technology Data Exchange (ETDEWEB)

    Katata, G.; Chino, M.; Kobayashi, T. [Japan Atomic Energy Agency (JAEA), Ibaraki (Japan); and others

    2015-07-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I{sub 2} and CH{sub 3}I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of

  9. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    International Nuclear Information System (INIS)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO 2 ) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO 2 pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs

  10. Some problems in the categorization of source terms

    International Nuclear Information System (INIS)

    Abbey, F.; Dunbar, I.H.; Hayns, M.R.; Nixon, W.

    1985-01-01

    In recent years techniques for calculating source terms have been considerably improved. It would be unfortunate if the new information were to be blurred by the use of old schemes for the categorization of source terms. In the past categorization schemes have been devised without the question of the general principles of categorization and the available options being addressed explicitly. In this paper these principles are set out, providing a framework within which categorization schemes used in past probabilistic risk assessments and possible future improvements are discussed. In particular the use of input from scoping consequence calculations in deciding how to group source terms, and the question of how modelling uncertainties may be expressed as uncertainties in a final category source terms are considered

  11. The inventory of sources, environmental releases and risk assessment for perfluorooctane sulfonate in China

    International Nuclear Information System (INIS)

    Zhang Lai; Liu Jianguo; Hu Jianxin; Liu Chao; Guo Weiguang; Wang Qiang; Wang Hong

    2012-01-01

    With about 100 t/y of the production volume, perfluorootane sulfonates (PFOS) are mainly used for metal plating, aqueous fire-fighting foams (AFFFs) and sulfluramidin China, and the use amount is about 30–40 t/y, 25–35 t/y and 4–8 t/y respectively. Based on the inventory of PFOS production and uses with geographic distribution educed from statistics, environmental risk assessment of PFOS was taken by using EUSES model, as well as its environmental releases were estimated both in local and regional levels in China. While the environmental release from manufacture is significant in Central China region, metal plating was identified as the major PFOS release source in regional level. The East China region shows the most strong emission strength of PFOS. Though the predicted environmental concentrations (PECs) were not exceed current relevant predicted no effect concentrations (PNECs) of the risk characterization for PFOS, higher PECs was estimated around major PFOS release sources showing undesirable environmental risk at local level. - Highlights: ► Inventory of production and uses of perfluorooctane sulfonate (PFOS) in China with geographical distribution. ► Characteristics of PFOS release sources and distribution consistent with social-economic situation in China. ► Effective model predicted results of PFOS environmental risk assessment in local and regional scales compared with relevant environmental monitoring data. - Inventory of PFOS production and use of with sectoral and regional distribution of China, environmental releases and risk status were indicated both in the local and regional level of the country.

  12. Estimation of marine source-term following Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    Bailly du Bois, P.; Laguionie, P.; Boust, D.; Korsakissok, I.; Didier, D.; Fiévet, B.

    2012-01-01

    Contamination of the marine environment following the accident in the Fukushima Dai-ichi nuclear power plant represented the most important artificial radioactive release flux into the sea ever known. The radioactive marine pollution came from atmospheric fallout onto the ocean, direct release of contaminated water from the plant and transport of radioactive pollution from leaching through contaminated soil. In the immediate vicinity of the plant (less than 500 m), the seawater concentrations reached 68 000 Bq.L −1 for 134 Cs and 137 Cs, and exceeded 100 000 Bq.L −1 for 131 I in early April. Due to the accidental context of the releases, it is difficult to estimate the total amount of radionuclides introduced into seawater from data obtained in the plant. An evaluation is proposed here, based on measurements performed in seawater for monitoring purposes. Quantities of 137 Cs in seawater in a 50-km area around the plant were calculated from interpolation of seawater measurements. The environmental halftime of seawater in this area is deduced from the time-evolution of these quantities. This halftime appeared constant at about 7 days for 137 Cs. These data allowed estimation of the amount of principal marine inputs and their evolution in time: a total of 27 PBq (12 PBq–41 PBq) of 137 Cs was estimated up to July 18. Even though this main release may be followed by residual inputs from the plant, river runoff and leakage from deposited sediments, it represents the principal source-term that must be accounted for future studies of the consequences of the accident on marine systems. The 137 Cs from Fukushima will remain detectable for several years throughout the North Pacific, and 137 Cs/ 134 Cs ratio will be a tracer for future studies. Highlights: ► Fukushima Dai-ichi accident is the most important artificial radioactive release flux into the sea. ► Quantities of 137 Cs in seawater are deduced from individual measurements. ► Local concentrations in

  13. Effect of different polyphenol sources on the efficiency of ellagic acid release by Aspergillus niger.

    Science.gov (United States)

    Sepúlveda, Leonardo; de la Cruz, Reynaldo; Buenrostro, José Juan; Ascacio-Valdés, Juan Alberto; Aguilera-Carbó, Antonio Francisco; Prado, Arely; Rodríguez-Herrera, Raúl; Aguilar, Cristóbal Noé

    2016-01-01

    Fungal hydrolysis of ellagitannins produces hexahydroxydiphenic acid, which is considered an intermediate molecule in ellagic acid release. Ellagic acid has important and desirable beneficial health properties. The aim of this work was to identify the effect of different sources of ellagitannins on the efficiency of ellagic acid release by Aspergillus niger. Three strains of A. niger (GH1, PSH and HT4) were assessed for ellagic acid release from different polyphenol sources: cranberry, creosote bush, and pomegranate used as substrate. Polyurethane foam was used as support for solid-state culture in column reactors. Ellagitannase activity was measured for each of the treatments. Ellagic acid was quantified by high performance liquid chromatography. When pomegranate polyphenols were used, a maximum value of ellagic acid (350.21 mg/g) was reached with A. niger HT4 in solid-state culture. The highest amount of ellagitannase (5176.81 U/l) was obtained at 8h of culture when cranberry polyphenols and strain A. niger PSH were used. Results demonstrated the effect of different polyphenol sources and A. niger strains on ellagic acid release. It was observed that the best source for releasing ellagic acid was pomegranate polyphenols and A. niger HT4 strain, which has the ability to degrade these compounds for obtaining a potent bioactive molecule such as ellagic acid. Copyright © 2015 Asociación Argentina de Microbiología. Publicado por Elsevier España, S.L.U. All rights reserved.

  14. Determination of the in-containment source term for a Large-Break Loss of Coolant Accident

    International Nuclear Information System (INIS)

    2001-04-01

    This is the report of a project that focused on one of the most important design basis accidents: the Large Break Loss Of Coolant Accident (LBLOCA) (for pressurised water reactors). The first step in the calculation of the radiological consequences of this accident is the determination of the source term inside the containment. This work deals with this part of the calculation of the LBLOCA radiological consequences for which a previous benchmark (1988) has shown wide variations in the licensing practices adopted by European countries. The calculation of this source term may naturally be split in several steps (see chapter II), corresponding to several physical stages in the release of fission products: fraction of core failure, release from the damaged fuel, airborne part of the release and the release into the reactor coolant system and the sumps, chemical behaviour of iodine in the aqueous and gas phases, natural and spray removal in the containment atmosphere. A chapter is devoted to each of these topics. In addition, two other chapters deal with the basic assumptions to define the accidental sequence and the nuclides to be considered when computing doses associated with the LBLOCA. The report describes where there is agreement between the partner organisations and where there are still differences in approach. For example, there is agreement concerning the percentage of failed fuel which could be used in future licensing assessments (however this subject is still under discussion in France, a lower value is thinkable). For existing plants, AVN (Belgium) wishes to keep the initial licensing assumptions. For the release from damaged fuel, there is not complete agreement: AVN (Belgium) wishes to maintain its present approach. IPSN (France), GRS (Germany) and NNC (UK) prefer to use their own methodologies that result in slightly different values to the proposed values for a common position. There are presently no recommendations of the release of fuel particulates

  15. Quantification of uncertainties in source term estimates for a BWR with Mark I containment

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Cazzoli, E.; Davis, R.; Ishigami, T.; Lee, M.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.

    1988-01-01

    A methodology for quantification and uncertainty analysis of source terms for severe accident in light water reactors (QUASAR) has been developed. The objectives of the QUASAR program are (1) to develop a framework for performing an uncertainty evaluation of the input parameters of the phenomenological models used in the Source Term Code Package (STCP), and (2) to quantify the uncertainties in certain phenomenological aspects of source terms (that are not modeled by STCP) using state-of-the-art methods. The QUASAR methodology consists of (1) screening sensitivity analysis, where the most sensitive input variables are selected for detailed uncertainty analysis, (2) uncertainty analysis, where probability density functions (PDFs) are established for the parameters identified by the screening stage and propagated through the codes to obtain PDFs for the outputs (i.e., release fractions to the environment), and (3) distribution sensitivity analysis, which is performed to determine the sensitivity of the output PDFs to the input PDFs. In this paper attention is limited to a single accident progression sequence, namely; a station blackout accident in a BWR with a Mark I containment buildings. Identified as an important accident in the draft NUREG-1150 a station blackout involves loss of both off-site power and DC power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation coding systems

  16. CONTAIN code calculations of the effects on the source term of CsI to I/sub 2/ conversion due to severe hydrogen burns

    International Nuclear Information System (INIS)

    Valdez, G.D.; Williams, D.C.

    1986-01-01

    In experiments conducted at Sandia National Laboratories large amounts of elemental iodine were produced when CsI-Al 2 O 3 aerosol was exposed to hydrogen/air combustion. To evaluate some of the implications of the iodide conversion (observed to occur with up to 75% efficiency) for the severe accident source term, computational simulations of representative accident sequences were conducted with the CONTAIN code. The following conclusions can be drawn from this preliminary source term assessment: (1) If the containment sprays are inoperative during the accident, or failed by the hydrogen burn, the late-time source term is almost tripled when the iodide is converted to I 2 . (2) With the sprays active, the amount released without conversion of the CsI aerosol is 63% higher than for the case when conversion occurs. (3) For the case where CsI is converted to I 2 continued operation of the sprays reduces the release by a factor of 40, relative to the case in which the sprays fail at the time of the hydrogen burn. When there is no conversion, the reduction factor for continued spray operation is about a factor of 9, relative to the failed spray case

  17. New source terms and the implications for emergency planning requirements at nuclear power plants in the United State

    International Nuclear Information System (INIS)

    Kaiser, G.D.; Cheok, M.C.

    1987-01-01

    This paper begins with a brief review of current approaches to source term driven changes to NRC emergency planning requirements and addresses significant differences between them. Approaches by IDCOR and EPRI, industry submittals to NRC and alternative risk-based evaluations have been considered. Important issues are discussed, such as the role of Protective Action Guides in determining the radius of the emergency planning zone (EPZ). The significance of current trends towards the prediction of longer warning times and longer durations of release in new source terms is assessed. These trends may help to relax the current notification time requirements. Finally, the implications of apparent support in the regulations for a threshold in warning time beyond which ad hoc protective measures are adequate is discussed

  18. Phase 1 immobilized low-activity waste operational source term

    International Nuclear Information System (INIS)

    Burbank, D.A.

    1998-01-01

    This report presents an engineering analysis of the Phase 1 privatization feeds to establish an operational source term for storage and disposal of immobilized low-activity waste packages at the Hanford Site. The source term information is needed to establish a preliminary estimate of the numbers of remote-handled and contact-handled waste packages. A discussion of the uncertainties and their impact on the source term and waste package distribution is also presented. It should be noted that this study is concerned with operational impacts only. Source terms used for accident scenarios would differ due to alpha and beta radiation which were not significant in this study

  19. Source term analyses under severe accidents for KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    2001-03-01

    In this study, in-containment source term for LOFW (Loss of Feed Water), which has appeared the most frequent core melt accident, is calculated and compared with NUREG-1465 source term. This study provides not only new source term data using MELCOR1.8.4 and its state-of-the-art models but also evaluating basis of KNGR design and its mitigation capability under severe accidents. As the selected accident is identical with LOFW-S17, which has been analyzed using MAAP by KEPCO with only difference of 2 SITs, mutual comparison of the results is especially expected.

  20. Study of source term evaluation from fuel solution under simulated nuclear criticality accident in TRACY

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Tashiro, Shinsuke; Nagai, Hitoshi; Koike, Tadao; Okagawa, Seigo; Murata, Mikio

    1999-01-01

    In a accident at the dissolver in a reprocessing plant, various fission products and radiolysis gases will be produced in the fuel solution and volatile radioactive nuclides and radiolysis gases and nitrogen oxide will be released into vent-gas spontaneously. Moreover other on-volatile nuclide will be releases as radioactive aerosol (mist) with bursting bubbles at surface of the solution. Therefore quantitative estimation of release and transport behavior of the radioactive material from solution as source term is very important. TRACY is a transient criticality experimental facility for studying the transient criticality characteristics of low enriched uranium. In this paper, experiment methods and results about the release behavior of the hydrogen, radioactive aerosol and iodine species from the fuel solutions are reported. As the results of the experiments, release patterns of H 2 , 140 Ba and 131 I could be grasped. Concentrations of H 2 in the vent-gas and 140 Ba in the gas phase in the core tank attained to the peak just after the transient criticality and decreased exponentially with time. On the other hand, concentrations of 131 I in the gas phase of the tank began to increase with a time lag of several minutes from the transient criticality and attained approximately constant values. (J.P.N.)

  1. The First Data Release of the All-sky NOAO Source Catalog

    Science.gov (United States)

    Nidever, David L.; NOAO Data Lab

    2018-06-01

    Roughly three quarters of the sky has been imaged with NOAO's telescopes from both hemispheres. While the large majority of these data were obtained for PI-led projects and surveys only a fraction have been released to the community via well-calibrated and easily accessible catalogs. We have remedied this by creating a catalog of sources from most of the public data taken on CTIO-4m+DECam as well as KPNO 4m+Mosaic3. The first data release (DR1) of this catalog, called the NOAO Source Catalog (NSC), contains 2.9 billion unique objects, 34 billion individual measurements, covers ~30,000 square degrees, has depths of ~23rd magnitude in most broadband filters with ~1-2% photometric accuracy and astrometric accuracy of ~2 mas. The NSC will be useful for exploring stellar streams, dwarf satellite galaxies, distant galaxies as well as variable stars and other transients. DR1 is now available through the NOAO Data Lab (datalab.noao.edu).

  2. GASPAR-II, Radiation Exposure to Man from Air Releases of Reactor Effluents

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Description of program or function: GASPAR implements the air release dose models of the NRC Regulatory Guide 1.109 for noble gases (semi-infinite plume only) and the radioiodine and particulate emissions. GASPAR computes both population (ALARA-As Low As Reasonably Achievable and NEPA-National Environmental Policy Act) and individual doses. Site data, meteorological data, radionuclide release source terms, and location meteorological data for selected individuals are specified as input data. The site data includes population data and milk, meat, and vegetation production. The meteorological data includes dispersion X/Q, X/Q decayed, X/Q decayed and depleted, and deposition. Population doses, individual doses, and cost benefit tables are calculated. 2 - Method of solution: There are two basic types of calculations, the population dose calculation and the individual dose calculation; however, both may be combined in a single GASPAR execution. There are usually several source terms corresponding to several release points. As the dose is computed for each source term it is accumulated so that the dose printed for the first source term is the actual dose for that term. For all subsequent source terms the dose printed is the accumulated dose with the dose printed for the last source term, the grand total for the problem. For the cost benefit table, individual source term doses are generated. Seven pathways by which the nuclides travel to man are considered. These are plume, ground, inhalation, vegetation, cows' milk, goats' milk, and meat. For the individual dose calculations, man is subdivided into the four age groups of infant (0 to 1 year), child (1-11 years), teenager (12-18 years) and adult (over 18 years). Each of these calculations take into account eight body organs - T. body, G.I. track, bone, liver, kidney, thyroid, lung, and skin. 3 - Restrictions on the complexity of the problem: None noted

  3. An Organizational-Technical Concept to Deal with Open Source Software License Terms

    Directory of Open Access Journals (Sweden)

    Sergius Dyck

    2016-06-01

    Full Text Available Open source software (OSS released under various license terms is widely used as third party libraries in today's software projects. To ensure open source compliance within an organization, a strategic approach to OSS management is needed. As basis for such an approach, we introduce an organizational-technical concept for dealing with the various OSS licenses by using procedural instructions and build automation software. The concept includes the careful consideration of OSS license conditions. The results obtained from this consideration and additional necessary commitments are documented in a so-called license playbook. We introduce procedure instructions enabling a consistent approach for software development using OSS libraries. The procedure instructions are described in a way such that they can be implemented for example for Java projects using the popular build automation tool Apache Maven and the software repository tool Nexus. We give guidance on how to realize such an implementation on basis of automation tools in practice.

  4. Modelling and simulation the radioactive source-term of fission products in PWR type reactors; Modelagem e simulacao do termo-fonte radioativo de produtos de fissao em reatores nucleares do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Porfirio, Rogilson Nazare da Silva

    1996-07-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  5. Design parameters and source terms: Volume 3, Source terms: Revision 0

    International Nuclear Information System (INIS)

    1987-09-01

    The Design Parameters and Source Terms Document was prepared in accordance with DOE request and to provide data for the environmental impact study to be performed in the future for the Deaf Smith County, Texas site for a nuclear waste repository in salt. This document updates a previous unpublished report to the level of the Site Characterization Plan /endash/ Conceptual Design Report, SCP-CDR. The previous unpublished SCC Study identifies the data needs for the Environmental Assessment effort for seven possible salt repository sites

  6. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

  7. The Phebus Fission Product and Source Term International Programmes

    International Nuclear Information System (INIS)

    Clement, B.; Zeyen, R.

    2005-01-01

    The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5 th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

  8. A comparison of world-wide uses of severe reactor accident source terms

    International Nuclear Information System (INIS)

    Ang, M.L.; Frid, W.; Kersting, E.J.; Friederichs, H.G.; Lee, R.Y.; Meyer-Heine, A.; Powers, D.A.; Soda, K.; Sweet, D.

    1994-09-01

    The definitions of source terms to reactor containments and source terms to the environment are discussed. A comparison is made between the TID-14844 example source term and the alternative source term described in NUREG-1465. Comparisons of these source terms to the containments and those used in France, Germany, Japan, Sweden, and the United Kingdom are made. Source terms to the environment calculated in NUREG-1500 and WASH-1400 are discussed. Again, these source terms are compared to those now being used in France, Germany, Japan, Sweden, and the United Kingdom. It is concluded that source terms to the containment suggested in NUREG-1465 are not greatly more conservative than those used in other countries. Technical bases for the source terms are similar. The regulatory use of the current understanding of radionuclide behavior varies among countries

  9. Supplemental analysis of accident sequences and source terms for waste treatment and storage operations and related facilities for the US Department of Energy waste management programmatic environmental impact statement

    International Nuclear Information System (INIS)

    Folga, S.; Mueller, C.; Nabelssi, B.; Kohout, E.; Mishima, J.

    1996-12-01

    This report presents supplemental information for the document Analysis of Accident Sequences and Source Terms at Waste Treatment, Storage, and Disposal Facilities for Waste Generated by US Department of Energy Waste Management Operations. Additional technical support information is supplied concerning treatment of transuranic waste by incineration and considering the Alternative Organic Treatment option for low-level mixed waste. The latest respirable airborne release fraction values published by the US Department of Energy for use in accident analysis have been used and are included as Appendix D, where respirable airborne release fraction is defined as the fraction of material exposed to accident stresses that could become airborne as a result of the accident. A set of dominant waste treatment processes and accident scenarios was selected for a screening-process analysis. A subset of results (release source terms) from this analysis is presented

  10. Modelling short term individual exposure from airborne hazardous releases in urban environments

    International Nuclear Information System (INIS)

    Bartzis, J.G.; Efthimiou, G.C.; Andronopoulos, S.

    2015-01-01

    Highlights: • The statistical behavior of the variability of individual exposure is described with a beta function. • The extreme value in the beta function is properly addressed by [5] correlation. • Two different datasets gave clear support to the proposed novel theory and its hypotheses. - Abstract: A key issue, in order to be able to cope with deliberate or accidental atmospheric releases of hazardous substances, is the ability to reliably predict the individual exposure downstream the source. In many situations, the release time and/or the health relevant exposure time is short compared to mean concentration time scales. In such a case, a significant scatter of exposure levels is expected due to the stochastic nature of turbulence. The problem becomes even more complex when dispersion occurs over urban environments. The present work is the first attempt to approximate on generic terms, the statistical behavior of the abovementioned variability with a beta distribution probability density function (beta-pdf) which has proved to be quite successful. The important issue of the extreme concentration value in beta-pdf seems to be properly addressed by the [5] correlation in which global values of its associated constants are proposed. Two substantially different datasets, the wind tunnel Michelstadt experiment and the field Mock Urban Setting Trial (MUST) experiment gave clear support to the proposed novel theory and its hypotheses. In addition, the present work can be considered as basis for further investigation and model refinements.

  11. Modelling short term individual exposure from airborne hazardous releases in urban environments

    Energy Technology Data Exchange (ETDEWEB)

    Bartzis, J.G., E-mail: bartzis@uowm.gr [University of Western Macedonia, Dept. of Mechanical Engineering, Sialvera & Bakola Str., 50100, Kozani (Greece); Efthimiou, G.C.; Andronopoulos, S. [Environmental Research Laboratory, INRASTES, NCSR Demokritos, Patriarchou Grigoriou & Neapoleos Str., 15310, Aghia Paraskevi (Greece)

    2015-12-30

    Highlights: • The statistical behavior of the variability of individual exposure is described with a beta function. • The extreme value in the beta function is properly addressed by [5] correlation. • Two different datasets gave clear support to the proposed novel theory and its hypotheses. - Abstract: A key issue, in order to be able to cope with deliberate or accidental atmospheric releases of hazardous substances, is the ability to reliably predict the individual exposure downstream the source. In many situations, the release time and/or the health relevant exposure time is short compared to mean concentration time scales. In such a case, a significant scatter of exposure levels is expected due to the stochastic nature of turbulence. The problem becomes even more complex when dispersion occurs over urban environments. The present work is the first attempt to approximate on generic terms, the statistical behavior of the abovementioned variability with a beta distribution probability density function (beta-pdf) which has proved to be quite successful. The important issue of the extreme concentration value in beta-pdf seems to be properly addressed by the [5] correlation in which global values of its associated constants are proposed. Two substantially different datasets, the wind tunnel Michelstadt experiment and the field Mock Urban Setting Trial (MUST) experiment gave clear support to the proposed novel theory and its hypotheses. In addition, the present work can be considered as basis for further investigation and model refinements.

  12. Effects of Symmetrically Arranged Heat Sources on the Heat Release Performance of Extruded-Type Heat Sinks

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Min Ye [Chonbuk National Univ., Chonju (Korea, Republic of)

    2016-02-15

    In this study we investigated the effects of symmetrically arranged heat sources on the heat release performances of extruded-type heat sinks through experiments and thermal fluid simulations. Also, based on the results we suggested a high-efficiency and cost-effective heat sink for a solar inverter cooling system. In this parametric study, the temperatures between heaters on the base plate and the heat release rates were investigated with respect to the arrangements of heat sources and amounts of heat input. Based on the results we believe that the use of both sides of the heat sink is the preferred method for releasing the heat from the heat source to the ambient environment rather than the use of a single side of the heat sink. Also from the results, it is believed that the symmetric arrangement of the heat sources is recommended to achieve a higher rate of heat transfer. From the results of the thermal fluid simulation, it was possible to confirm the qualitative agreement with the experimental results. Finally, quantitative comparison with respect to mass flow rates, heat inputs, and arrangements of the heat source was also performed.

  13. Source term development for tritium at the Sheffield disposal site

    International Nuclear Information System (INIS)

    MacKenzie, D.R.; Barletta, R.E.; Smalley, J.F.; Kempf, C.R.; Davis, R.E.

    1984-01-01

    The Sheffield low-level radioactive waste disposal site, which ceased operation in 1978, has been the focus of modeling efforts by the NRC for the purpose of predicting long-term site behavior. To provide the NRC with the information required for its modeling effort, a study to define the source term for tritium in eight trenches at the Sheffield site has been undertaken. Tritium is of special interest since significant concentrations of the isotope have been found in groundwater samples taken at the site and at locations outside the site boundary. Previous estimates of tritium site inventory at Sheffield are in wide disagreement. In this study, the tritium inventory in the eight trenches was estimated by reviewing the radioactive shipping records (RSRs) for waste buried in these trenches. It has been found that the tritium shipped for burial at the site was probably higher than previously estimated. In the eight trenches surveyed, which amount to roughly one half the total volume and activity buried at Sheffield, approximately 2350 Ci of tritium from non-fuel cycle sources were identified. The review of RSRs also formed the basis for obtaining waste package descriptions and for contacting large waste generators to obtain more detailed information regarding these waste packages. As a result of this review and the selected generator contacts, the non-fuel cycle tritium waste was categorized. The tritium releases from each of these waste categories were modeled. The results of this modeling effort are presented for each of the eight trenches selected. 3 references, 2 figures

  14. 41 CFR 102-37.465 - May a SASP modify or release any of the terms and conditions of donation?

    Science.gov (United States)

    2010-07-01

    ... release any of the terms and conditions of donation? 102-37.465 Section 102-37.465 Public Contracts and... REGULATION PERSONAL PROPERTY 37-DONATION OF SURPLUS PERSONAL PROPERTY Donations to Public Agencies, Service... SASP modify or release any of the terms and conditions of donation? You may alter or grant releases...

  15. Site-sensitive hazards of potential airborne radioactive release from sources on the Kola peninsula

    International Nuclear Information System (INIS)

    Bergman, R.; Thaning, L.; Baklanov, A.

    1998-02-01

    In this work we focus on cases of airborne releases from some of the sources on the Kola Peninsula - primarily nuclear reactors on submarines and the Kola Nuclear Power Plant (KNPP). The purpose of our study is to illustrate, and discuss some features - dependent on site and release characteristics - of the deposition patterns resulting from assumed unit radioactive releases to the atmosphere from a location at a fjord and from the KNPP in Polyarnye Zori. Using meteorological data for one real weather situation, the analysis is based on simulating the transport in air of assumed radioactive releases and estimating the deposition pattern on local, meso- and regional scales. By allowing unit releases to occur simultaneously from the site at the fjord and from the power plant (and with the same release profile in time) comparisons are made of differences in deposition patterns in and outside the Kola region. In this case study a set of assumed release heights, durations of the release, and particle size distributions are applied to indicate the dependence for the resulting deposition pattern on these parameters

  16. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    Energy Technology Data Exchange (ETDEWEB)

    Routamo, T [Imatran Voima Oy, Vantaa (Finland)

    1996-12-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I{sub 2} hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I{sub 2} is produced from I{sup -}very rapidly. High temperature in the reactor coolant causes I{sub 2} to be transformed into HIO and through the boiling of the coolant volatile I{sub 2} and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I{sup -} release to be much lower than those of I{sub 2} and HIO. (author) 15 figs., 1 tab., refs.

  17. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    International Nuclear Information System (INIS)

    Routamo, T.

    1996-01-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I 2 hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I 2 is produced from I - very rapidly. High temperature in the reactor coolant causes I 2 to be transformed into HIO and through the boiling of the coolant volatile I 2 and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I - release to be much lower than those of I 2 and HIO. (author) 15 figs., 1 tab., refs

  18. Sensitivity analysis and benchmarking of the BLT low-level waste source term code

    International Nuclear Information System (INIS)

    Suen, C.J.; Sullivan, T.M.

    1993-07-01

    To evaluate the source term for low-level waste disposal, a comprehensive model had been developed and incorporated into a computer code, called BLT (Breach-Leach-Transport) Since the release of the original version, many new features and improvements had also been added to the Leach model of the code. This report consists of two different studies based on the new version of the BLT code: (1) a series of verification/sensitivity tests; and (2) benchmarking of the BLT code using field data. Based on the results of the verification/sensitivity tests, the authors concluded that the new version represents a significant improvement and it is capable of providing more realistic simulations of the leaching process. Benchmarking work was carried out to provide a reasonable level of confidence in the model predictions. In this study, the experimentally measured release curves for nitrate, technetium-99 and tritium from the saltstone lysimeters operated by Savannah River Laboratory were used. The model results are observed to be in general agreement with the experimental data, within the acceptable limits of uncertainty

  19. A source classification framework supporting pollutant source mapping, pollutant release prediction, transport and load forecasting, and source control planning for urban environments

    DEFF Research Database (Denmark)

    Lützhøft, Hans-Christian Holten; Donner, Erica; Wickman, Tonie

    2012-01-01

    for this purpose. Methods Existing source classification systems were examined by a multidisciplinary research team, and an optimised SCF was developed. The performance and usability of the SCF were tested using a selection of 25 chemicals listed as priority pollutants in Europe. Results The SCF is structured...... in the form of a relational database and incorporates both qualitative and quantitative source classification and release data. The system supports a wide range of pollution monitoring and management applications. The SCF functioned well in the performance test, which also revealed important gaps in priority...

  20. Instant-release fractions for the assessment of used nuclear fuel disposal

    International Nuclear Information System (INIS)

    Garisto, N.C.; Vance, E.R.; Stroes-Gascoyne, S.; Johnson, L.H.

    1989-02-01

    Quantitative estimates of instant-release fractions for the potential release of radionuclides from used CANDU fuel in an underground disposal vault have been made in terms of probability- density functions, taking variability and uncertainty into account. The radionuclides included in this study are 129 I, 135 Cs, 79 Se, 126 Sn, 99 Tc, 14 C, and 3 H. The probability-density functions are based on experimental data on the short term release of radionuclides upon contact with groundwater, and on a knowledge of the solid-state chemistry of used fuel. They provide source terms for the environmental and safety assessment of used nuclear fuel disposal

  1. Utility view of the source term and air cleaning

    International Nuclear Information System (INIS)

    Littlefield, P.S.

    1985-01-01

    The utility view of the source term and air cleaning is discussed. The source term is made up of: (1) noble gases, which there has been a tendency to ignore in the past because it was thought there was nothing that could be done with them anyway, (2) the halogens, which have been dealt with in Air Cleaning Conferences in the past in terms of charcoal and other systems for removing them, and (3) the solid components of the source term which particulate filters are designed to handle. Air cleaning systems consist of filters, adsorbers, containment sprays, suppression pools in boiling water reactors and ice beds in ice condenser-equipped plants. The feasibility and cost of air cleaning systems are discussed

  2. Long-Term Outcome of Open Plantar Fascia Release.

    Science.gov (United States)

    MacInnes, Alasdair; Roberts, Sam C; Kimpton, Jessica; Pillai, Anand

    2016-01-01

    Plantar fasciitis is thought to be a self-limiting condition best treated by conservative measures, but despite this many patients have a prolonged duration of symptoms and surgery may be indicated. Partial plantar fascial release is reported to have a short-term success rate of up to 80%, but anecdotally this was not thought to represent our local experience. An audit of long-term patient-reported outcomes following open partial plantar fascia release was performed. A total of 30 patients (33 feet) were identified over a 10-year period and case notes were reviewed. Patients were contacted by letter and invited to complete 2 validated patient-reported outcome score questionnaires (Visual Analog Scale-Foot and Ankle [VAS-FA] and Manchester Oxford Foot Questionnaire [MOXFQ]). Responses were received from 24 patients (26 feet). The average ages were 42.4 (range 24-61) for male and 46.2 (range 33-60) for female patients, with a female/male ratio of 2.7:1. The average duration of treatment prior to operative intervention was 3.1 years (range 1-5). Preoperatively, our cohort underwent a range of conservative measures. All patients were reviewed postoperatively, and average time from surgery to completion of questionnaires was 80 months (range 14-130). The outcomes were worse in patients who had received preoperative steroid injections and this was found to be statistically significant. The mean MOXFQ score was 33.6 ± 3.9 (0-64). Mean VAS-FA score was 57.8 ± 4.9 (24-100). This study found a negative correlation between duration of follow-up and outcome, in both MOXFQ and VAS-FA, showing that patients continued to improve many years postoperatively. The authors also found worse outcomes with preoperative steroid injections, better outcomes in older patients, and a weak gender bias, suggesting results in men were better than those in women. A prolonged recovery period and generally poor outcomes leads the authors to suggest that open plantar fascia release is of

  3. Non-Darwinian evolution for the source detection of atmospheric releases

    Science.gov (United States)

    Cervone, Guido; Franzese, Pasquale

    2011-08-01

    A non-Darwinian evolutionary algorithm is presented as search engine to identify the characteristics of a source of atmospheric pollutants, given a set of concentration measurements. The algorithm drives iteratively a forward dispersion model from tentative sources toward the real source. The solutions of non-Darwinian evolution processes are not generated through pseudo-random operators, unlike traditional evolutionary algorithms, but through a reasoning process based on machine learning rule generation and instantiation. The new algorithm is tested with both a synthetic case and with the Prairie Grass field experiment. To further test the capabilities of the algorithm to work in real-world scenarios, the source identification of all Prairie Grass releases was performed with a decreasing number of sensor measurements, and a relationship is found between the precision of the solution, the number of sensors available, and the levels of concentration measured by the sensors. The proposed methodology can be used for a variety of optimization problems, and is particularly suited for problems where the operations needed for evaluating new candidate solutions are computationally expensive.

  4. STEPS: source term estimation based on plant status phase 0 - the technical specifications of the containment module

    International Nuclear Information System (INIS)

    Vamanu, D.V.

    1998-01-01

    In the framework of Project RODOS (Real-Time On-Line Decision Support System for Nuclear Emergencies in Europe), the European Atomic Energy Community (EAEC) of the Commission of the European Communities has commissioned the development of a unified concept, body of knowledge, models and software package meant to assist the evaluation of the source term of severe nuclear accidents in light water reactors of the types prevailing on the Continent. Code-named STEPS, for 'Source Term Estimation based on Plant Status', the project has evolved as Contract RODOS D (FI4P-CT96-0048), between EAEC and consortium of expert European centres including Commissariat a l'Energie Atomique, Institut de Protection et de Surete Nucleaire, (CEA-DPI-SEAC) as Coordinator and Forschungszentrum Karlsruhe GmbH (FZK-INR), the Finnish Centre for Radiation and Nuclear Safety, (STUK-NSD), the Technical Research Centre of Finland, Energy, Nuclear Energy (VTT-ET-NE), and Eidgenossische Technische Hochschule - ETH Zurich, Centre of Excellence (ETH-CERS) as Contractors. For the Phase 0 of the project, an IFIN-HH expert has been assigned by ETH-CERS to develop the Technical Specifications of the delivery component of the intended STEPS package, the CONTAINMENT Module. Sponsored by ETH-CERS headquarters in Zurich, the work was done on the premises and with the logistic support of CEA D PI-SEAC at Fontenay-aux-Roses, with the feedback processing and computer code development subsequently performed in Bucharest. The Technical Specifications of the STEPS CONTAINMENT Module were guided by specific terms of reference, including: (i) the capability of the software to function as a source term interface between targeted nuclear power plants and the RODOS System; (ii) the comparable capability of the system to be operated as a stand-alone assessment and decision support tool for a comprehensive variety of plants, nuclear emergency classes. On the technical side, the specifications had to focus on the possible

  5. Scenarios and analytical methods for UF6 releases at NRC-licensed fuel cycle facilities

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Dykstra, J.; Holt, D.D.; Huxtable, W.P.; Just, R.A.; Williams, W.R.

    1984-06-01

    This report identifies and discusses potential scenarios for the accidental release of UF 6 at NRC-licensed UF 6 production and fuel fabrication facilities based on a literature review, site visits, and DOE enrichment plant experience. Analytical tools needed for evaluating source terms for such releases are discussed, and the applicability of existing methods is reviewed. Accident scenarios are discussed under the broad headings of cylinder failures, UF 6 process system failures, nuclear criticality events, and operator errors and are categorized by location, release source, phase of UF 6 prior to release, release flow characteristics, release causes, initiating events, and UF 6 inventory at risk. At least three types of releases are identified for further examination: (1) a release from a liquid-filled cylinder outdoors, (2) a release from a pigtail or cylinder in a steam chest, (3) an indoor release from either (a) a pigtail or liquid-filled cylinder or (b) other indoor source depending on facility design and operating procedures. Indoor release phenomena may be analyzed to determine input terms for a ventilation model by using a time-dependent homogeneous compartment model or a more complex hydrodynamic model if time-dependent, spatial variations in concentrations, temperature, and pressure are important. Analytical tools for modeling directed jets and explosive releases are discussed as well as some of the complex phenomena to be considered in analyzing UF 6 releases both indoors and outdoors

  6. Unclassified Sources Term and Radionuclide Data for Corrective Action Unit 97: Yucca Flat/Climax Mine, Nevada Test Site, Nevada, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Peter Martian

    2009-08-01

    This report documents the evaluation of the information and data available on the unclassified source term and radionuclide contamination for CAU 97: Yucca Flat/Climax Mine. The total residual inventory of radionuclides associated with one or more tests is known as the radiologic source term (RST). The RST is comprised of radionuclides in water, glass, or other phases or mineralogic forms. The hydrologic source term (HST) of an underground nuclear test is the portion of the total RST that is released into the groundwater over time following the test. In this report, the HST represents radionuclide release some time after the explosion and does not include the rapidly evolving mechanical, thermal, and chemical processes during the explosion. The CAU 97: Yucca Flat/Climax Mine has many more detonations and a wider variety of settings to consider compared to other CAUs. For instance, the source term analysis and evaluation performed for CAUs 101 and 102: Central and Western Pahute Mesa and CAU 98: Frenchman Flat did not consider vadose zone attenuation because many detonations were located near or below the water table. However, the large number of Yucca Flat/Climax Mine tests and the location of many tests above the water table warrant a more robust analysis of the unsaturated zone. The purpose of this report is to develop and document conceptual models of the Yucca Flat/Climax Mine HST for use in implementing source terms for the Yucca Flat/Climax Mine models. This document presents future plans to incorporate the radionuclide attenuation mechanisms due to unsaturated/multiphase flow and transport within the Yucca Flat CAU scale modeling. The important processes that influence radionuclide migration for the unsaturated and saturated tests in alluvial, volcanic, and carbonate settings are identified. Many different flow and transport models developed by Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), including original

  7. Estimate of airborne release of plutonium from Babcock and Wilcox plant as a result of severe wind hazard and earthquake

    International Nuclear Information System (INIS)

    Mishima, J.; Schwendiman, L.C.; Ayer, J.E.

    1978-10-01

    As part of an interdisciplinary study to evaluate the potential radiological consequences of wind hazard and earthquake upon existing commercial mixed oxide fuel fabrication plants, the potential mass airborne releases of plutonium (source terms) from such events are estimated. The estimated souce terms are based upon the fraction of enclosures damaged to three levels of severity (crush, puncture penetrate, and loss of external filter, in order of decreasing severity), called damage ratio, and the airborne release if all enclosures suffered that level of damage. The discussion of damage scenarios and source terms is divided into wind hazard and earthquake scenarios in order of increasing severity. The largest airborne releases from the building were for cases involving the catastrophic collapse of the roof over the major production areas--wind hazard at 110 mph and earthquakes with peak ground accelerations of 0.20 to 0.29 g. Wind hazards at higher air velocities and earthquakes with higher ground acceleration do not result in significantly greater source terms. The source terms were calculated as additional mass of respirable particles released with time up to 4 days; and, under these assumptions, approximately 98% of the mass of material of concern is made airborne from 2 h to 4 days after the event. The overall building source terms from the damage scenarios evaluated are shown in a table. The contribution of individual areas to the overall building source term is presented in order of increasing severity for wind hazard and earthquake

  8. Reassessment of the technical bases for estimating source terms. Final report

    International Nuclear Information System (INIS)

    Silberberg, M.; Mitchell, J.A.; Meyer, R.O.; Ryder, C.P.

    1986-07-01

    This document describes a major advance in the technology for calculating source terms from postulated accidents at US light-water reactors. The improved technology consists of (1) an extensive data base from severe accident research programs initiated following the TMI accident, (2) a set of coupled and integrated computer codes (the Source Term Code Package), which models key aspects of fission product behavior under severe accident conditions, and (3) a number of detailed mechanistic codes that bridge the gap between the data base and the Source Term Code Package. The improved understanding of severe accident phenonmena has also allowed an identification of significant sources of uncertainty, which should be considered in estimating source terms. These sources of uncertainty are also described in this document. The current technology provides a significant improvement in evaluating source terms over that available at the time of the Reactor Safety Study (WASH-1400) and, because of this significance, the Nuclear Regulatory Commission staff is recommending its use

  9. Regulatory impact of nuclear reactor accident source term assumptions. Technical report

    International Nuclear Information System (INIS)

    Pasedag, W.F.; Blond, R.M.; Jankowski, M.W.

    1981-06-01

    This report addresses the reactor accident source term implications on accident evaluations, regulations and regulatory requirements, engineered safety features, emergency planning, probabilistic risk assessment, and licensing practice. Assessment of the impact of source term modifications and evaluation of the effects in Design Basis Accident analyses, assuming a change of the chemical form of iodine from elemental to cesium iodide, has been provided. Engineered safety features used in current LWR designs are found to be effective for all postulated combinations of iodine source terms under DBA conditions. In terms of potential accident consequences, it is not expected that the difference in chemical form between elemental iodine and cesium iodide would be significant. In order to account for the current information on source terms, a spectrum of accident scenerios is discussed to realistically estimate the source terms resulting from a range of potential accident conditions

  10. Helium release from radioisotope heat sources

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

    1984-05-01

    Diffusion of helium in /sup 238/PuO/sub 2/ fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel.

  11. Helium release from radioisotope heat sources

    International Nuclear Information System (INIS)

    Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

    1984-05-01

    Diffusion of helium in 238 PuO 2 fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel

  12. Elucidating the Role of Carbon Sources on Abiotic and Biotic Release of Arsenic into Cambodian Aquifers

    Science.gov (United States)

    Koeneke, M.

    2017-12-01

    Arsenic (As) is a naturally occurring contaminant in Cambodia that has been contaminating well-water sources of millions of people. Commonly, studies look into the biotic factors that cause the arsenic to be released from aquifer sediments to groundwater. However, abiotic release of As from sediments, though little studied, may also play key roles in As contamination of well water. The goal of this research is to quantitatively compare organic-carbon mediated abiotic and biotic release of arsenic from sediments to groundwater. Batch anaerobic incubation experiments under abiotic (sodium azide used to immobilize microbes) and biotic conditions were conducted using Cambodian aquifer sediments, four different organic carbon sources (sodium lactate, sodium citrate, sodium oxalate, and humic acid), and six different carbon concentrations (0, 1, 2.5, 5, 10, 25mg C/L). Dissolved arsenic, iron(Fe), and manganese(Mn) concentrations in the treatments were measured 112 days . In addition, sediment and solution carbon solution was measured . Collectively, these show how different carbon sources, different carbon concentrations, and how abiotic and biotic factors impact the release of arsenic from Cambodian sediments into aquifers. Overall, an introduction of organic carbon to the soil increases the amount of As released from the sediment. The biotic + abiotic and abiotic conditions seemed to play a minimal role in the amount of As released. Dissolved species analysis showed us that 100% of the As was As(V), Our ICP-MS results vary due to the heterogeneity of samples, but when high levels are Fe are seen in solution, we also see high levels of As. We also see higher As concentrations when there is a smaller amount of Mn in solution.

  13. Hazardous constituent source term. Revision 2

    International Nuclear Information System (INIS)

    1994-01-01

    The Department of Energy (DOE) has several facilities that either generate and/or store transuranic (TRU)-waste from weapons program research and production. Much of this waste also contains hazardous waste constituents as regulated under Subtitle C of the Resource Conservation and Recovery Act (RCRA). Toxicity characteristic metals in the waste principally include lead, occurring in leaded rubber gloves and shielding. Other RCRA metals may occur as contaminants in pyrochemical salt, soil, debris, and sludge and solidified liquids, as well as in equipment resulting from decontamination and decommissioning activities. Volatile organic compounds (VOCS) contaminate many waste forms as a residue adsorbed on surfaces or occur in sludge and solidified liquids. Due to the presence of these hazardous constituents, applicable disposal regulations include land disposal restrictions established by Hazardous and Solid Waste Amendments (HSWA). The DOE plans to dispose of TRU-mixed waste from the weapons program in the Waste Isolation Pilot Plant (WIPP) by demonstrating no-migration of hazardous constituents. This paper documents the current technical basis for methodologies proposed to develop a post-closure RCRA hazardous constituent source term. For the purposes of demonstrating no-migration, the hazardous constituent source term is defined as the quantities of hazardous constituents that are available for transport after repository closure. Development of the source term is only one of several activities that will be involved in the no-migration demonstration. The demonstration will also include uncertainty and sensitivity analyses of contaminant transport

  14. Source Term Model for Fine Particle Resuspension from Indoor Surfaces

    National Research Council Canada - National Science Library

    Kim, Yoojeong; Gidwani, Ashok; Sippola, Mark; Sohn, Chang W

    2008-01-01

    This Phase I effort developed a source term model for particle resuspension from indoor surfaces to be used as a source term boundary condition for CFD simulation of particle transport and dispersion in a building...

  15. Environmental transport and long-term exposure for tritium released in the biosphere

    International Nuclear Information System (INIS)

    Bergman, R.; Bergstroem, U.; Evans, S.

    1979-01-01

    Global cycling of tritium is studied with regard to long-term exposure and dose. Dose and dose commitment are calculated for releases at different latitudes to the troposphere, land and upper ocean layer, with particular regard to effects from release into recipients of intermediate size as, for example, the Baltic Sea. The global transport of tritium appears to be governed by first order kinetics. Compartment models based on linear differential equation systems, as used in this study, should therefore be adequate. The realism and applicability of ecological compartment models are analysed with respect to completeness of the systems of reservoirs and pathways as well as accuracy in assumed reservoir sizes and exchange rates. By introducing different biospheric reservoirs and transfer mechanisms, important carriers and recipients are identified for the analysis of tritium released to air, land and water. Terrestrial biota and groundwater are shown to be significant both with regard to reservoir sizes and influence on the land-troposphere and land-sea exchange of tritium. Model studies regarding the conversion of HT to HTO in different biospheric reservoirs indicate that an atmospheric release of HT may yield up to 1.7 times the dose commitment obtained after release of the same amount of tritium as HTO. The global collective dose commitment from a tropospheric release of tritium is 0.002-0.004 man.rem per Ci depending on the latitude at the release point. Release to the surface ocean layers gives a ten times lower collective dose. (author)

  16. Scoping Analysis of Source Term and Functional Containment Attenuation Factors

    Energy Technology Data Exchange (ETDEWEB)

    Pete Lowry

    2012-10-01

    In order to meet future regulatory requirements, the Next Generation Nuclear Plant (NGNP) Project must fully establish and validate the mechanistic modular high temperature gas-cooled reactor (HTGR) source term. This is not possible at this stage in the project, as significant uncertainties in the final design remain unresolved. In the interim, however, there is a need to establish an approximate characterization of the source term. The NGNP team developed a simplified parametric model to establish mechanistic source term estimates for a set of proposed HTGR configurations.

  17. Scoping Analysis of Source Term and Functional Containment Attenuation Factors

    Energy Technology Data Exchange (ETDEWEB)

    Pete Lowry

    2012-02-01

    In order to meet future regulatory requirements, the Next Generation Nuclear Plant (NGNP) Project must fully establish and validate the mechanistic modular high temperature gas-cooled reactor (HTGR) source term. This is not possible at this stage in the project, as significant uncertainties in the final design remain unresolved. In the interim, however, there is a need to establish an approximate characterization of the source term. The NGNP team developed a simplified parametric model to establish mechanistic source term estimates for a set of proposed HTGR configurations.

  18. Scoping Analysis of Source Term and Functional Containment Attenuation Factors

    Energy Technology Data Exchange (ETDEWEB)

    Pete Lowry

    2012-01-01

    In order to meet future regulatory requirements, the Next Generation Nuclear Plant (NGNP) Project must fully establish and validate the mechanistic modular high temperature gas-cooled reactor (HTGR) source term. This is not possible at this stage in the project, as significant uncertainties in the final design remain unresolved. In the interim, however, there is a need to establish an approximate characterization of the source term. The NGNP team developed a simplified parametric model to establish mechanistic source term estimates for a set of proposed HTGR configurations.

  19. An appreciation of the events, models and data used for LMFBR radiological source term estimations

    International Nuclear Information System (INIS)

    Keir, D.; Clough, P.N.

    1989-01-01

    In this report, the events, models and data currently available for analysis of accident source terms in liquid metal cooled fast neutron reactors are reviewed. The types of hypothetical accidents considered are the low probability, more extreme types of severe accident, involving significant degradation of the core and which may lead to the release of radionuclides. The base case reactor design considered is a commercial scale sodium pool reactor of the CDFR type. The feasibility of an integrated calculational approach to radionuclide transport and speciation (such as is used for LWR accident analysis) is explored. It is concluded that there is no fundamental obstacle, in terms of scientific data or understanding of the phenomena involved, to such an approach. However this must be regarded as a long-term goal because of the large amount of effort still required to advance development to a stage comparable with LWR studies. Particular aspects of LMFBR severe accident phenomenology which require attention are the behaviour of radionuclides during core disruptive accident bubble formation and evolution, and during the less rapid sequences of core melt under sodium. The basic requirement for improved thermal hydraulic modelling of core, coolant and structural materials, in these and other scenarios, is highlighted as fundamental to the accuracy and realism of source term estimations. The coupling of such modelling to that of radionuclide behaviour is seen as the key to future development in this area

  20. Actinide Source Term Program, position paper. Revision 1

    International Nuclear Information System (INIS)

    Novak, C.F.; Papenguth, H.W.; Crafts, C.C.; Dhooge, N.J.

    1994-01-01

    The Actinide Source Term represents the quantity of actinides that could be mobilized within WIPP brines and could migrate with the brines away from the disposal room vicinity. This document presents the various proposed methods for estimating this source term, with a particular focus on defining these methods and evaluating the defensibility of the models for mobile actinide concentrations. The conclusions reached in this document are: the 92 PA open-quotes expert panelclose quotes model for mobile actinide concentrations is not defensible; and, although it is extremely conservative, the open-quotes inventory limitsclose quotes model is the only existing defensible model for the actinide source term. The model effort in progress, open-quotes chemical modeling of mobile actinide concentrationsclose quotes, supported by a laboratory effort that is also in progress, is designed to provide a reasonable description of the system and be scientifically realistic and supplant the open-quotes Inventory limitsclose quotes model

  1. Development of source term PIRT of Fukushima Daiichi NPPs accident

    International Nuclear Information System (INIS)

    Suehiro, S.; Okamoto, K.

    2017-01-01

    The severe accident evaluation committee of AESJ (Atomic Energy Society of Japan) developed the thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and the source term PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aimed to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the code. The source term PIRT was divided into 3 phases for the time domain and 9 categories for the spatial domain. The 68 phenomena were extracted and the importance from viewpoint of the source term was ranked through brainstorming and discussion. This paper describes the developed source term PIRT list and summarized the high ranked phenomena in each phase. (author)

  2. Modeling of fission product release in integral codes

    International Nuclear Information System (INIS)

    Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.

    2014-01-01

    The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

  3. SARNET integrated European Severe Accident Research-Conclusions in the source term area

    Energy Technology Data Exchange (ETDEWEB)

    Haste, T., E-mail: tim.haste@irsn.f [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Giordano, P. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Herranz, L. [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain); Girault, N.; Dubourg, R. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Sabroux, J.-C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Saclay Research Centre, BP 68, F-91192 Gif-sur-Yvette Cedex (France); Cantrel, L. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Bottomley, D. [European Commission Joint Research Centre, Transuranium Institute, P.O. Box 2340, D-76125 Karlsruhe (Germany); Parozzi, F. [ENEA - Ricerca sul Sistema Elettrico (ERSE) SpA., Via Rubattino 54, I-20134 Milano (Italy); Auvinen, A. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT Espoo (Finland); Dickinson, S. [National Nuclear Laboratory, Harwell Business Centre, Didcot, OX11 0QJ (United Kingdom); Lamy, J.-C. [Electricite de France, 12-14 avenue Dutrievoz, F-69100 Villeurbanne (France); Weber, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsgelaende, D-85748 Garching (Germany); Albiol, T. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France)

    2009-12-15

    The overall aim of the SARNET (Severe Accident Research NETwork), in the EU 6th Framework programme was to integrate in a sustainable manner the research capabilities of fifty-one European organisations from eighteen member states of the European Union (EU) plus the Joint Research Centres, with one Canadian company, to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasised integrating activities, spreading of excellence (including knowledge transfer) and jointly executed research, with the knowledge gained being encapsulated in the European severe accident modelling code ASTEC. This paper summarises the achievements over the whole project in the Source Term Topic, which dealt with potential radioactive release to the environment, covering release of fission products and structural materials from the core, their transport in the primary circuit, and their behaviour in the containment. The main technical areas covered, as emphasised by the earlier EURSAFE project, were the effect of oxidative conditions on fission product release and transport (especially the behaviour of the highly radiotoxic ruthenium under air ingress conditions), iodine volatility in the primary circuit, control rod aerosol release (Ag-In-Cd) that affects iodine transport, containment by-pass in the case of steam generator tube rupture, aerosol retention in containment cracks, aerosol remobilisation in the circuit, and iodine/ruthenium behaviour in the containment especially concerning the volatile fraction in the atmosphere. The studies also covered performance of new experiments, analysis of existing data, and formulation and improvement of theoretical models. Significant progress was made in each area. Looking to the future, the 7th Framework successor project SARNET2 covers the remaining issues concerning iodine and ruthenium, including practical

  4. SARNET. Severe Accident Research Network - key issues in the area of source term

    International Nuclear Information System (INIS)

    Giordano, P.; Micaelli, J.C.; Haste, T.; Herranz, L.

    2005-01-01

    scenario studies assisted in defining test conditions for new experiments. Regarding predictability of iodine species exiting the Reactor Coolant System (RCS), which affects its source into the containment, iodine behaviour in the circuit and silver-indium-cadmium release from the core have been reviewed (these absorber elements play a significant role in iodine chemistry). The design of new experiments has been discussed. Concerning the radioactive aerosol source term, work is conducted in the risk-relevant areas of by-pass accident sequences, transport through cracks in containment walls and revaporisation from previous deposits in the RCS that could lead to a delayed source term. Modelling approaches for aerosol retention in containment cracks and interpretation of experimental results on retention in secondary side of SG are already proposed. Amongst the numerous physical and chemical processes which affect the intense reactivity of iodine in containments, mass transfer phenomena were first addressed in the network, looking at SISYPHE experimental interpretation and associated modelling. (author)

  5. Source terms derived from analyses of hypothetical accidents, 1950-1986

    International Nuclear Information System (INIS)

    Stratton, W.R.

    1987-01-01

    This paper reviews the history of reactor accident source term assumptions. After the Three Mile Island accident, a number of theoretical and experimental studies re-examined possible accident sequences and source terms. Some of these results are summarized in this paper

  6. Analysis of source term aspects in the experiment Phebus FPT1 with the MELCOR and CFX codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Fuertes, F. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. E-mail: francisco.martinfuertes@upm.es; Barbero, R. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Martin-Valdepenas, J.M. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Jimenez, M.A. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain)

    2007-03-15

    Several aspects related to the source term in the Phebus FPT1 experiment have been analyzed with the help of MELCOR 1.8.5 and CFX 5.7 codes. Integral aspects covering circuit thermalhydraulics, fission product and structural material release, vapours and aerosol retention in the circuit and containment were studied with MELCOR, and the strong and weak points after comparison to experimental results are stated. Then, sensitivity calculations dealing with chemical speciation upon release, vertical line aerosol deposition and steam generator aerosol deposition were performed. Finally, detailed calculations concerning aerosol deposition in the steam generator tube are presented. They were obtained by means of an in-house code application, named COCOA, as well as with CFX computational fluid dynamics code, in which several models for aerosol deposition were implemented and tested, while the models themselves are discussed.

  7. Estimated airborne release of plutonium from Westinghouse Cheswick site as a result of postulated damage from severe wind and seismic hazard

    International Nuclear Information System (INIS)

    Mishima, J.; Schwendiman, L.C.; Ayer, J.E.

    1979-06-01

    The potential airborne releases of plutonium (source terms) from postulated damage sustained by the Westinghouse Plutonium Fuel Development Laboratories at the Cheswick site in Pennsylvania as a result of various levels of wind and seismic hazard are estimated. The source terms are based on damage scenarios originated by other specialists and range up to 260 mph for wind hazard and in excess of 0.39 g ground acceleration for seismic hazard. The approaches and factors used to estimate the source terms (inventories of dispersible materials at risk, damage levels and ratios, fractional airborne releases of dispersible materials under stress, atmosphere exchange rates, and source term ranges) are discussed. Source term estimates range from less than 10 -7 g plutonium to greater than 130 g plutonium over a four-day period

  8. Non-domestic phosphorus release in rivers during low-flow: Mechanisms and implications for sources identification

    Science.gov (United States)

    Dupas, Rémi; Tittel, Jörg; Jordan, Phil; Musolff, Andreas; Rode, Michael

    2018-05-01

    A common assumption in phosphorus (P) load apportionment studies is that P loads in rivers consist of flow independent point source emissions (mainly from domestic and industrial origins) and flow dependent diffuse source emissions (mainly from agricultural origin). Hence, rivers dominated by point sources will exhibit highest P concentration during low-flow, when flow dilution capacity is minimal, whereas rivers dominated by diffuse sources will exhibit highest P concentration during high-flow, when land-to-river hydrological connectivity is maximal. Here, we show that Soluble Reactive P (SRP) concentrations in three forested catchments free of point sources exhibited seasonal maxima during the summer low-flow period, i.e. a pattern expected in point source dominated areas. A load apportionment model (LAM) is used to show how point sources contribution may have been overestimated in previous studies, because of a biogeochemical process mimicking a point source signal. Almost twenty-two years (March 1995-September 2016) of monthly monitoring data of SRP, dissolved iron (Fe) and nitrate-N (NO3) were used to investigate the underlying mechanisms: SRP and Fe exhibited similar seasonal patterns and opposite to that of NO3. We hypothesise that Fe oxyhydroxide reductive dissolution might be the cause of SRP release during the summer period, and that NO3 might act as a redox buffer, controlling the seasonality of SRP release. We conclude that LAMs may overestimate the contribution of P point sources, especially during the summer low-flow period, when eutrophication risk is maximal.

  9. Heat and mass release for some transient fuel source fires: A test report

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1986-10-01

    Nine fire tests using five different trash fuel source packages were conducted by Sandia National Laboratories. This report presents the findings of these tests. Data reported includes heat and mass release rates, total heat and mass release, plume temperatures, and average fuel heat of combustion. These tests were conducted as a part of the US Nuclear Regulatory Commission sponsored fire safety research program. Data from these tests were intended for use in nuclear power plant probabilistic risk assessment fire analyses. The results were also used as input to a fire test program at Sandia investigating the vulnerability of electrical control cabinets to fire. The fuel packages tested were chosen to be representative of small to moderately sized transient trash fuel sources of the type that would be found in a nuclear power plant. The highest fire intensity encountered during these tests was 145 kW. Plume temperatures did not exceed 820 0 C

  10. Source term analysis for a RCRA mixed waste disposal facility

    International Nuclear Information System (INIS)

    Jordan, D.L.; Blandford, T.N.; MacKinnon, R.J.

    1996-01-01

    A Monte Carlo transport scheme was used to estimate the source strength resulting from potential releases from a mixed waste disposal facility. Infiltration rates were estimated using the HELP code, and transport through the facility was modeled using the DUST code, linked to a Monte Carlo driver

  11. Analyses of postulated accidental releases of UF6 inside process buildings

    International Nuclear Information System (INIS)

    Oliveira Neto, Jose Messias de; Nunes, Beatriz Guimaraes; Dias, Cristiane

    2009-01-01

    Uranium Hexafluoride is a material used in the various processes which comprise the front end of the nuclear fuel cycle (conversion, enrichment and fuel fabrication). Confinement of UF 6 is a very important safety requirement since this material is highly reactive and presents safety hazards to humans. The present paper discusses the safety relevant aspects of accidental releases of UF 6 inside process confinement buildings. Postulated accidental scenarios are analyzed and their consequences evaluated. Implant releases rates are estimated using computer code predictions. A time dependent homogeneous compartment model is used to predict concentrations of UF 6 , hydrogen fluoride and uranyl fluoride inside a confinement building, as well as to evaluate source terms released to the atmosphere. These source terms can be used as input to atmospheric dispersion models to evaluate consequences to the environment. The results can also be used to define adequate protective measures for emergency situations. (author)

  12. A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

    International Nuclear Information System (INIS)

    Ang, M.L.; Grindon, E.; Dutton, L.M.C.; Garcia-Sedano, P.; Santamaria, C.S.; Centner, B.; Auglaire, M.; Routamo, T.; Outa, S.; Jokiniemi, J.; Gustavsson, V.; Wennerstrom, H.; Spanier, L.; Gren, M.; Boschiero, M-H; Droulas, J-L; Friederichs, H-G; Sonnenkalb, M.

    2001-01-01

    The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

  13. Radionuclide release rates from spent fuel for performance assessment modeling

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1994-01-01

    In a scenario of aqueous transport from a high-level radioactive waste repository, the concentration of radionuclides in water in contact with the waste constitutes the source term for transport models, and as such represents a fundamental component of all performance assessment models. Many laboratory experiments have been done to characterize release rates and understand processes influencing radionuclide release rates from irradiated nuclear fuel. Natural analogues of these waste forms have been studied to obtain information regarding the long-term stability of potential waste forms in complex natural systems. This information from diverse sources must be brought together to develop and defend methods used to define source terms for performance assessment models. In this manuscript examples of measures of radionuclide release rates from spent nuclear fuel or analogues of nuclear fuel are presented. Each example represents a very different approach to obtaining a numerical measure and each has its limitations. There is no way to obtain an unambiguous measure of this or any parameter used in performance assessment codes for evaluating the effects of processes operative over many millennia. The examples are intended to suggest by example that in the absence of the ability to evaluate accuracy and precision, consistency of a broadly based set of data can be used as circumstantial evidence to defend the choice of parameters used in performance assessments

  14. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

    Directory of Open Access Journals (Sweden)

    Sunarko

    2017-01-01

    Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

  15. A simplified radionuclide source term for total-system performance assessment

    International Nuclear Information System (INIS)

    Wilson, M.L.

    1991-11-01

    A parametric model for releases of radionuclides from spent-nuclear-fuel containers in a waste repository is presented. The model is appropriate for use in preliminary total-system performance assessments of the potential repository site at Yucca Mountain, Nevada; for this reason it is simpler than the models used for detailed studies of waste-package performance. Terms are included for releases from the spent fuel pellets, from the pellet/cladding gap and the grain boundaries within the fuel pellets, from the cladding of the fuel rods, and from the radioactive fuel-assembly parts. Multiple barriers are considered, including the waste container, the fuel-rod cladding, the thermal ''dry-out'', and the waste form itself. The basic formulas for release from a single fuel rod or container are extended to formulas for expected releases for the whole repository by using analytic expressions for probability distributions of some important parameters. 39 refs., 4 figs., 4 tabs

  16. Topical report on release scenario analysis of long-term management of high-level defense waste at the Hanford Site

    International Nuclear Information System (INIS)

    Wallace, R.W.; Landstrom, D.K.; Blair, S.C.; Howes, B.W.; Robkin, M.A.; Benson, G.L.; Reisenauer, A.E.; Walters, W.H.; Zimmerman, M.G.

    1980-11-01

    Potential release scenarios for the defense high-level waste (HLW) on the Hanford Site are presented. Presented in this report are the three components necessary for evaluating the various alternatives under consideration for long-term management of Hanford defense HLW: identification of scenarios and events which might directly or indirectly disrupt radionuclide containment barriers; geotransport calculations of waste migration through the site media; and consequence (dose) analyses based on groundwater and air pathways calculations. The scenarios described in this report provide the necessary parameters for radionuclide transport and consequence analysis. Scenarios are categorized as either bounding or nonbounding. Bounding scenarios consider worst case or what if situations where an actual and significant release of waste material to the environment would happen if the scenario were to occur. Bounding scenarios include both near-term and long-term scenarios. Near-term scenarios are events which occur at 100 years from 1990. Long term scenarios are potential events considered to occur at 1000 and 10,000 years from 1990. Nonbounding scenarios consider events which result in insignificant releases or no release at all to the environment. Three release mechanisms are described in this report: (1) direct exposure of waste to the biosphere by a defined sequence of events (scenario) such as human intrusion by drilling; (2) radionuclides contacting an unconfined aquifer through downward percolation of groundwater or a rising water table; and (3) cataclysmic or explosive release of radionuclides by such mechanisms as meteorite impact, fire and explosion, criticality, or seismic events. Scenarios in this report present ways in which these release mechanisms could occur at a waste management facility. The scenarios are applied to the two in-tank waste management alternatives: in-situ disposal and continued present action

  17. Flowsheets and source terms for radioactive waste projections

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1985-03-01

    Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF 6 conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables

  18. Identifying Patterns in the Weather of Europe for Source Term Estimation

    Science.gov (United States)

    Klampanos, Iraklis; Pappas, Charalambos; Andronopoulos, Spyros; Davvetas, Athanasios; Ikonomopoulos, Andreas; Karkaletsis, Vangelis

    2017-04-01

    During emergencies that involve the release of hazardous substances into the atmosphere the potential health effects on the human population and the environment are of primary concern. Such events have occurred in the past, most notably involving radioactive and toxic substances. Examples of radioactive release events include the Chernobyl accident in 1986, as well as the more recent Fukushima Daiichi accident in 2011. Often, the release of dangerous substances in the atmosphere is detected at locations different from the release origin. The objective of this work is the rapid estimation of such unknown sources shortly after the detection of dangerous substances in the atmosphere, with an initial focus on nuclear or radiological releases. Typically, after the detection of a radioactive substance in the atmosphere indicating the occurrence of an unknown release, the source location is estimated via inverse modelling. However, depending on factors such as the spatial resolution desired, traditional inverse modelling can be computationally time-consuming. This is especially true for cases where complex topography and weather conditions are involved and can therefore be problematic when timing is critical. Making use of machine learning techniques and the Big Data Europe platform1, our approach moves the bulk of the computation before any such event taking place, therefore allowing for rapid initial, albeit rougher, estimations regarding the source location. Our proposed approach is based on the automatic identification of weather patterns within the European continent. Identifying weather patterns has long been an active research field. Our case is differentiated by the fact that it focuses on plume dispersion patterns and these meteorological variables that affect dispersion the most. For a small set of recurrent weather patterns, we simulate hypothetical radioactive releases from a pre-known set of nuclear reactor locations and for different substance and temporal

  19. Evaluation of fuel release rate and mechanism tests under RBCB conditions

    International Nuclear Information System (INIS)

    Adamson, M.G.

    1981-09-01

    This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory

  20. Nitrifying aerobic granular sludge fermentation for releases of carbon source and phosphorus: The role of fermentation pH.

    Science.gov (United States)

    Zou, Jinte; Pan, Jiyang; He, Hangtian; Wu, Shuyun; Xiao, Naidong; Ni, Yongjiong; Li, Jun

    2018-07-01

    The effect of fermentation pH (uncontrolled, 4 and 10) on the releases of carbon source and phosphorus from nitrifying aerobic granular sludge (N-AGS) was investigated. Meanwhile, metal ion concentration and microbial community characterization were explored during N-AGS fermentation. The results indicated that N-AGS fermentation at pH 10 significantly promoted the releases of soluble chemical oxygen demand (SCOD) and total volatile fatty acids (TVFAs). However, SCOD and TVFA released from N-AGS were inhibited at pH 4. Moreover, acidic condition promoted phosphorus release (mainly apatite) from N-AGS during anaerobic fermentation. Nevertheless, alkaline condition failed to increase phosphorus concentration due to the formation of chemical-phosphate precipitates. Compared with the previously reported flocculent sludge fermentation, N-AGS fermentation released more SCOD and TVFAs, possibly due to the greater extracellular polymeric substances content and some hydrolytic-acidogenic bacteria in N-AGS. Therefore, N-AGS alkaline fermentation facilitated the carbon source recovery, while N-AGS acidic fermentation benefited the phosphorus recovery. Copyright © 2018. Published by Elsevier Ltd.

  1. Relation between source term and emergency planning for nuclear power plants

    International Nuclear Information System (INIS)

    Shi Zhongqi; Yang Ling

    1992-01-01

    Some background information of the severe accidents and source terms related to the nuclear power plant emergency planning are presented. The new source term information in NUREG-0956 and NUREG-1150, and possible changes in emergency planning requirements in U.S.A. are briefly provided. It is suggested that a principle is used in selecting source terms for establishing the emergency planning policy and a method is used in determining the Emergency Planning Zone (EPZ) size in China. Based on the research results of (1) EPZ size of PWR nuclear power plants being built in China, and (2) impact of reactor size and selected source terms on the EPZ size, it is concluded that the suggested principle and the method are suitable and feasible for PWR nuclear power plants in China

  2. Modeling a point-source release of 1,1,1-trichloroethane using EPA's SCREEN model

    International Nuclear Information System (INIS)

    Henriques, W.D.; Dixon, K.R.

    1994-01-01

    Using data from the Environmental Protection Agency's Toxic Release Inventory 1988 (EPA TRI88), pollutant concentration estimates were modeled for a point source air release of 1,1,1-trichloroethane at the Savannah River Plant located in Aiken, South Carolina. Estimates were calculating using the EPA's SCREEN model utilizing typical meteorological conditions to determine maximum impact of the plume under different mixing conditions for locations within 100 meters of the stack. Input data for the SCREEN model were then manipulated to simulate the impact of the release under urban conditions (for the purpose of assessing future landuse considerations) and under flare release options to determine if these parameters lessen or increase the probability of human or wildlife exposure to significant concentrations. The results were then compared to EPA reference concentrations (RfC) in order to assess the size of the buffer around the stack which may potentially have levels that exceed this level of safety

  3. A modified SILCS contraceptive diaphragm for long-term controlled release of the HIV microbicide dapivirine.

    Science.gov (United States)

    Major, Ian; Boyd, Peter; Kilbourne-Brook, Maggie; Saxon, Gene; Cohen, Jessica; Malcolm, R Karl

    2013-07-01

    There is considerable interest in developing new multipurpose prevention technologies to address women's reproductive health needs. This study describes an innovative barrier contraceptive device--based on the SILCS diaphragm--that also provides long-term controlled release of the lead candidate anti-HIV microbicide dapivirine. Diaphragm devices comprising various dapivirine-loaded polymer spring cores overmolded with a nonmedicated silicone elastomer sheath were fabricated by injection molding processes. In vitro release testing, thermal analysis and mechanical characterization were performed on the devices. A diaphragm device containing a polyoxymethylene spring core loaded with 10% w/w dapivirine provided continuous and controlled release of dapivirine over a 6-month period, with a mean in vitro daily release rate of 174 mcg/day. The mechanical properties of the new diaphragm were closely matched to the SILCS diaphragm. The study demonstrates proof of concept for a dapivirine-releasing diaphragm with daily release quantities potentially capable of preventing HIV transmission. In discontinuous clinical use, release of dapivirine may be readily extended over 1 or more years. Copyright © 2013 Elsevier Inc. All rights reserved.

  4. Source terms; isolation and radiological consequences of carbon-14 waste in the Swedish SFR repository

    International Nuclear Information System (INIS)

    Hesboel, R.; Puigdomenech, I.; Evans, S.

    1990-01-01

    The source term, isolation capacity, and long-term radiological exposure of 14 C from the Swedish underground repository for low and intermediate level waste (SFR) is assessed. The prospective amount of 14 C in the repository is assumed to be 5 TBq. Spent ion exchange resins will be the dominant source of 14 C. The pore water in the concrete repository is expected to maintain a pH of >10.5 for a period of at least 10 6 y. The cement matrix of the repository will retain most of the 14 CO 3 2- initially present. Bacterial production of CO 2 and CH 4 from degradation of ion-exchange resins and bitumen may contribute to 14 C release to the biosphere. However, CH 4 contributes only to a small extent to the overall carbon loss from freshwater ecosystems. The individual doses to local and regional individuals peaked with 5x10 -3 and regional individuals peaked with 5x10 -3 and 8x10 -4 μSv y -1 respectively at about 2.4x10 4 years. A total leakage of 8.4 GBq of 14 C from the repository will cause a total collective dose commitment of 1.1 manSv or 130 manSv TBq -1 . (authors)

  5. Rn-222 release to the environment: comparison between different granite sources

    International Nuclear Information System (INIS)

    Mamoon, M.; Kamal, S.M.

    2005-01-01

    In this work three different types of granites were studied, namely: pure granite, alkali granite and altered (hydrated) alkali granite. General radioactivity of the granites was studied along with the potential for 222 Rn emanation. The study indicated that altered alkali granite releases, relatively, the highest 222 Rn emanation to the surrounding air while alkali granite emits the more intense gamma radiation of the three granites. Hence, altered alkali granite can be used as a laboratory source for 222 Rn.

  6. Long-term stability study of Prussian blue-A quality assessment of water content and cyanide release.

    Science.gov (United States)

    Mohammad, A; Yang, Y; Khan, M A; Faustino, P J

    2015-02-01

    Prussian blue, ferric hexacyanoferrate is approved for (oral) treatment of internal contamination with radioisotopes of cesium or thallium. Cyanide makes up 35-40% of Prussian blue's molecular composition; thus, cyanide may be released during transit through the digestive tract under physiological pH conditions. The purpose of this study is to assess the long-term stability of Prussian blue drug products and active pharmaceutical ingredients and its impact on cyanide release. The study involves the determination and comparison of the loss in water content and cyanide released from Prussian blue under pH conditions that bracket human physiological exposure. Test samples of active pharmaceutical ingredient and drug product were stored for 10 years at ambient temperatures that mimic warehouse storage conditions. Water loss from Prussian blue was measured using thermogravimetric analysis. An in vitro physiological pH model that brackets gastric exposure and gastrointestinal transit was utilized for cyanide release. Prussian blue was incubated in situ at pH: 1.0, 5.0, and 7.0 @ 37°C for 1-24 h. Cyanide was measured using a validated colorimetric method by UV-Vis spectroscopy. Although the water content (quality attribute) of Prussian blue active pharmaceutical ingredient and drug product decreased by about 10.5% and 13.8%, respectively, since 2003, the cyanide release remained comparable. At pH of 7.0 for 24 h cyanide released from active pharmaceutical ingredient-1 was 21.33 ± 1.76 μg/g in 2004, and 28.45 ± 3.15 μg/g in 2013; cyanide released from drug product-1 was 21.89 ± 0.56 μg/g in 2004, and 27.31 ± 5.78 μg/g in 2013. At gastric pH of 1.0 and upper gastrointestinal pH of 5.0, the data for active pharmaceutical ingredients and drug products were also comparable in 2013. The cyanide release is still pH-dependent and follows the same trend as observed in 2003 with minimum release at pH of 5.0 and maximal release at pH of 1.0. In summary, this is the long-term

  7. Results for SEAFP-subtask A 10: Assessments of individual and collective doses to the public for routine and accidental releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.

    1995-04-01

    Dose calculations have been performed for accidental releases of tritium and activation products. Unit releases of 1 GBq per nuclide have been investigated for 31 fusion relevant nuclides. Additionally, unit releases of 1g of tritium and several activated materials have been studied. Under normal operation conditions, dose calculations have been performed for real source terms of tritium and activated materials. The individual dose values at the fence of the site (1 km) as well as the collective dose to the public (from 1 km to 100 km) have been obtained. As site specific parameters are still missing, different so called ''worst case'' release conditions have been applied. To have a first guess of the influence of the release duration on the dose to the Most Exposed Individual (MEI) in the vicinity of a reactor, different release durations, ranging from 1 hour up to 168 hours have been investigated, too. Finally, dose calculations have been performed for mobilisation source terms which take account of deposition and retention in the plant. This has been done for several RPM and APM source terms. The dose values of these final source terms seem to be less than every criteria to start emergency actions, however, some problems e.g. the behaviour of tritium in the plant, remain unsolved. (orig.)

  8. Mechanism for optimization of signal-to-noise ratio of dopamine release based on short-term bidirectional plasticity.

    Science.gov (United States)

    Da Cunha, Claudio; McKimm, Eric; Da Cunha, Rafael M; Boschen, Suelen L; Redgrave, Peter; Blaha, Charles D

    2017-07-15

    Repeated electrical stimulation of dopamine (dopamine) fibers can cause variable effects on further dopamine release; sometimes there are short-term decreases while in other cases short-term increases have been reported. Previous studies have failed to discover what factors determine in which way dopamine neurons will respond to repeated stimulation. The aim of the present study was therefore to investigate what determines the direction and magnitude of this particular form of short-term plasticity. Fixed potential amperometry was used to measure dopamine release in the nucleus accumbens in response to two trains of electrical pulses administered to the ventral tegmental area of anesthetized mice. When the pulse trains were of equal magnitude we found that low magnitude stimulation was associated with short-term suppression and high magnitude stimulation with short-term facilitation of dopamine release. Secondly, we found that the magnitude of the second pulse train was critical for determining the sign of the plasticity (suppression or facilitation), while the magnitude of the first pulse train determined the extent to which the response to the second train was suppressed or facilitated. This form of bidirectional plasticity might provide a mechanism to enhance signal-to-noise ratio of dopamine neurotransmission. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Initial assessment of MHTGR confinement releases

    International Nuclear Information System (INIS)

    Maneke, J.L.; Lanning, D.D.; Lidsky, L.M.

    1986-01-01

    Initial investigation of Modular High Temperature Gas-Cooled Reactor (MHTGR) designs suggest that source terms during postulated accidents will be considerably lower than Light Water Reactor (LWR) source term estimates. These lower postulated accident releases are not only a safety incentive, but also an economic incentive for the development of this reactor type. For example, it is hoped that a filtered confinement building, rather than a more expensive LWR-like containment building, would adequately protect the public from radiological releases. The ability of a confinement building to satisfy safety requirements for the MHTGR depends on several reactor parameters, such as fuel quality, reactor design, and the design of the reactor building. SCIMCA, a Simple Code for Initial MHTGR Confinement Assessment has been developed for preliminary MHTGR building requirement calculations. The code is capable of modeling a decay chain with a maximum of five regions. Phenomena such as fission product decay and buildup, natural deposition, building filtration, and intercompartmental transport are incorporated. SCIMCA models reduction mechanisms, such as dispersion and decay, occurring as radioactivity is transported through the environment. A subroutine for calculating doses at specified distances has also been included

  10. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    International Nuclear Information System (INIS)

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ''MELCOR Verification, Benchmarking, and Applications,'' whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR

  11. Radioactive releases of nuclear power plants: the code ASTEC

    International Nuclear Information System (INIS)

    Sdouz, G.; Pachole, M.

    1999-11-01

    In order to adopt potential countermeasures to protect the population during the course of an accident in a nuclear power plant a fast prediction of the radiation exposure is necessary. The basic input value for such a dispersion calculation is the source term, which is the description of the physical and chemical behavior of the released radioactive nuclides. Based on a source term data base a pilot system has been developed to determine a relevant source term and to generate the input file for the dispersion code TAMOS of the Zentralanstalt fuer Meteorologie und Geodynamik (ZAMG). This file can be sent directly as an attachment of e-mail to the TAMOS user for further processing. The source terms for 56 European nuclear power plant units are included in the pilot version of the code ASTEC (Austrian Source Term Estimation Code). The use of the system is demonstrated in an example based on an accident in the unit TEMELIN-1. In order to calculate typical core inventories for the data bank the international computer code OBIGEN 2.1 was installed and applied. The report has been completed with a discussion on the optimal data transfer. (author)

  12. Evaluation of JRC source term methodology using MAAP5 as a fast-running crisis tool for a BWR4 Mark I reactor

    International Nuclear Information System (INIS)

    Vela-García, M.; Simola, K.

    2016-01-01

    JRC participated in the OECD/NEA FASTRUN benchmark reviewing fast-running software tools to model fission product releases during accidents at nuclear power plants. The main goal of fast-running software tools is to foresee the accident progression, so that mitigating actions can be taken and the population can be adequately protected. Within the FASTRUN, JRC used the MAAP 4.0.8 code and developed a methodology to obtain the source term (as activity released per radioisotope) of PWR and BWR station black-out accident scenarios. The modifications made in the MAAP models were limited to a minimum number of important parameters. This aims at reproducing a crisis situation with a limited time to adapt a generic input deck. This paper presents further studies, where JRC analysed the FASTRUN BWR scenario using MAAP 5.0.2 that has the capability of calculating doses. A sensitivity study was performed with the MAAP 5.0.2 DOSE package deactivated, using the same methodology as in the case of MAAP 4.0.8 for source term calculation. The results were close to the reference LTSBO SOARCA case, independently of the methodology used. One of the benefits of using the MAAP code is the short runtime of the simulations.

  13. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  14. Low-level radioactive waste source term model development and testing: Topical report

    International Nuclear Information System (INIS)

    Sullivan, T.M.; Kempf, C.R.; Suen, C.J.; Mughabghab, S.M.

    1988-08-01

    The Low-Level Waste Source Term Evaluation Project has the objective to develop a system model capable of predicting radionuclide release rates from a shallow land burial facility. The previous topical report for this project discussed the framework and methodology for developing a system model and divided the problem into four compartments: water flow, container degradation, waste form leaching, and radionuclide transport. Each of these compartments is described by submodels which will be coupled into the system model. From February 1987 to March 1988, computer models have been selected to predict water flow (FEMWATER) and radionuclide transport (FEMWASTE) and separate models have been developed to predict pitting corrosion of steel containers and leaching from porous waste forms contained in corrodible containers. This report discusses each of the models in detail and presents results obtained from applying the models to shallow land burial trenches over a range of expected conditions. 68 refs., 34 figs., 14 tabs

  15. Experimental Physics and Industrial Control System (EPICS): Application source/release control for EPICS R3.11.6

    International Nuclear Information System (INIS)

    Zieman, B.; Kraimer, M.

    1994-01-01

    This manual describes a set of tools that can be used to develop software for EPICS based control systems. It provides the following features: Multiple applications; the entire system is composed of an arbitrary number of applications: Source/Release Control; all files created or modified by the applications developers can be put under sccs (a UNIX Source/Release control utility): Multiple Developers; it allows a number of applications developers to work separately during the development phase but combine their applications for system testing and for a production system; Makefiles: makefiles are provided to automatically rebuild various application components. For C and state notation programs, Imagefiles are provided

  16. Selective application of revised source terms to operating nuclear power plants

    International Nuclear Information System (INIS)

    Moon, Joo Hyun; Song, Jae Hyuk; Lee, Young Wook; Ko, Hyun Seok; Kang, Chang Sun

    2001-01-01

    More than 30 years later since 1962 when TID-14844 was promulgated, there has been big change of the US NRC's regulatory position in using accident source terms for radiological assessment following a design basis accident (DBA). To replace the instantaneous source terms of TID-14844, the time-dependent source terms of NUREG-1465 was published in 1995. In the meantime, the radiological acceptance criteria for reactor site evaluation in 10 CFR Part 100 were also revised. In particular, the concept of total effective dose equivalent has been incorporated in accordance with the radiation protection standards set forth in revised 10 CFR Part 20. Subsequently, the publication of Regulatory Guide 1.183 and the revision of Standard Review Plan 15.0.1 followed in 2000, which provided the licensee of operating nuclear power reactor with the acceptable guidance of applying the revised source term. The guidance allowed the holder of an operating license issued prior to January 10, 1997 to voluntarily revise the accident source terms used in the radiological consequence analyses of DBA. Regarding to its type of application, there suggested full and selective applications, Whether it is full or selective, based upon the scope and nature of associated plant modifications being proposed, the actual application of the revised source terms to an operating plant is expected to give a large impact on its facility design basis. Considering scope and cost of the analyses required for licensing, selective application is seemed to be more appealing to an licensee of the operating plant rather than full application. In this paper, hence, the selective application methodology is reviewed and is actally applied to the assessment of offsite radiological consequence following a LOCA at Ulchin Unit 3 and 4, in order to identify and analyze the potential impacts due to application of revised source terms and to assess the considerations taken in each application prior to its actual

  17. Robotic UV-Vis apparatus for long-term characterization of drug release from nanochannels

    International Nuclear Information System (INIS)

    Geninatti, T; Grattoni, A; Small, E

    2014-01-01

    Reliable monitoring of the kinetics of molecular release from drug delivery devices is crucial for their therapeutic success. Commercially available UV-Vis spectrophotometers provide reliable quantification of analyte concentrations directly correlated to the absorbance of fluids. However, they are not suitable for long-term measurements requiring high frequency of sampling from a large number of replicates and continuous fluid mixing, all of which are necessary for evaluation of drug delivery devices. To address this need, we developed a novel robotic apparatus serially connected to a commercial UV-Vis spectrophotometer. The robotic apparatus enables us to automatically and reliably acquire long-term data for up to 48 samples with high frequency of measurements and independent magnetic stirring. We equipped the robotic apparatus with independent connectors that allowed us to apply an electric potential to each sample for electrokinetic studies. The apparatus repeatability and accuracy was demonstrated in comparison to a commercial UV-Vis spectrophotometer. The system was successfully employed to characterize the diffusion kinetics of acetone and doxorubicin through nanochannel membranes (nDS) designed for long-term drug delivery. Dendritic fullerene 1 was used to show that the robotic apparatus routes the electric potential to nanochannel membranes enabling us to investigate the actively controlled release of molecules. Our results demonstrate that the robotic apparatus could widely broaden the range of applications of UV-Vis spectrophotometry, especially in the case of large sample processing and for long-term diffusive and electrokinetic studies in drug delivery. (technical design note)

  18. Development of a comprehensive source term model for the Subsurface Disposal Area at the Idaho National Engineering and Environmental Laboratory

    International Nuclear Information System (INIS)

    1997-01-01

    The first detailed comprehensive simulation study to evaluate fate and transport of wastes disposed in the Subsurface Disposal Area (SDA), at the Radioactive Waste Management Complex (RWMC), Idaho National Engineering and Environmental Laboratory (INEEL) has recently been conducted. One of the most crucial parts of this modeling was the source term or release model. The current study used information collected over the last five years defining contaminant specific information including: the amount disposed, the waste form (physical and chemical properties) and the type of container used for each contaminant disposed. This information was used to simulate the release of contaminants disposed in the shallow subsurface at the SDA. The DUST-MS model was used to simulate the release. Modifications were made to allow the yearly disposal information to be incorporated. The modeling includes unique container and release rate information for each of the 42 years of disposal. The results from this simulation effort are used for both a groundwater and a biotic uptake evaluation. As part of this modeling exercise, inadequacies in the available data relating to the release of contaminants have been identified. The results from this modeling study have been used to guide additional data collection activities at the SDA for purposes of increasing confidence in the appropriateness of model predictions

  19. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 x 10 -3 person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 x 10 -2 person-Sv/Pbq (2.4 x 10 -4 person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river

  20. Performance and first results of fission product release and transport provided by the VERDON facility

    Energy Technology Data Exchange (ETDEWEB)

    Gallais-During, A., E-mail: annelise.gallais-during@cea.fr [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Bonnin, J.; Malgouyres, P.-P. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Morin, S. [IRSN, F-13108 Saint-Paul-lez-Durance (France); Bernard, S.; Gleizes, B.; Pontillon, Y.; Hanus, E.; Ducros, G. [CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-01

    Highlights: • A new facility to perform experimental LWR severe accidents sequences is evaluated. • In the furnace a fuel sample is heated up to 2600 °C under a controlled gas atmosphere. • Innovative thermal gradient tubes are used to study fission product transport. • The new VERDON facility shows an excellent consistency with results from VERCORS. • Fission product re-vapourization results confirm the correct functioning of the gradient tubes. - Abstract: One of the most important areas of research concerning a hypothetical severe accident in a light water reactor (LWR) is determining the source term, i.e. quantifying the nature, release kinetics and global released fraction of the fission products (FPs) and other radioactive materials. In line with the former VERCORS programme to improve source term estimates, the new VERDON laboratory has recently been implemented at the CEA Cadarache Centre in the LECA-STAR facility. The present paper deals with the evaluation of the experimental equipment of this new VERDON laboratory (furnace, release and transport loops) and demonstrates its capability to perform experimental sequences representative of LWR severe accidents and to supply the databases necessary for source term assessments and FP behaviour modelling.

  1. Effects of Operating Conditions on Gas Release Thermal ...

    African Journals Online (AJOL)

    The gas release rates and the flame length of the potential jet fires were initially estimated using Simplex Source Term Models which pay limited attention to operating conditions. Finally a more detailed follow-up study, accounting for a range of practical factors was conducted. A number of useful risk management metrics ...

  2. Directional Unfolded Source Term (DUST) for Compton Cameras.

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Dean J.; Mitchell, Dean J.; Horne, Steven M.; O' Brien, Sean; Thoreson, Gregory G

    2018-03-01

    A Directional Unfolded Source Term (DUST) algorithm was developed to enable improved spectral analysis capabilities using data collected by Compton cameras. Achieving this objective required modification of the detector response function in the Gamma Detector Response and Analysis Software (GADRAS). Experimental data that were collected in support of this work include measurements of calibration sources at a range of separation distances and cylindrical depleted uranium castings.

  3. Environmental radiation safety source term evaluation program

    International Nuclear Information System (INIS)

    Moss, O.R.; Filipy, R.E.; Cannon, W.C.; Craig, D.K.

    1977-04-01

    Plutonium-238 is currently used in the form of a pure refractory oxide as a power source on a number of space vehicles that have already been or will be launched during the next few years. Although the sources are designed and built to withstand re-entry into the earth's atmosphere and impact with the earth's surface without releasing any plutonium, the possibility of such an event can never be absolutely excluded. Three separate tasks were undertaken in this study. The interactions between soils and 238 PuO 2 aerosols which might be created in a space launch about environment were examined. Aging of the plutonium-soil mixture under a humid atmosphere showed a trend toward the slow coagulation of two dilute aerosols. Studies on marine animals were conducted to assess the response of 238 PuO 2 pellets to conditions found 60 feet below the ocean surface. Ultrafilterability studies measured the solubility of 238 PuO 2 as a function of time, temperature, suspension concentration and molality of solvent

  4. Development of source term evaluation method for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Keon Jae; Cheong, Jae Hak; Park, Jin Baek; Kim, Guk Gee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-10-15

    This project had investigate several design features of radioactive waste processing system and method to predict nuclide concentration at primary coolant basic concept of next generation reactor and safety goals at the former phase. In this project several prediction methods of source term are evaluated conglomerately and detailed contents of this project are : model evaluation of nuclide concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant(NPP), investigation of prediction parameter of source term evaluation, basic parameter of PWR, operational parameter, respectively, radionuclide removal system and adjustment values of reference NPP, suggestion of source term prediction method of next generation NPP.

  5. Source term estimation for small sized HTRs

    International Nuclear Information System (INIS)

    Moormann, R.

    1992-08-01

    Accidents which have to be considered are core heat-up, reactivity transients, water of air ingress and primary circuit depressurization. The main effort of this paper belongs to water/air ingress and depressurization, which requires consideration of fission product plateout under normal operation conditions; for the latter it is clearly shown, that absorption (penetration) mechanisms are much less important than assumed sometimes in the past. Source term estimation procedures for core heat-up events are shortly reviewed; reactivity transients are apparently covered by them. Besides a general literature survey including identification of areas with insufficient knowledge this paper contains some estimations on the thermomechanical behaviour of fission products in water in air ingress accidents. Typical source term examples are also presented. In an appendix, evaluations of the AVR experiments VAMPYR-I and -II with respect to plateout and fission product filter efficiency are outlined and used for a validation step of the new plateout code SPATRA. (orig.)

  6. Long-Term Effectiveness and Safety of Dexmethylphenidate Extended-Release Capsules in Adult ADHD

    Science.gov (United States)

    Adler, Lenard A.; Spencer, Thomas; McGough, James J.; Jiang, Hai; Muniz, Rafael

    2009-01-01

    Objective: This study evaluates dexmethylphenidate extended release (d-MPH-ER) in adults with ADHD. Method: Following a 5-week, randomized, controlled, fixed-dose study of d-MPH-ER 20 to 40 mg/d, 170 adults entered a 6-month open-label extension (OLE) to assess long-term safety, with flexible dosing of 20 to 40 mg/d. Exploratory effectiveness…

  7. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  8. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

  9. Inverse modelling of radionuclide release rates using gamma dose rate observations

    Science.gov (United States)

    Hamburger, Thomas; Evangeliou, Nikolaos; Stohl, Andreas; von Haustein, Christoph; Thummerer, Severin; Wallner, Christian

    2015-04-01

    Severe accidents in nuclear power plants such as the historical accident in Chernobyl 1986 or the more recent disaster in the Fukushima Dai-ichi nuclear power plant in 2011 have drastic impacts on the population and environment. Observations and dispersion modelling of the released radionuclides help to assess the regional impact of such nuclear accidents. Modelling the increase of regional radionuclide activity concentrations, which results from nuclear accidents, underlies a multiplicity of uncertainties. One of the most significant uncertainties is the estimation of the source term. That is, the time dependent quantification of the released spectrum of radionuclides during the course of the nuclear accident. The quantification of the source term may either remain uncertain (e.g. Chernobyl, Devell et al., 1995) or rely on estimates given by the operators of the nuclear power plant. Precise measurements are mostly missing due to practical limitations during the accident. The release rates of radionuclides at the accident site can be estimated using inverse modelling (Davoine and Bocquet, 2007). The accuracy of the method depends amongst others on the availability, reliability and the resolution in time and space of the used observations. Radionuclide activity concentrations are observed on a relatively sparse grid and the temporal resolution of available data may be low within the order of hours or a day. Gamma dose rates, on the other hand, are observed routinely on a much denser grid and higher temporal resolution and provide therefore a wider basis for inverse modelling (Saunier et al., 2013). We present a new inversion approach, which combines an atmospheric dispersion model and observations of radionuclide activity concentrations and gamma dose rates to obtain the source term of radionuclides. We use the Lagrangian particle dispersion model FLEXPART (Stohl et al., 1998; Stohl et al., 2005) to model the atmospheric transport of the released radionuclides. The

  10. Development of source term evaluation method for Korean Next Generation Reactor(III)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geon Jae; Park, Jin Baek; Lee, Yeong Il; Song, Min Cheonl; Lee, Ho Jin [Korea Advanced Institue of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    This project had investigated irradiation characteristics of MOX fuel method to predict nuclide concentration at primary and secondary coolant using a core containing 100% of all MOX fuel and development of source term evaluation tool. In this study, several prediction methods of source term are evaluated. Detailed contents of this project are : an evaluation of model for nuclear concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant using purely MOX fuel, suggestion of source term prediction method of NPP with a core using MOX fuel.

  11. Reuse of conditional released materials from decommissioning; a review of approaches and scenarios with long-term constructions - 59149

    International Nuclear Information System (INIS)

    Daniska, Vladimir; Pritrsky, Jozef; Ondra, Frantisek; Zachar, Matej; Necas, Vladimir

    2012-01-01

    Paper presents the overall scope and actual results of the project for evaluation of representative scenarios for reuse of conditionally released materials from decommissioning. Aim of the project is to evaluate the possibilities of reuse of conditionally released steels and concrete in technical constructions which guarantee the long-term preservation of design properties over periods of 50-100 years. Interaction of conditionally released materials with public is limited and predictable due to design and purpose of selected constructions and due to fact that in many scenarios these materials are embedded in non-radioactive materials such as bars in reinforced concrete. Worker's scenarios for preparation, operation and maintenance of these constructions are analysed in detail including the manufacturing of elements for these constructions. Project aims to evaluate the scenarios of reuse of conditionally released materials in a complex way in order to develop the data for designers of scenarios and to evaluate the volumes of conditionally released materials based on facility (to be decommissioned) inventory data. The long-term constructions considered are bridges, tunnels, roads, railway constructions, industrial buildings, power industry equipment and others. Evaluation covers following areas: - Analysis of activities for manufacturing of reinforcement bars, rolled steel sheets and other steel elements and analysis of activities for construction of evaluated scenarios in order to evaluate the external exposure of professionals performing these activities; - Analysis of external exposure of professionals involved in operation and maintenance of the long-term constructions; analysis of external exposure of public groups which are exposed to evaluated constructions; - Analysis of internal exposure of public groups from the radionuclides released from the evaluated scenarios based on models for migration of radionuclides from the long-term constructions to critical

  12. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  13. Sources of radioactive waste from light-water reactors and their physical and chemical properties

    International Nuclear Information System (INIS)

    Bell, M.J.; Collins, J.T.

    1979-01-01

    The general physical and chemical properties of waste streams in light-water reactors (LWRs) are described. The principal mechanisms for release and the release pathways to the environment are discussed. The calculation of liquid and gaseous source terms using one of the available models is presented. These calculated releases are compared with observed releases from operating LWRs. The computerized mathematical model used is the GALE Code which is the Nuclear Regulatory Commission (NRC) staff's model for calculating source terms for effluents from LWRs (USNRC76a, USNRC76b). Programs currently being conducted at operating reactors by the NRC, Electric Power Research Institute, and various utilities to better define the characteristics of waste streams and the performance of radwaste process equipment are described

  14. The influence of short-term concentration peaks on exposure risks in the vicinity of an episodic release of hydrogen sulphide

    International Nuclear Information System (INIS)

    1981-01-01

    In order to propose a methodology by which the influence of short-term concentration peaks on exposure risks could be estimated in the vicinity of an atmospheric release of hydrogen sulphide (H 2 S), an extensive and up-to-date review of H 2 S toxicity was conducted, with emphasis on acute and sub-acute poisoning. The literature on animal studies and cases of human exposure were used to derive a lethal dose relationship(concentration-exposure time) appropriate for the general population. A statistical model was developed which calculates the probability of short-term concentrations exceeding the lethal level given the downwind range from the release, the short-term averaging time of interest, the long-term average concentration, and metereological and terrain conditions . Results were obtained for passive releases of H 2 S under a range of hypothetical conditions. Interpretation of these results is given in terms of the overall probability of lethal exposure during a 30-minute episode. The likely influence of heavy water plant gas dispersion systems is also addressed. (author)

  15. Effect of a long-term release of plutonium and americium into an estuarine-coastal sea ecosystem. 1. Development of an assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Murray, C N; Avogadro, A [Commission of the European Communities, Ispra (Italy). Joint Research Centre

    1979-05-01

    A method is developed for the assessment of the distribution and the associated hazard due to a long-term release of long-lived alpha-emitting radionuclides into surface waters of a marine ecosystem. The methodology is designed to identify various environmental compartments and the processes occurring within them, which are of importance in affecting the behaviour and thus the distribution of actinide elements in such systems. The compartment model system (box model) chosen is applied to an imaginary coastal area. Various processes in the environmental compartments are considered separately and then assembled to show their combined interactions. Using the concept of critical nuclide-pathway-group analysis, an attempt has been made to define a number of the most important pathways by which actinides released into the aquatic environment could return to man, and especially those related to the exploitation of aquatic food resources. The concentration levels for the example considered produce rather low dose rates to man of less than 3% of the maximum permissible intake. The low dose levels strongly depend upon the concentration factors of the various biological species, as well as upon source-term activities. The concentration factors used for the biological transfer of actinides relate to the water activity only. In the case where highly radioactive sediments or sedimentary-associated material were closely involved in the uptake pathway, actinide transfer to man could become more relevant. This study shows that sedimentation and bottom sediment absorption represent the major reconcentration processes for actinides released into surface waters.

  16. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  17. Overview of major hazards. Part 2: Source term; dispersion; combustion; blast, missiles, venting; fire; radiation; runaway reactions; toxic substances; dust explosions

    Science.gov (United States)

    Vilain, J.

    Approaches to major hazard assessment and prediction are reviewed. Source term: (phenomenology/modeling of release, influence on early stages of dispersion); dispersion (atmospheric advection, diffusion and deposition, emphasis on dense/cold gases); combustion (flammable clouds and mists covering flash fires, deflagration, transition to detonation; mostly unconfined/partly confined situations); blast formation, propagation, interaction with structures; catastrophic fires (pool fires, torches and fireballs; highly reactive substances) runaway reactions; features of more general interest; toxic substances, excluding toxicology; and dust explosions (phenomenology and protective measures) are discussed.

  18. Source-term development for a contaminant plume for use by multimedia risk assessment models

    International Nuclear Information System (INIS)

    Whelan, Gene; McDonald, John P.; Taira, Randal Y.; Gnanapragasam, Emmanuel K.; Yu, Charley; Lew, Christine S.; Mills, William B.

    1999-01-01

    Multimedia modelers from the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE) are collaborating to conduct a comprehensive and quantitative benchmarking analysis of four intermedia models: DOE's Multimedia Environmental Pollutant Assessment System (MEPAS), EPA's MMSOILS, EPA's PRESTO, and DOE's RESidual RADioactivity (RESRAD). These models represent typical analytically, semi-analytically, and empirically based tools that are utilized in human risk and endangerment assessments for use at installations containing radioactive and/or hazardous contaminants. Although the benchmarking exercise traditionally emphasizes the application and comparison of these models, the establishment of a Conceptual Site Model (CSM) should be viewed with equal importance. This paper reviews an approach for developing a CSM of an existing, real-world, Sr-90 plume at DOE's Hanford installation in Richland, Washington, for use in a multimedia-based benchmarking exercise bet ween MEPAS, MMSOILS, PRESTO, and RESRAD. In an unconventional move for analytically based modeling, the benchmarking exercise will begin with the plume as the source of contamination. The source and release mechanism are developed and described within the context of performing a preliminary risk assessment utilizing these analytical models. By beginning with the plume as the source term, this paper reviews a typical process and procedure an analyst would follow in developing a CSM for use in a preliminary assessment using this class of analytical tool

  19. In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

  20. A Study on Improvement of Algorithm for Source Term Evaluation

    International Nuclear Information System (INIS)

    Park, Jeong Ho; Park, Do Hyung; Lee, Jae Hee

    2010-03-01

    The program developed by KAERI for source term assessment of radwastes from the advanced nuclear fuel cycle consists of spent fuel database analysis module, spent fuel arising projection module, and automatic characterization module for radwastes from pyroprocess. To improve the algorithm adopted the developed program, following items were carried out: - development of an algorithm to decrease analysis time for spent fuel database - development of setup routine for a analysis procedure - improvement of interface for spent fuel arising projection module - optimization of data management algorithm needed for massive calculation to estimate source terms of radwastes from advanced fuel cycle The program developed through this study has a capability to perform source term estimation although several spent fuel assemblies with different fuel design, initial enrichment, irradiation history, discharge burnup, and cooling time are processed at the same time in the pyroprocess. It is expected that this program will be very useful for the design of unit process of pyroprocess and disposal system

  1. Radionuclide release rate inversion of nuclear accidents in nuclear facility based on Kalman filter

    International Nuclear Information System (INIS)

    Tang Xiuhuan; Bao Lihong; Li Hua; Wan Junsheng

    2014-01-01

    The rapidly and continually back-calculating source term is important for nuclear emergency response. The Gaussian multi-puff atmospheric dispersion model was used to produce regional environment monitoring data virtually, and then a Kalman filter was designed to inverse radionuclide release rate of nuclear accidents in nuclear facility and the release rate tracking in real time was achieved. The results show that the Kalman filter combined with Gaussian multi-puff atmospheric dispersion model can successfully track the virtually stable, linear or nonlinear release rate after being iterated about 10 times. The standard error of inversion results increases with the true value. Meanwhile extended Kalman filter cannot inverse the height parameter of accident release as interceptive error is too large to converge. Kalman filter constructed from environment monitoring data and Gaussian multi-puff atmospheric dispersion model can be applied to source inversion in nuclear accident which is characterized by static height and position, short and continual release in nuclear facility. Hence it turns out to be an alternative source inversion method in nuclear emergency response. (authors)

  2. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  3. Reassessment of the technical bases for estimating source terms. Draft report for comment

    International Nuclear Information System (INIS)

    Silberberg, M.; Mitchell, J.A.; Meyer, R.O.; Pasedag, W.F.; Ryder, C.P.; Peabody, C.A.; Jankowski, M.W.

    1985-07-01

    NUREG-0956 describes the NRC staff and contractor efforts to reassess and update the agency's analytical procedures for estimating accident source terms for nuclear power plants. The effort included development of a new source term analytical procedure - a set of computer codes - that is intended to replace the methodology of the Reactor Safety Study (WASH-1400) and to be used in reassessing the use of TID-14844 assumptions (10 CFR 100). NUREG-0956 describes the development of these codes, the demonstration of the codes to calculate source terms for specific cases, the peer review of this work, some perspectives on the overall impact of new source terms on plant risks, the plans for related research projects, and the conclusions and recommendations resulting from the effort

  4. Mineral sources and transport pathways for arsenic release in a coastal watershed, USA

    Science.gov (United States)

    Foley, Nora K.; Ayuso, Robert A.

    2008-01-01

    Metasedimentary bedrock of coastal Maine contains a diverse suite of As-bearing minerals that act as significant sources of elements found in ground and surface waters in the region. Arsenic sources in the Penobscot Formation include, in order of decreasing As content by weight: löllingite and realgar (c.70%), arsenopyrite, cobaltite, glaucodot, and gersdorffite (in the range of 34–45%), arsenian pyrite (Formation, the relative stability of primary As-bearing minerals follows a pattern where the most commonly observed highly altered minerals are pyrrhotite, realgar, niccolite, löllingite > glaucodot, arsenopyrite-cobaltian > arsenopyrite, cobaltite, gersdorffite, fine-grained pyrite, Ni-pyrite > coarse-grained pyrite. Reactions illustrate that oxidation of Fe-As disulphide group and As-sulphide minerals is the primary release process for As. Liberation of As by carbonation of realgar and orpiment in contact with high-pH groundwaters may contribute locally to elevated contents of As in groundwater, especially where As is decoupled from Fe. Released metals are sequestered in secondary minerals by sorption or by incorporation in crystal structures. Secondary minerals acting as intermediate As reservoirs include claudetite (c.75%), orpiment (61%), scorodite (c. 45%), secondary arsenopyrite (c. 46%), goethite (minerals. Reductive dissolution of Fe-oxide minerals may govern the ultimate release of iron and arsenic – especially As(V) – to groundwater; however, dissolution of claudetite (arsenic trioxide) may directly contribute As(III). Processes thought to explain the release of As from minerals in bedrock include oxidation of arsenian pyrite or arsenopyrite, or carbonation of As-sulphides, and most models based on these generally rely on discrete minerals or on a fairly limited series of minerals. In contrast, in the Penobscot Formation and other metasedimentary rocks of coastal Maine, oxidation of As-bearing Fe-cobalt-nickel-sulphide minerals, dissolution (by

  5. Environmental contamination from a ground-level release of fission products

    International Nuclear Information System (INIS)

    Stupka, R.C.; Kephart, G.S.; Rittmann, P.D.

    1986-08-01

    On January 11, 1985, a ground-level release of fission products, primarily 90 Sr, occurred at the Hanford Site in southeastern Washington State. The release was detected during routine surveys and the majority of the contamination was confined to the immediate area where the release occurred. Response to the incident was complicated by a strong inversion that resulted in a buildup of 222 Rn daughter products on environmental air samples and outdoor surfaces. The cause of the release appears to have been the operation of a transfer jet that inadvertently pressurized an unblanked line leading to the 241-C-151 Diversion Box. A buildup of pressure inside the diversion box forced contaminated air through gaps in the diversion box cover blocks resulting in an unmonitored, short duration release to the environment. The source term was estimated using data obtained from environmental air samplers. The ground deposition speed was calculated using the integrated exposure (air samples) and surface contamination levels obtained from recently fallen snow. The total release was estimated to be 1.4 Ci 90 Sr and 0.02 Ci 137 Cs. Based on this source term, the maximum 50-yr dose commitment to onsite pesonnel was 50 mrem whole body and 600 mrem bone. No detectable internal deposition occurred during the incident and corrective action which followed; this was probably due to several factors: (1) prompt detection of the release; (2) localized contamination control; (3) excellent personnel protection practices; and (4) the protection offered by building ventilation systems. The theoretical maximum offsite individual would receive a potential 1-yr dose commitment of 0.01 mrem whole body and 0.2 mrem bone from this incident. The potential 50-yr dose commitment would be 0.13 mrem whole body and 2.0 mrem bone. In actuality, neither onsite or offsite individuals would be expected to receive even these small dose commitments

  6. Pennsylvania Source Term Tracking System

    International Nuclear Information System (INIS)

    1992-08-01

    The Pennsylvania Source Term Tracking System tabulates surveys received from radioactive waste generators in the Commonwealth of radioactive waste is collected each quarter from generators using the Low-Level Radioactive Waste Management Quarterly Report Form (hereafter called the survey) and then entered into the tracking system data base. This personal computer-based tracking system can generate 12 types of tracking reports. The first four sections of this reference manual supply complete instructions for installing and setting up the tracking system on a PC. Section 5 presents instructions for entering quarterly survey data, and Section 6 discusses generating reports. The appendix includes samples of each report

  7. Effect of Fuel Structure Materials on Radiation Source Term in Reactor Core Meltdown

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Ha, Kwang Soon

    2014-01-01

    The fission product (Radiation Source) releases from the reactor core into the containment is obligatorily evaluated to guarantee the safety of Nuclear Power Plant (NPP) under the hypothetical accident involving a core meltdown. The initial core inventory is used as a starting point of all radiological consequences and effects on the subsequent results of accident assessment. Hence, a proper evaluation for the inventory can be regarded as one of the most important part over the entire procedure of accident analysis. The inventory of fission products is typically evaluated on the basis of the uranium material (e.g., UO2 and USi2) loaded in nuclear fuel assembly, except for the structure materials such as the end fittings, grids, and some kinds of springs. However, the structure materials are continually activated by the neutrons generated from the nuclear fission, and some nuclides of them (e.g., 14 C and 60 Co) can significantly influence on accident assessment. During the severe core accident, the structure components can be also melted with the melting points of temperature relatively lower than uranium material. A series of the calculation were performed by using ORIGEN-S module in SCALE 6.1 package code system. The total activity in each part of structure materials was specifically analyzed from these calculations. The fission product inventory is generally evaluated based on the uranium materials of fuel only, even though the structure components of the assembly are continually activated by the neutrons generated from the nuclear fission. In this study, the activation calculation of the fuel structure materials was performed for the initial source term assessment in the accident of reactor core meltdown. As a result, the lower end fitting and the upper plenum greatly contribute to the total activity except for the cladding material. The nuclides of 56 Mn, '5 1 Cr, 55 Fe, 58 Co, 54 Mn, and 60 Co are analyzed to mainly effect on the activity. This result

  8. Metal lost and found: dissipative uses and releases of copper in the United States 1975-2000.

    Science.gov (United States)

    Lifset, Reid J; Eckelman, Matthew J; Harper, E M; Hausfather, Zeke; Urbina, Gonzalo

    2012-02-15

    Metals are used in a variety of ways, many of which lead to dissipative releases to the environment. Such releases are relevant from both a resource use and an environmental impact perspective. We present a historical analysis of copper dissipative releases in the United States from 1975 to 2000. We situate all dissipative releases in copper's life cycle and introduce a conceptual framework by which copper dissipative releases may be categorized in terms of intentionality of use and release. We interpret our results in the context of larger trends in production and consumption and government policies that have served as drivers of intentional copper releases from the relevant sources. Intentional copper releases are found to be both significant in quantity and highly variable. In 1975, for example, the largest source of intentional releases was from the application of copper-based pesticides, and this decreased more than 50% over the next 25 years; all other sources of intentional releases increased during that period. Overall, intentional copper releases decreased by approximately 15% from 1975 to 2000. Intentional uses that are unintentionally released such as copper from roofing, increased by the same percentage. Trace contaminant sources such as fossil fuel combustion, i.e., sources where both the use and the release are unintended, increased by nearly 50%. Intentional dissipative uses are equivalent to 60% of unintentional copper dissipative releases and more than five times that from trace sources. Dissipative copper releases are revealed to be modest when compared to bulk copper flows in the economy, and we introduce a metric, the dissipation index, which may be considered an economy-wide measure of resource efficiency for a particular substance. We assess the importance of dissipative releases in the calculation of recycling rates, concluding that the inclusion of dissipation in recycling rate calculations has a small, but discernible, influence, and should

  9. Release fraction of PWR after severe accidents. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    Aziz, M; El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Fission fragments and gases are emitted after accidents as a result of core meltdown and core concrete interactions. These aerosols are transported and fill the reactor containment. With increasing the pressure above pressure design bases, a failure of containment may occur and subsequently these aerosols will release into the external environment leading to a source term of radioactivity that affects the safety of workers and public. The amount of aerosol which escapes to the environment can be described by the release fraction which is defined as the total accumulated aerosol which initially enters the containment. The factors that affect the release fraction is studied, and the aerosol dynamics equation is used to model the release of aerosol to the outside atmosphere. These factors are containment pressure, failure time,break area, the size of aerosol particle. It found that early failure time and higher pressure increase the release fraction, also the release faction is affected by the area and the aerosol particle size. 7 figs., 2 tabs.

  10. Release fraction of PWR after severe accidents. Vol. 4

    International Nuclear Information System (INIS)

    Aziz, M.; El-Messeiry, A.M.

    1996-01-01

    Fission fragments and gases are emitted after accidents as a result of core meltdown and core concrete interactions. These aerosols are transported and fill the reactor containment. With increasing the pressure above pressure design bases, a failure of containment may occur and subsequently these aerosols will release into the external environment leading to a source term of radioactivity that affects the safety of workers and public. The amount of aerosol which escapes to the environment can be described by the release fraction which is defined as the total accumulated aerosol which initially enters the containment. The factors that affect the release fraction is studied, and the aerosol dynamics equation is used to model the release of aerosol to the outside atmosphere. These factors are containment pressure, failure time,break area, the size of aerosol particle. It found that early failure time and higher pressure increase the release fraction, also the release faction is affected by the area and the aerosol particle size. 7 figs., 2 tabs

  11. Source terms: an investigation of uncertainties, magnitudes, and recommendations for research. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Levine, S.; Kaiser, G. D.; Arcieri, W. C.; Firstenberg, H.; Fulford, P. J.; Lam, P. S.; Ritzman, R. L.; Schmidt, E. R.

    1982-03-01

    The purpose of this document is to assess the state of knowledge and expert opinions that exist about fission product source terms from potential nuclear power plant accidents. This is so that recommendations can be made for research and analyses which have the potential to reduce the uncertainties in these estimated source terms and to derive improved methods for predicting their magnitudes. The main reasons for writing this report are to indicate the major uncertainties involved in defining realistic source terms that could arise from severe reactor accidents, to determine which factors would have the most significant impact on public risks and emergency planning, and to suggest research and analyses that could result in the reduction of these uncertainties. Source terms used in the conventional consequence calculations in the licensing process are not explicitly addressed.

  12. Estimated airborne release of plutonium from Atomics International's Nuclear Materials Development Facility in the Santa Susana site, California, as a result of postulated damage from severe wind and earthquake hazard

    International Nuclear Information System (INIS)

    Mishima, J.; Ayer, J.E.

    1981-09-01

    The potential mass of airborne releases of plutonium (source term) that could result from wind and seismic damage is estimated for the Atomics International Company's Nuclear Materials Development Facility (NMDF) at the Santa Susana site in California. The postulated source terms will be useful as the basis for estimating the potential dose to the maximum exposed individual by inhalation and to the total population living within a prescribed radius of the site. The respirable fraction of airborne particles is thus the principal concern. The estimated source terms are based on the damage ratio, and the potential airborne releases if all enclosures suffer particular levels of damage. In an attempt to provide a realistic range of potential source terms that include most of the normal processing conditions, a best estimate bounded by upper and lower limits is provided. The range of source terms is calculated by combining a high best estimate and a low damage ratio, based on a fraction of enclosures suffering crush or perforation, with the airborne release from enclosures based upon an upper limit, average, and lower limit inventory of dispersible materials at risk. Two throughput levels are considered. The factors used to evaluate the fractional airborne release of materials and the exchange rates between enclosed and exterior atmospheres are discussed. The postulated damage and source terms are discussed for wind and earthquake hazard scenarios in order of their increasing severity

  13. Alternate source term models for Yucca Mountain performance assessment based on natural analog data and secondary mineral solubility

    International Nuclear Information System (INIS)

    Murphy, W.M.; Codell, R.B.

    1999-01-01

    Performance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Pena Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Pena Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg/y, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Pena Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses

  14. Alternate source term models for Yucca Mountain performance assessment based on natural analog data and secondary mineral solubility

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, W.M.; Codell, R.B.

    1999-07-01

    Performance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Pena Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Pena Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg/y, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Pena Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.

  15. EDF source term reduction project main outcomes and further developments

    International Nuclear Information System (INIS)

    Ranchoux, Gilles; Bonnefon, Julien; Benfarah, Moez; Wintergerst Matthieu; Gressier, Frederic; Leclercq, Stephanie

    2012-09-01

    The dose reduction is a strategic purpose for EDF in link with the stakes of, nuclear acceptability, respect of regulation and productivity gains. This consists not only in improving the reactor shutdown organization (time spent in control area, biological shielding,...) but also in improving the radiological state of the unit and the efficiency of the source term reduction operations. Since 2003, EDF has been running an innovative project called 'Source Term Reduction' federating the different EDF research and engineering centers in order to: - participate to the long term view about Radiological Protection issues (international feedback analyses), - develop contamination prediction tools (OSCAR software) suitable for the industrial needs (operating units and EPR design), - develop scientific models useful for the understanding of contamination mechanisms to support the strategic decision processes, - carry on with updating and analyzing of contamination measurements feedback in corrosion products (EMECC and CZT campaigns), - carry on with the operational support at short or middle term by optimizing startup and shutdown processes, pre-oxidation or and by improving purification efficiency or material characteristics. This paper will show in a first part the main 2011 results in occupational exposure (collective and individual dose, RCS index...). In a second part, an overview of the main EDF outcomes of the last 3 years in the field of source term reduction will be presented. Future developments extended to contamination issues in EDF NPPs will be also pointed out in this paper. (authors)

  16. Development of DUST: A computer code that calculates release rates from a LLW disposal unit

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1992-01-01

    Performance assessment of a Low-Level Waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the disposal unit source term). The major physical processes that influence the source term are water flow, container degradation, waste form leaching, and radionuclide transport. A computer code, DUST (Disposal Unit Source Term) has been developed which incorporates these processes in a unified manner. The DUST code improves upon existing codes as it has the capability to model multiple container failure times, multiple waste form release properties, and radionuclide specific transport properties. Verification studies performed on the code are discussed

  17. Natural hazards that may trigger a radiological release from a plutonium processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Selvidge, J. E.

    1977-04-28

    Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10/sup -4/ per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10/sup -7/ grams. The source term for high-velocity, downslope winds is higher--2.2 x 10/sup -3/ grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10/sup -7/ per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated.

  18. Natural hazards that may trigger a radiological release from a plutonium processing facility

    International Nuclear Information System (INIS)

    Selvidge, J.E.

    1977-01-01

    Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10 -4 per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10 -7 grams. The source term for high-velocity, downslope winds is higher--2.2 x 10 -3 grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10 -7 per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated

  19. Backup Sourcing Decisions for Coping with Supply Disruptions under Long-Term Horizons

    Directory of Open Access Journals (Sweden)

    Jing Hou

    2016-01-01

    Full Text Available This paper studies a buyer’s inventory control problem under a long-term horizon. The buyer has one major supplier that is prone to disruption risks and one backup supplier with higher wholesale price. Two kinds of sourcing methods are available for the buyer: single sourcing with/without contingent supply and dual sourcing. In contingent sourcing, the backup supplier is capacitated and/or has yield uncertainty, whereas in dual sourcing the backup supplier has an incentive to offer output flexibility during disrupted periods. The buyer’s expected cost functions and the optimal base-stock levels using each sourcing method under long-term horizon are obtained, respectively. The effects of three risk parameters, disruption probability, contingent capacity or uncertainty, and backup flexibility, are examined using comparative studies and numerical computations. Four sourcing methods, namely, single sourcing with contingent supply, dual sourcing, and single sourcing from either of the two suppliers, are also compared. These findings can be used as a valuable guideline for companies to select an appropriate sourcing strategy under supply disruption risks.

  20. Lessons Learned from Characterization, Performance Assessment, and EPA Regulatory Review of the 1996 Actinide Source Term for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Larson, K.W.; Moore, R.C.; Nowak, E.J.; Papenguth, H.W.; Jow, H.

    1999-01-01

    The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy (DOE) facility for the permanent disposal of transuranic waste from defense activities. In 1996, the DOE submitted the Title 40 CFR Part 191 Compliance Certification Application for the Waste Isolation Pilot Plant (CCA) to the US Environmental Protection Agency (EPA). The CCA included a probabilistic performance assessment (PA) conducted by Sandia National Laboratories to establish compliance with the quantitative release limits defined in 40 CFR 191.13. An experimental program to collect data relevant to the actinide source term began around 1989, which eventually supported the 1996 CCA PA actinide source term model. The actinide source term provided an estimate of mobile dissolved and colloidal Pu, Am, U, Th, and Np concentrations in their stable oxidation states, and accounted for effects of uncertainty in the chemistry of brines in waste disposal areas. The experimental program and the actinide source term included in the CCA PA underwent EPA review lasting more than 1 year. Experiments were initially conducted to develop data relevant to the wide range of potential future conditions in waste disposal areas. Interim, preliminary performance assessments and actinide source term models provided insight allowing refinement of experiments and models. Expert peer review provided additional feedback and confidence in the evolving experimental program. By 1995, the chemical database and PA predictions of WIPP performance were considered reliable enough to support the decision to add an MgO backfill to waste rooms to control chemical conditions and reduce uncertainty in actinide concentrations, especially for Pu and Am. Important lessons learned through the characterization, PA modeling, and regulatory review of the actinide source term are (1) experimental characterization and PA should evolve together, with neither activity completely dominating the other, (2) the understanding of physical processes

  1. Distinct transmitter release properties determine differences in short-term plasticity at functional and silent synapses.

    Science.gov (United States)

    Cabezas, Carolina; Buño, Washington

    2006-05-01

    Recent evidence suggests that functional and silent synapses are not only postsynaptically different but also presynaptically distinct. The presynaptic differences may be of functional importance in memory formation because a proposed mechanism for long-term potentiation is the conversion of silent synapses into functional ones. However, there is little direct experimentally evidence of these differences. We have investigated the transmitter release properties of functional and silent Schaffer collateral synapses and show that on the average functional synapses displayed a lower percentage of failures and higher excitatory postsynaptic current (EPSC) amplitudes than silent synapses at +60 mV. Moreover, functional but not silent synapses show paired-pulse facilitation (PPF) at +60 mV and thus presynaptic short-term plasticity will be distinct in the two types of synapse. We examined whether intraterminal endoplasmic reticulum Ca2+ stores influenced the release properties of these synapses. Ryanodine (100 microM) and thapsigargin (1 microM) increased the percentage of failures and decreased both the EPSC amplitude and PPF in functional synapses. Caffeine (10 mM) had the opposite effects. In contrast, silent synapses were insensitive to both ryanodine and caffeine. Hence we have identified differences in the release properties of functional and silent synapses, suggesting that synaptic terminals of functional synapses express regulatory molecular mechanisms that are absent in silent synapses.

  2. A simple method for estimating potential source term bypass fractions from confinement structures

    International Nuclear Information System (INIS)

    Kalinich, D.A.; Paddleford, D.F.

    1997-01-01

    Confinement structures house many of the operating processes at the Savannah River Site (SRS). Under normal operating conditions, a confinement structure in conjunction with its associated ventilation systems prevents the release of radiological material to the environment. However, under potential accident conditions, the performance of the ventilation systems and integrity of the structure may be challenged. In order to calculate the radiological consequences associated with a potential accident (e.g. fires, explosion, spills, etc.), it is necessary to determine the fraction of the source term initially generated by the accident that escapes from the confinement structure to the environment. While it would be desirable to estimate the potential bypass fraction using sophisticated control-volume/flow path computer codes (e.g. CONTAIN, MELCOR, etc.) in order to take as much credit as possible for the mitigative effects of the confinement structure, there are many instances where using such codes is not tractable due to limits on the level-of-effort allotted to perform the analysis. Moreover, the current review environment, with its emphasis on deterministic/bounding-versus probabilistic/best-estimate-analysis discourages using analytical techniques that require the consideration of a large number of parameters. Discussed herein is a simplified control-volume/flow path approach for calculating source term bypass fraction that is amenable to solution in a spreadsheet or with a commercial mathematical solver (e.g. MathCad or Mathematica). It considers the effects of wind and fire pressure gradients on the structure, ventilation system operation, and Halon discharges. Simple models are used to characterize the engineered and non-engineered flow paths. By making judicious choices for the limited set of problem parameters, the results from this approach can be defended as bounding and conservative

  3. Consideration of emergency source terms for pebble-bed high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tao, Liu; Jun, Zhao; Jiejuan, Tong; Jianzhu, Cao

    2009-01-01

    Being the last barrier in the nuclear power plant defense-in-depth strategy, emergency planning (EP) is an integrated project. One of the key elements in this process is emergency source terms selection. Emergency Source terms for light water reactor (LWR) nuclear power plant (NPP) have been introduced in many technical documents, and advanced NPP emergency planning is attracting attention recently. Commercial practices of advanced NPP are undergoing in the world, pebble-bed high-temperature gas-cooled reactor (HTGR) power plant is under construction in China which is considered as a representative of advanced NPP. The paper tries to find some pieces of suggestion from our investigation. The discussion of advanced NPP EP will be summarized first, and then the characteristics of pebble-bed HTGR relating to EP will be described. Finally, PSA insights on emergency source terms selection and current pebble-bed HTGR emergency source terms suggestions are proposed

  4. Review of tellurium release rates from LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Beahm, E.C.; Wichner, R.P.

    1983-01-01

    Although fission product tellurium presents a potentially significant radiohazard, its release and transport in source-term experiments is frequently overlooked because it does not possess a readily measurable, gamma emission; moreover, a recent study emphasized noble gas, iodine and cesium release from LWR fuel elements because of the large data base that exists for these materials. Some new tests show that in some cases tellurium may be held up in core material to a greater degree than previously assumed - an observation that prompts a careful reappraisal of the existing tellurium-release data and its chemical foundation

  5. 14 CFR Appendix I to Part 417 - Methodologies for Toxic Release Hazard Analysis and Operational Procedures

    Science.gov (United States)

    2010-01-01

    ... Quotient/Hazard Index (HQ/HI) formulation to determine the toxic concentration threshold for mixtures of... cloud center of mass altitude. (vi) Worst case initial source term assuming instantaneous release of...

  6. PFLOTRAN-RepoTREND Source Term Comparison Summary.

    Energy Technology Data Exchange (ETDEWEB)

    Frederick, Jennifer M

    2018-03-01

    Code inter-comparison studies are useful exercises to verify and benchmark independently developed software to ensure proper function, especially when the software is used to model high-consequence systems which cannot be physically tested in a fully representative environment. This summary describes the results of the first portion of the code inter-comparison between PFLOTRAN and RepoTREND, which compares the radionuclide source term used in a typical performance assessment.

  7. Generation and release of radioactive gases in LLW disposal facilities

    Energy Technology Data Exchange (ETDEWEB)

    Yim, M.S. [Harvard School Public Health, Boston, MA (United States); Simonson, S.A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-02-01

    The atmospheric release of radioactive gases from a generic engineered LLW disposal facility and its radiological impacts were examined. To quantify the generation of radioactive gases, detailed characterization of source inventory for carbon-14, tritium, iodine-129, krypton-85, and radon-222, was performed in terms of their activity concentrations; their distribution within different waste classes, waste forms and containers; and their subsequent availability for release in volatile or gaseous form. The generation of gases was investigated for the processes of microbial activity, radiolysis, and corrosion of waste containers and metallic components in wastes. The release of radionuclides within these gases to the atmosphere was analyzed under the influence of atmospheric pressure changes.

  8. Source term estimation for small sized HTRs: status and further needs - a german approach

    International Nuclear Information System (INIS)

    Moormann, R.; Schenk, W.; Verfondern, K.

    2000-01-01

    The main results of German studies on source term estimation for small pebble-bed HTRs with their strict safety demands are outlined. Core heat-up events are no longer dominant for modern high quality fuel, but fission product transport during water ingress accidents (steam cycle plants) and depressurization is relevant, mainly due to remobilization of fission products which were plated-out in the course of normal operation or became dust borne. An important lack of knowledge was identified as concerns data on plate-out under normal operation, as well as on the behaviour of dust borne activity as a whole. Improved knowledge in this field is also important for maintenance/repair and design/shielding. For core heat-up events the influence of burn-up on temperature induced fission product release has to be measured for future high burn-up fuel. Also, transport mechanisms out of the He circuit into the environment require further examination. For water/steam ingress events mobilization of plated-out fission products by steam or water has to be considered in detail, along with steam interaction with kernels of particles with defective coatings. For source terms of depressurization, a more detailed knowledge of the flow pattern and shear forces on the various surfaces is necessary. In order to improve the knowledge on plate-out and dust in normal operation and to generate specimens for experimental remobilization studies, planning/design of plate-out/dust examination facilities which could be added to the next generation of HTRs (HTR10,HTTR) is proposed. For severe air ingress and reactivity accidents, behaviour of future advanced fuel elements has to be experimentally tested. (authors)

  9. Low-level waste disposal performance assessments - Total source-term analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilhite, E.L.

    1995-12-31

    Disposal of low-level radioactive waste at Department of Energy (DOE) facilities is regulated by DOE. DOE Order 5820.2A establishes policies, guidelines, and minimum requirements for managing radioactive waste. Requirements for disposal of low-level waste emplaced after September 1988 include providing reasonable assurance of meeting stated performance objectives by completing a radiological performance assessment. Recently, the Defense Nuclear Facilities Safety Board issued Recommendation 94-2, {open_quotes}Conformance with Safety Standards at Department of Energy Low-Level Nuclear Waste and Disposal Sites.{close_quotes} One of the elements of the recommendation is that low-level waste performance assessments do not include the entire source term because low-level waste emplaced prior to September 1988, as well as other DOE sources of radioactivity in the ground, are excluded. DOE has developed and issued guidance for preliminary assessments of the impact of including the total source term in performance assessments. This paper will present issues resulting from the inclusion of all DOE sources of radioactivity in performance assessments of low-level waste disposal facilities.

  10. PLOTLIB: a computerized nuclear waste source-term library storage and retrieval system

    International Nuclear Information System (INIS)

    Marshall, J.R.; Nowicki, J.A.

    1978-01-01

    The PLOTLIB code was written to provide computer access to the Nuclear Waste Source-Term Library for those users with little previous computer programming experience. The principles of user orientation, quick accessibility, and versatility were extensively employed in the development of the PLOTLIB code to accomplish this goal. The Nuclear Waste Source-Term Library consists of 16 ORIGEN computer runs incorporating a wide variety of differing light water reactor (LWR) fuel cycles and waste streams. The typical isotopic source-term data consist of information on watts, curies, grams, etc., all of which are compiled as a function of time after reactor discharge and unitized on a per metric ton heavy metal basis. The information retrieval code, PLOTLIB, is used to process source-term information requests into computer plots and/or user-specified output tables. This report will serve both as documentation of the current data library and as an operations manual for the PLOTLIB computer code. The accompanying input description, program listing, and sample problems make this code package an easily understood tool for the various nuclear waste studies under way at the Office of Waste Isolation

  11. EPICS release 3.11 specific documentation -- EPICS release notes for 3.11

    International Nuclear Information System (INIS)

    1994-01-01

    EPICS release 3.11 is now ready for user testing. A person who wants to set up a simplified application environment to boot an IOC and create databases using R3.11 should follow the directions in Appendix B, page 27, of the EPICS Source/Release Control Manual, Sept. 20, 1993. The R3.11 EPICS path at ANL/APS is /net/phebos/epics/R3.11 so the command to get the new release is /net/phebos/epics/R3.11/Unix/share/bin/getrel /net/phebos/epics/R3.11. An existing R3.8 short form report can be copied to this new directory and used to create a database. ANL/APS is currently testing an Application Developers Source/Release control system. It is not yet ready for general distribution. Attached are the EPICS R3.11 release notes

  12. Low-level radioactive waste source terms for the 1992 integrated data base

    International Nuclear Information System (INIS)

    Loghry, S.L.; Kibbey, A.H.; Godbee, H.W.; Icenhour, A.S.; DePaoli, S.M.

    1995-01-01

    This technical manual presents updated generic source terms (i.e., unitized amounts and radionuclide compositions) which have been developed for use in the Integrated Data Base (IDB) Program of the U.S. Department of Energy (DOE). These source terms were used in the IDB annual report, Integrated Data Base for 1992: Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics, DOE/RW-0006, Rev. 8, October 1992. They are useful as a basis for projecting future amounts (volume and radioactivity) of low-level radioactive waste (LLW) shipped for disposal at commercial burial grounds or sent for storage at DOE solid-waste sites. Commercial fuel cycle LLW categories include boiling-water reactor, pressurized-water reactor, fuel fabrication, and uranium hexafluoride (UF 6 ) conversion. Commercial nonfuel cycle LLW includes institutional/industrial (I/I) waste. The LLW from DOE operations is category as uranium/thorium fission product, induced activity, tritium, alpha, and open-quotes otherclose quotes. Fuel cycle commercial LLW source terms are normalized on the basis of net electrical output [MW(e)-year], except for UF 6 conversion, which is normalized on the basis of heavy metal requirement [metric tons of initial heavy metal ]. The nonfuel cycle commercial LLW source term is normalized on the basis of volume (cubic meters) and radioactivity (curies) for each subclass within the I/I category. The DOE LLW is normalized in a manner similar to that for commercial I/I waste. The revised source terms are based on the best available historical data through 1992

  13. Impact of source terms on distances to which reactor accident consequences occur

    International Nuclear Information System (INIS)

    Ostmeyer, R.M.

    1982-01-01

    Estimates of the distances over which reactor accident consequences might occur are important for development of siting criteria and for emergency response planning. This paper summarizes the results of a series of CRAC2 calculations performed to estimate these distances. Because of the current controversy concerning the magnitude of source terms for severe accidents, the impact of source term reductions upon distance estimates is also examined

  14. The long-term problems of contaminated land: Sources, impacts and countermeasures

    International Nuclear Information System (INIS)

    Baes, C.F. III.

    1986-11-01

    This report examines the various sources of radiological land contamination; its extent; its impacts on man, agriculture, and the environment; countermeasures for mitigating exposures; radiological standards; alternatives for achieving land decontamination and cleanup; and possible alternatives for utilizing the land. The major potential sources of extensive long-term land contamination with radionuclides, in order of decreasing extent, are nuclear war, detonation of a single nuclear weapon (e.g., a terrorist act), serious reactor accidents, and nonfission nuclear weapons accidents that disperse the nuclear fuels (termed ''broken arrows'')

  15. Effects of indoor residence on radiation doses from routine releases of radionuclides to the atmosphere

    International Nuclear Information System (INIS)

    Kocher, D.C.

    1980-01-01

    Dose reduction factors from indoor residence during routine releases of radionuclides to the atmosphere were studied using models that are suitable for application to arbitrary source terms. Dose reduction factors for internal exposure to inhaled radionuclides account for air ventilation and deposition on inside building surfaces. Estimated internal dose reduction factors are approx. 0.2 to 0.8 for particulates and 0.07 to 0.4 for radioiodine. Dose reduction factors for external photon exposure from airborne and surface-deposited sources are based on the point-kernel integration method. Values for source terms from a fuel reprocessing plant and a hypothetical reactor accident are within a factor of 2 of the value 0.5 adopted by the US Nuclear Regulatory Commission (NRC) for population dose assessments. For the release at Three Mile Island nuclear station, however, the external dose reduction factor may be an order of magnitude less than the value adopted by the NRC

  16. Discussion on the source of radon in uranium exploration method using radon-released thermal effect in minerals

    International Nuclear Information System (INIS)

    Su Shoutian.

    1985-01-01

    This paper considers the problem of the source of radon in uranium exploration method using radon-released thermal effect. In minerals by means of scintillation emanometry, we have carried out the measurement on radon content in minerals at various temperature in barren and ore-bearing granites of the granite-type uranium deposit No. 752, and inclusion decrepitation method has also been used to determine the temperature of decrepitation and its relative frequency. It was found from experiments that heated samples may release most of radon prior to inclusion decrepitation, radon released from thermal effect was, on the contrary, very little at temperature intervals of inclusion decrepitation on a large scale basis. When inclusions were ground after radon releasing, it would still release from inclusions after reheating. The radon content calculated from uranium content in inclusions is lower than the sensitivity of the determination method, so it is too difficult to be determined, indicating that the radon content released is not related to inclusions. Samples were determined by uranium chemical analysis and radium radiochemical analysis and it is obvious to note that the radon content released from thermal effect in minerals is positively correlated to the uranium and radium content. Various kinds of experiments suggest that radon is not derived from inclusions but from the whole mineral

  17. Benchmarking of fast-running software tools used to model releases during nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Devitt, P.; Viktorov, A., E-mail: Peter.Devitt@cnsc-ccsn.gc.ca, E-mail: Alex.Viktorov@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Fukushima highlighted the importance of effective nuclear accident response. However, its complexity greatly impacted the ability to provide timely and accurate information to national and international stakeholders. Safety recommendations provided by different national and international organizations varied notably. Such differences can partially be attributed to different methods used in the initial assessment of accident progression and the amount of radioactivity release.Therefore, a comparison of methodologies was undertaken by the NEA/CSNI and its highlights are presented here. For this project, the prediction tools used by various emergency response organizations for estimating the source terms and public doses were examined. Those organizations that have a capability to use such tools responded to a questionnaire describing each code's capabilities and main algorithms. Then the project's participants analyzed five accident scenarios to predict the source term, dispersion of releases and public doses. (author)

  18. CERN Web Application Detection. Refactoring and release as open source software

    CERN Document Server

    Lizonczyk, Piotr

    2015-01-01

    This paper covers my work during my assignment as participant of CERN Summer Students 2015 programme. The project was aimed at refactoring and publication of the Web Application Detection tool, which was developed at CERN and priorly used internally by the Computer Security team. The range of tasks performed include initial refactoring of code, which was developed like a script rather than a Python package, through extracting components that were not specific to CERN usage, the subsequent final release of the source code on GitHub and the integration with third-party software i.e. the w3af tool. Ultimately, Web Application Detection software received positive responses, being downloaded ca. 1500 times at the time of writing this report.

  19. The long-term problems of contaminated land: Sources, impacts and countermeasures

    Energy Technology Data Exchange (ETDEWEB)

    Baes, C.F. III

    1986-11-01

    This report examines the various sources of radiological land contamination; its extent; its impacts on man, agriculture, and the environment; countermeasures for mitigating exposures; radiological standards; alternatives for achieving land decontamination and cleanup; and possible alternatives for utilizing the land. The major potential sources of extensive long-term land contamination with radionuclides, in order of decreasing extent, are nuclear war, detonation of a single nuclear weapon (e.g., a terrorist act), serious reactor accidents, and nonfission nuclear weapons accidents that disperse the nuclear fuels (termed ''broken arrows'').

  20. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  1. The initial phase of sudden releases of superheated liquid

    International Nuclear Information System (INIS)

    Schmidli, J.

    1994-04-01

    The catastrophic failure of a pressure vessel containing a liquefied substance, leading to an instantaneous release of its whole contents is considered as one of the major technological hazards. Due to the rapid depressurization caused by vessel failure, the fluid becomes superheated and unstable. Part of the fluid will evaporate using its internal energy and the two-phase mixture forming will be accelerated. This flashing process can be very violent, as experiments and incidents actually happened have shown. In the past, a number of dispersion models were developed to predict the history of an instantaneous release. In most of these models the source term is considered to be a gas volume at rest and not a rapidly expanding aerosol, as could be observed. Furthermore, it is usually assumed that all of the remaining fluid is entrained into the expanding cloud and nothing is deposited on the ground to form a pool. This work concentrates on the initial phase of the sudden release of superheated liquids with the aim to gain a better understanding of the flashing process and of the physical mechanisms involved, leading to a reliable prediction of the source term. Therefore, more than 400 experiments with propane, butane, refrigerant 12 and 114 were conducted. The experiments were initiated by shattering spherical glass flasks of different sizes. The main parameters varied were the liquid superheat and the filling level of the vessel. Using high-speed video and movie recordings and very fast responding measurement devices, it was possible to study the initial phase of such releases during which gravity plays no role. For sufficiently large released internal energy, the initial evolution of the release was always spherical with a constant radial expansion velocity during he first milliseconds until instabilities appeared at the surface of the droplet/vapor cloud that was formed. For all the experimental conditions, the fraction of the initial liquid falling on the ground

  2. The current status of ARAC [Atmospheric Release Advisory Capability] and its application to the Chernobyl event

    International Nuclear Information System (INIS)

    Gudiksen, P.H.; Sullivan, T.J.; Harvey, T.F.

    1986-10-01

    The Atmospheric Release Advisory Capability (ARAC) project, developed by the Lawrence Livermore National Laboratory (LLNL), provides real-time dose assessments and estimates of the extent of surface contamination that may result from an atmospheric release of radioactivity. It utilizes advanced computer-based data communication and processing systems to acquire the meteorological and source term information needed by the three-dimensional atmospheric dispersion models to derive the consequence assessments. The ARAC responded to the recent Chernobyl reactor accident in the Soviet Union by estimating the source term and the radiation dose distribution due to exposure to the radioactive cloud over Europe and the Northern Hemisphere. This analysis revealed that approximately 50% of the estimated core inventories of I-131 and Cs-137 were released. The estimated committed effective dose equivalent due to inhalation of radioactivty during cloud passage is of the order of 10 mrem within parts of Scandinavia and eastern Europe, while most of the populations within central Europe were exposed to levels ranging from 1 to 10 mrem. The amount of Cs-137 released by the Chernobyl accident far exceeds that released by previous reactor accidents, but is only about 6% of the Cs-137 produced by the atmospheric weapon testing programs. 9 refs., 4 figs., 2 tabs

  3. Release kinetics of tocopherol and quercetin from binary antioxidant controlled-release packaging films.

    Science.gov (United States)

    Chen, Xi; Lee, Dong Sun; Zhu, Xuntao; Yam, Kit L

    2012-04-04

    This paper investigated the feasibility of manipulating packaging polymers with various degrees of hydrophobicity to release two antioxidants, tocopherol and quercetin, at rates suitable for long-term inhibition of lipid oxidation in food. For example, one antioxidant can be released at a fast rate to provide short-term/intermediate protection, whereas the other antioxidant can be released at a slower rate to provide intermediate/long-term protection of lipid oxidation. Controlled-release packaging films containing tocopherol and quercetin were produced using ethylene vinyl alcohol (EVOH), ethylene vinyl acetate (EVA), low-density polyethylene (LDPE), and polypropylene (PP) polymers; the release of these antioxidants to 95% ethanol (a fatty food simulant) was measured using UV-vis spectrophotometry, and Fickian diffusion models with appropriate initial and boundary conditions were used to fit the data. For films containing only quercetin, the results show that the release of quercetin was much faster but lasted for a much shorter time for hydrophilic polymers (EVOH and EVA) than for hydrophobic polymers (LDPE and PP). For binary antioxidant films containing tocopherol and quercetin, the results show that tocopherol released more rapidly but for a shorter period of time than quercetin in LDPE and EVOH films, and the difference is more pronounced for LDPE films than EVOH films. The results also show the presence of tocopherol can accelerate the release of quercetin. Although none of the films produced is acceptable for long-term lipid oxidation inhibition, the study provides encouraging results suggesting that acceptable films may be produced in the future using polymer blend films.

  4. Data Release Report for Source Physics Experiment 1 (SPE-1), Nevada National Security Site

    Energy Technology Data Exchange (ETDEWEB)

    Townsend, Margaret [NSTec; Mercadente, Jennifer [NSTec

    2014-04-28

    The first Source Physics Experiment shot (SPE-1) was conducted in May 2011. The explosive source was a ~100-kilogram TNT-equivalent chemical set at a depth of 60 meters. It was recorded by an extensive set of instrumentation that includes sensors both at near-field (less than 100 meters) and far-field (more than 100 meters) distances. The near-field instruments consisted of three-component accelerometers deployed in boreholes around the shot and a set of singlecomponent vertical accelerometers on the surface. The far-field network comprised a variety of seismic and acoustic sensors, including short-period geophones, broadband seismometers, three-component accelerometers, and rotational seismometers at distances of 100 meters to 25 kilometers. This report coincides with the release of these data for analysts and organizations that are not participants in this program. This report describes the first Source Physics Experiment and the various types of near-field and far-field data that are available.

  5. Reducing plant radiation fields by source term reduction - tracking cobalt and antimony to their sources at Gentilly-2

    International Nuclear Information System (INIS)

    Gauthier, P.; Guzonas, D.A.

    2006-01-01

    Gentilly-2 NGS is experiencing high radiation fields in the fuelling machine vaults. These high fields make maintenance outages more expensive and their management more complicated. As part of the station refurbishment project, a task group was created to identify the cause of the high fields and make recommendations to prevent their reoccurrence in the second (post-refurbishment) operating cycle. To identify the root cause of the problem, the task group decided to analyse the primary heat transport system (PHTS), the fuel handling system and their inter-relation. Gentilly-2 has had to manage a unique (to CANDU) problem arising from antimony released from the main heat transport pump seals. Antimony deposits on in-core surfaces, becomes activated, and subsequently can be released, especially under oxidizing coolant conditions. It then becomes incorporated into the magnetite deposits on PHTS piping, including the steam generators and inlet feeders. Gentilly-2 has focused a great deal of effort on managing antimony over the last 15 years. As a result of these initiatives, radioantimony fields have been quite effectively managed since 1997, resulting in a decrease in their relative contribution to the total fields. The decrease in radioantimony fields highlighted the significant contribution of 60 Co cobalt activity; the high levels of both radioantimony and 60 Co differentiate Gentilly-2 from other CANDU 6 plants. Two types of 59 Co sources are present in the CANDU PHTS. High surface area materials such as steam generator tubes and feeder pipes contain trace concentrations of 59 Co as an impurity, which can be released by corrosion. Low surface area materials such as Stellites contain high concentrations of 59 Co that can be released as either corrosion or wear products. After assessing potential cobalt sources, the task group concluded that PHTS materials were not likely the origin of the high 60 Co fields. The major PHTS components identified as cobalt sources have

  6. Short-term Outcomes following Concussion in the NFL: An 11-year Retrospective Study of Player Release Rate and Financial Loss

    Science.gov (United States)

    Ramkumar, Prem; Navarro, Sergio Michael

    2017-01-01

    Objectives: The primary goal of this study was to assess the short-term outcomes among National Football League (NFL) players following concussion in terms of: (1) DNP protocol activation, (2) release rate at one and three years, and (3) mean salary reduction. A secondary goal of the study was to stratify the post-concussive release rate by franchise and player position. Methods: NFL player transaction records and publicly available weekly injury reports from August 2005 to January 2016 for NFL players were analyzed. All players immediately sustaining recorded concussions were evaluated for a change to inactive or do-not-play (DNP) status. The one-year and three-year release rate following concussion was defined as any player transitioning to inactivation, retirement, free agency, or any failure to return for a successive season on the same team’s active roster after one or three years from the initial concussion. Student’s t-test was used to compare release rates between non-concussed and concussed players at one and three years. Mean salary reduction per year following concussion was calculated using publicly available player contracts. Additionally, franchise-level and position-based analyses of the release rate were performed. Results: Of the total 5,451 NFL players retrospectively analyzed over the 11-year period, 373 sustained publicly reported concussions resulting in DNP protocol activation. The release rate of the post-concussive versus non-concussive player was 26% vs. 20% at 1 year (pfranchise to release an athlete following concussion within one and three years. Table 1 reports a position-based analysis in terms of concussion rate, mean salary reduction, and NFL career longevity. Conclusion: Our retrospective study demonstrates that NFL concussions resulting in DNP protocol activation leads to a statistically greater release rate among concussed NFL players than non-concussed players. Released players suffered reduction in year-over-year accumulated

  7. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  8. Added Value of uncertainty Estimates of SOurce term and Meteorology (AVESOME)

    DEFF Research Database (Denmark)

    Sørensen, Jens Havskov; Schönfeldt, Fredrik; Sigg, Robert

    In the early phase of a nuclear accident, two large sources of uncertainty exist: one related to the source term and one associated with the meteorological data. Operational methods are being developed in AVESOME for quantitative estimation of uncertainties in atmospheric dispersion prediction.......g. at national meteorological services, the proposed methodology is feasible for real-time use, thereby adding value to decision support. In the recent NKS-B projects MUD, FAUNA and MESO, the implications of meteorological uncertainties for nuclear emergency preparedness and management have been studied...... uncertainty in atmospheric dispersion model forecasting stemming from both the source term and the meteorological data is examined. Ways to implement the uncertainties of forecasting in DSSs, and the impacts on real-time emergency management are described. The proposed methodology allows for efficient real...

  9. The influence of "host release factor" on carbon release by zooxanthellae isolated from fed and starved Aiptasia pallida (Verrill).

    Science.gov (United States)

    Davy, S K; Cook, C B

    2001-06-01

    Symbiotic dinoflagellates (zooxanthellae) typically respond to extracts of host tissue with enhanced release of short-term photosynthetic products. We examined this "host release factor" (HRF) response using freshly isolated zooxanthellae of differing nutritional status. The nutritional status was manipulated by either feeding or starving the sea anemone Aiptasia pallida (Verrill). The release of fixed carbon from isolated zooxanthellae was measured using 14C in 30 min experiments. Zooxanthellae in filtered seawater alone released approximately 5% of photosynthate irrespective of host feeding history. When we used a 10-kDa ultrafiltrate of A. pallida host tissue as a source of HRF, approximately 14% of photosynthate was released to the medium. This increased to over 25% for zooxanthellae from anemones starved for 29 days or more. The cell-specific photosynthetic rate declined with starvation in these filtrate experiments, but the decline was offset by the increased percentage release. Indeed, the total amount of released photosynthate remained unchanged, or even increased, as zooxanthellae became more nutrient deficient. Similar trends were also observed when zooxanthellae from A. pallida were incubated in a 3-kDa ultrafiltrate of the coral Montastraea annularis, suggesting that HRF in the different filtrates operated in a similar manner. Our results support the suggestion that HRF diverts surplus carbon away from storage compounds to translocated compounds such as glycerol.

  10. Design parameters and source terms: Volume 1, Design parameters: Revision 0

    International Nuclear Information System (INIS)

    1987-10-01

    The Design Parameters and Source Terms Document was prepared in accordance with DOE request and to provide data for the environmental impact study to be performed in the future for the Deaf Smith County, Texas site for a nuclear waste repository in salt. This document updates a previous unpublished report by Stearns Catalytic Corporation (SCC), entitled ''Design Parameters and Source Terms for a Two-Phase Repository in Salt,'' 1985, to the level of the Site Characterization Plan - Conceptual Design Report. The previous unpublished SCC Study identifies the data needs for the Environmental Assessment effort for seven possible Salt Repository sites

  11. Radiological consequence evaluation of DBAs with alternative source term method for a Chinese PWR

    International Nuclear Information System (INIS)

    Li, J.X.; Cao, X.W.; Tong, L.L.; Huang, G.F.

    2012-01-01

    Highlights: ► Radiological consequence evaluation of DBAs with alternative source term method for a Chinese 900 MWe PWR has been investigated. ► Six typical DBA sequences are analyzed. ► The doses of control room, EAB and outer boundary of LPZ are acceptable. ► The differences between AST method and TID-14844 method are investigated. - Abstract: Since a large amount of fission products may releases into the environment, during the accident progression in nuclear power plants (NPPs), which is a potential hazard to public risk, the radiological consequence should be evaluated for alleviating the hazard. In most Chinese NPPs the method of TID-14844, in which the whole body and thyroid dose criteria is employed as dose criteria, is currently adopted to evaluate the radiological consequences for design-basis accidents (DBAs), but, due to the total effective dose equivalent is employed as dose criteria in alternative radiological source terms (AST) method, it is necessary to evaluate the radiological consequences for DBAs with AST method and to discuss the difference between two methods. By using an integral safety analysis code, an analytical model of the 900 MWe pressurized water reactor (PWR) is built and the radiological consequences in DBAs at control room (CR), exclusion area boundary (EAB), low population zone (LPZ) are analyzed, which includes LOCA and non-LOCA DBAs, such as fuel handling accident (FHA), rod ejection accident (REA), main steam line break (MSLB), steam generator tube rupture (SGTR), locked rotor accident (LRA) by using the guidance of the RG 1.183. The results show that the doses in CR, EAB and LPZ are acceptable compared with dose criteria in RG 1.183 and the differences between AST method and TID-14844 method are also discussed.

  12. Scenarios for potential radionuclide release from marine reactors dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Lynn, N.; Mount, M.; Gussgard, K.

    1995-01-01

    The largest inventory of radioactive materials dumped in the Kara Sea by the former Soviet Union comes from the spent nuclear fuel (SNF) of seven marine reactors, the current (1994) inventory of which makes a total of approximately 4.7x10 15 Bq. In progressing its work for the International Arctic Seas Assessment Project, under the auspices of the International Atomic Energy Agency, the Source Term Working Group has analysed the Source Term and subsequently developed a number of model scenarios for the potential release patterns of radionuclides into the Kara Sea from the SNF and activated components dumped within the marine reactors.These models are based on the present and future conditions of the barrier materials and their configuration within the dumped objects. They account for progressive corrosion of the outer and inner steel barriers, breakdown of the organic fillers, and degradation and leaching from the SNFs. Annual release rates are predicted to four thousand years into the future. 5 refs., 1 fig., 1 tab

  13. Comparison of source-term calculations using the AREST and SYVAC-Vault models: [Final report

    International Nuclear Information System (INIS)

    Apted, M.J.; Engel, D.W.; Garisto, N.C.; LeNeveu, D.M.

    1988-07-01

    A comparison of the calculated radionuclide release from a waste package in a geologic repository has been performed using the verified SYVAC-Vault Model and AREST Model. the purpose of this comparison is to further establish the credibility of these codes for predictive performance assessment and to identify improvements that may be required. A reference case for a Canadian conceptual design with spent fuel as the waste form was chosen to make an initial comparison. The results from the two models were in good agreement, including peak release rates, time to reach peak release, and long term release rates. Differences in results from the two models are attributed to differences in computational approaches. Studies of the effects of sorption, convective flow, distributed containment failure, and precipitation are identified as key areas for further comparisons and are currently in progress. 11 refs., 3 figs., 5 tabs

  14. Coarse Grid Modeling of Turbine Film Cooling Flows Using Volumetric Source Terms

    Science.gov (United States)

    Heidmann, James D.; Hunter, Scott D.

    2001-01-01

    The recent trend in numerical modeling of turbine film cooling flows has been toward higher fidelity grids and more complex geometries. This trend has been enabled by the rapid increase in computing power available to researchers. However, the turbine design community requires fast turnaround time in its design computations, rendering these comprehensive simulations ineffective in the design cycle. The present study describes a methodology for implementing a volumetric source term distribution in a coarse grid calculation that can model the small-scale and three-dimensional effects present in turbine film cooling flows. This model could be implemented in turbine design codes or in multistage turbomachinery codes such as APNASA, where the computational grid size may be larger than the film hole size. Detailed computations of a single row of 35 deg round holes on a flat plate have been obtained for blowing ratios of 0.5, 0.8, and 1.0, and density ratios of 1.0 and 2.0 using a multiblock grid system to resolve the flows on both sides of the plate as well as inside the hole itself. These detailed flow fields were spatially averaged to generate a field of volumetric source terms for each conservative flow variable. Solutions were also obtained using three coarse grids having streamwise and spanwise grid spacings of 3d, 1d, and d/3. These coarse grid solutions used the integrated hole exit mass, momentum, energy, and turbulence quantities from the detailed solutions as volumetric source terms. It is shown that a uniform source term addition over a distance from the wall on the order of the hole diameter is able to predict adiabatic film effectiveness better than a near-wall source term model, while strictly enforcing correct values of integrated boundary layer quantities.

  15. Unclassified Source Term and Radionuclide Data for Corrective Action Unit 98: Frenchman Flat Nevada Test Site, Nevada

    International Nuclear Information System (INIS)

    Farnham, Irene

    2005-01-01

    Frenchman Flat is one of several areas of the Nevada Test Site (NTS) used for underground nuclear testing (Figure 1-1). These nuclear tests resulted in groundwater contamination in the vicinity of the underground test areas. As a result, the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) is currently conducting a corrective action investigation (CAI) of the Frenchman Flat underground test areas. Since 1996, the Nevada Division of Environmental Protection (NDEP) has regulated NNSA/NSO corrective actions through the ''Federal Facility Agreement and Consent Order'' ([FFACO], 1996). Appendix VI of the FFACO agreement, ''Corrective Action Strategy'', was revised on December 7, 2000, and describes the processes that will be used to complete corrective actions, including those in the Underground Test Area (UGTA) Project. The individual locations covered by the agreement are known as corrective action sites (CASs), which are grouped into corrective action units (CAUs). The UGTA CASs are grouped geographically into five CAUs: Frenchman Flat, Central Pahute Mesa, Western Pahute Mesa, Yucca Flat/Climax Mine, and Rainier Mesa/Shoshone Mountain (Figure 1-1). These CAUs have distinctly different contaminant source, geologic, and hydrogeologic characteristics related to their location (FFACO, 1996). The Frenchman Flat CAU consists of 10 CASs located in the northern part of Area 5 and the southern part of Area 11 (Figure 1-1). This report documents the evaluation of the information and data available on the unclassified source term and radionuclide contamination for Frenchman Flat, CAU 98. The methodology used to estimate hydrologic source terms (HSTs) for the Frenchman Flat CAU is also documented. The HST of an underground nuclear test is the portion of the total inventory of radionuclides that is released over time into the groundwater following the test. The total residual inventory of radionuclides associated with one or

  16. Estimation of in-plant source term release behaviors from Fukushima daiichi reactor cores by forward method and comparison with reverse method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Won; Rhee, Bo Wook; Song, Jin Ho; Kim, Sung Il; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012–018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3

  17. Operator aids for prediction of source term attenuation

    International Nuclear Information System (INIS)

    Powers, D.A.

    2004-01-01

    Simplified expressions for the attenuation of radionuclide releases by sprays and by water pools are devised. These expressions are obtained by correlation of the 10th, 50th and 90th percentiles of uncertainty distributions for the water pool decontamination factor and the spray decontamination coefficient. These uncertainty distributions were obtained by Monte Carlo uncertainty analyses using detailed, mechanistic models of the pools and sprays. Uncertainties considered in the analyses include uncertainties in the phenomena and uncertainties in the initial and boundary conditions dictated by the progression of severe accidents. Final results are graphically displayed in terms of the decontamination factor achieved at selected levels of conservatism versus pool depth and water subcooling or, in the case of sprays, versus time. (author)

  18. Estimation of the time-dependent radioactive source-term from the Fukushima nuclear power plant accident using atmospheric transport modelling

    Science.gov (United States)

    Schoeppner, M.; Plastino, W.; Budano, A.; De Vincenzi, M.; Ruggieri, F.

    2012-04-01

    Several nuclear reactors at the Fukushima Dai-ichi power plant have been severely damaged from the Tōhoku earthquake and the subsequent tsunami in March 2011. Due to the extremely difficult on-site situation it has been not been possible to directly determine the emissions of radioactive material. However, during the following days and weeks radionuclides of 137-Caesium and 131-Iodine (amongst others) were detected at monitoring stations throughout the world. Atmospheric transport models are able to simulate the worldwide dispersion of particles accordant to location, time and meteorological conditions following the release. The Lagrangian atmospheric transport model Flexpart is used by many authorities and has been proven to make valid predictions in this regard. The Flexpart software has first has been ported to a local cluster computer at the Grid Lab of INFN and Department of Physics of University of Roma Tre (Rome, Italy) and subsequently also to the European Mediterranean Grid (EUMEDGRID). Due to this computing power being available it has been possible to simulate the transport of particles originating from the Fukushima Dai-ichi plant site. Using the time series of the sampled concentration data and the assumption that the Fukushima accident was the only source of these radionuclides, it has been possible to estimate the time-dependent source-term for fourteen days following the accident using the atmospheric transport model. A reasonable agreement has been obtained between the modelling results and the estimated radionuclide release rates from the Fukushima accident.

  19. PCDD/PCDF release inventories

    Energy Technology Data Exchange (ETDEWEB)

    Fiedler, H. [UNEP Chemicals, Chatelaine (Switzerland)

    2004-09-15

    The Stockholm Convention on Persistent Organic Pollutants (POPs) entered into force on 17 May 2004 with 50 Parties. In May 2004, 59 countries had ratified or acceded the Convention. The objective of the Convention is ''to protect human health and the environment from persistent organic pollutants''. For intentionally produced POPs, e.g., pesticides and industrial chemicals such as hexachlorobenzene and polychlorinated biphenyls, this will be achieved by stop of production and use. For unintentionally generated POPs, such as polychlorinated dibenzo-pdioxins (PCDD) and polychlorinated dibenzofurans (PCDF), measures have to be taken to ''reduce the total releases derived from anthropogenic sources''; the final goal is ultimate elimination, where feasible. Under the Convention, Parties have to establish and maintain release inventories to prove the continuous release reduction. Since many countries do not have the technical and financial capacity to measure all releases from all potential PCDD/PCDF sources, UNEP Chemicals has developed the ''Standardized Toolkit for the Identification of Quantification of Dioxin and Furan Releases'' (''Toolkit'' for short), a methodology to estimate annual releases from a number of sources. With this methodology, annual releases can be estimated by multiplying process-specific default emission factors provided in the Toolkit with national activity data. At the seventh session of the Intergovernmental Negotiating Committee, the Toolkit was recommended to be used by countries when reporting national release data to the Conference of the Parties. The Toolkit is especially used by developing countries and countries with economies in transition where no measured data are available. Results from Uruguay, Thailand, Jordan, Philippines, and Brunei Darussalam have been published.

  20. Selected source term topics. Report to CSNI by an OECD/NEA Group of experts

    International Nuclear Information System (INIS)

    1987-04-01

    CSNI Report 136 summarizes the results of the work performed by the Group of Experts on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 and 1986. This report is complementary to Part 1, 'Technical Status of the Source Term' of CSNI Report 135, 'Report to CSNI on Source Term Assessment, Containment atmosphere control systems, and accident consequences'; it considers in detail a number of very specific issues thought to be important in the source term area. It consists of: an executive summary (prepared by the Chairman of the Group), a section on conclusions and recommendations, and five technical chapters (fission product chemistry in the primary circuit of a LWR during severe accidents; resuspension/re-entrainment of aerosols in LWRs following a meltdown accident; iodine chemistry under severe accident conditions; effects of combustion, steam explosions and pressurized melt ejection on fission product behaviour; radionuclide removal by pool scrubbing), a technical annex and two appendices

  1. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  2. A note on variational multiscale methods for high-contrast heterogeneous porous media flows with rough source terms

    KAUST Repository

    Calo, Victor M.

    2011-09-01

    In this short note, we discuss variational multiscale methods for solving porous media flows in high-contrast heterogeneous media with rough source terms. Our objective is to separate, as much as possible, subgrid effects induced by the media properties from those due to heterogeneous source terms. For this reason, enriched coarse spaces designed for high-contrast multiscale problems are used to represent the effects of heterogeneities of the media. Furthermore, rough source terms are captured via auxiliary correction equations that appear in the formulation of variational multiscale methods [23]. These auxiliary equations are localized and one can use additive or multiplicative constructions for the subgrid corrections as discussed in the current paper. Our preliminary numerical results show that one can capture the effects due to both spatial heterogeneities in the coefficients (such as permeability field) and source terms (e.g., due to singular well terms) in one iteration. We test the cases for both smooth source terms and rough source terms and show that with the multiplicative correction, the numerical approximations are more accurate compared to the additive correction. © 2010 Elsevier Ltd.

  3. Effects of long-term inhibition of neuronal nitric oxide synthase on blood pressure and renin release

    DEFF Research Database (Denmark)

    Ollerstam, A.; Skøtt, O.; Ek, J.

    2001-01-01

    Nitric oxide (NO) produced by neuronal NO-synthase (nNOS) in macula densa cells may be involved in the control of renin release. 7-Nitro indazole (7-NI) inhibits nNOS, and we investigated the effect of short- (4 days) and long-term (4 weeks) 7-NI treatment on blood pressure (BP), plasma renin...... LS rats (107 +/- 15 vs. 56 +/- 1 mGU mL(-1)). Stimulation of PRC in LS rats was further enhanced by 7-NI after 4 days of treatment, but not affected in rats treated for 4 weeks. This suggests that inhibition of nNOS stimulates renin release but that this stimulatory effect in the long run might...

  4. Estimated airborne release of radionuclides from the Battelle Memorial Institute Columbus Laboratories JN-1b building at the West Jefferson site as a result of postulated damage from severe wind and earthquake hazard

    International Nuclear Information System (INIS)

    Mishima, J.; Ayer, J.E.

    1981-11-01

    The potential airborne releases of radionuclides (source terms) that could result from wind and earthquake dmage are estimated for the Battelle Memorial Institute Columbus Laboratories JN-1b Building at the West Jefferson site in Ohio. The estimated source terms are based on the damage to barriers containing the radionuclides, the inventory of radionuclides at risk, and the fraction of the inventory made airborne as a result of the loss of containment. In an attempt to provide a realistic range of potential source terms that include most of the normal operating conditions, a best estimate bounded by upper and lower limits is calculated by combining the upper-bound, best-estimate, and lower-bound inventories-at-risk with an airborne release factor (upper-bound, best-estimate, and lower-bound if possible) for the situation. The factors used to evaluate the fractional airborne release of materials and the exchange rates between enclosed and exterior atmospheres are discussed. The postulated damage and source terms are discussed for wind and earthquake hazard scenarios in order of their increasing severity

  5. Inverse identification of unknown finite-duration air pollutant release from a point source in urban environment

    Science.gov (United States)

    Kovalets, Ivan V.; Efthimiou, George C.; Andronopoulos, Spyros; Venetsanos, Alexander G.; Argyropoulos, Christos D.; Kakosimos, Konstantinos E.

    2018-05-01

    In this work, we present an inverse computational method for the identification of the location, start time, duration and quantity of emitted substance of an unknown air pollution source of finite time duration in an urban environment. We considered a problem of transient pollutant dispersion under stationary meteorological fields, which is a reasonable assumption for the assimilation of available concentration measurements within 1 h from the start of an incident. We optimized the calculation of the source-receptor function by developing a method which requires integrating as many backward adjoint equations as the available measurement stations. This resulted in high numerical efficiency of the method. The source parameters are computed by maximizing the correlation function of the simulated and observed concentrations. The method has been integrated into the CFD code ADREA-HF and it has been tested successfully by performing a series of source inversion runs using the data of 200 individual realizations of puff releases, previously generated in a wind tunnel experiment.

  6. Source strength and dispersion of CO2 releases from high-pressure pipelines: CFD model using real gas equation of state

    International Nuclear Information System (INIS)

    Liu, Xiong; Godbole, Ajit; Lu, Cheng; Michal, Guillaume; Venton, Philip

    2014-01-01

    Highlights: • Validated CFD models for decompression and dispersion of CO 2 releases from pipelines. • Incorporation of real gas EOS into CFD code for source strength estimation. • Demonstration of better performance of SST k–ω turbulence model for jet flow. • Demonstration of better performance of real gas EOS compared to ideal gas EOS. • Demonstration of superiority of CFD models over a commercial risk assessment package. - Abstract: Transportation of CO 2 in high-pressure pipelines forms a crucial link in the ever-increasing application of Carbon Capture and Storage (CCS) technologies. An unplanned release of CO 2 from a pipeline presents a risk to human and animal populations and the environment. Therefore it is very important to develop a deeper understanding of the atmospheric dispersion of CO 2 before the deployment of CO 2 pipelines, to allow the appropriate safety precautions to be taken. This paper presents a two-stage Computational Fluid Dynamics (CFD) study developed (1) to estimate the source strength, and (2) to simulate the subsequent dispersion of CO 2 in the atmosphere, using the source strength estimated in stage (1). The Peng–Robinson (PR) EOS was incorporated into the CFD code. This enabled accurate modelling of the CO 2 jet to achieve more precise source strength estimates. The two-stage simulation approach also resulted in a reduction in the overall computing time. The CFD models were validated against experimental results from the British Petroleum (BP) CO 2 dispersion trials, and also against results produced by the risk management package Phast. Compared with the measurements, the CFD simulation results showed good agreement in both source strength and dispersion profile predictions. Furthermore, the effect of release direction on the dispersion was studied. The presented research provides a viable method for the assessment of risks associated with CCS

  7. A model for short and medium range dispersion of radionuclides released to the atmosphere

    International Nuclear Information System (INIS)

    Clarke, R.H.

    1979-09-01

    A Working Group was established to give practical guidance on the estimation of the dispersion of radioactive releases to the atmosphere. The dispersion is estimated in the short and medium range, that is from about 100 m to a few tens of kilometres from the source, and is based upon a Gaussian plume model. A scheme is presented for categorising atmospheric conditions and values of the associated dispersion parameters are given. Typical results are presented for releases in specific meteorological conditions and a scheme is included to allow for durations of release of up to 24 hours. Consideration has also been given to predicting longer term average concentrations, typically annual averages, and results are presented which facilitate site specific calculations. The results of the models are extended to 100 km from the source, but the increasing uncertainty with which results may be predicted beyond a few tens of kilometres from the source is emphasised. Three technical appendices provide some of the rationale behind the decisions made in adopting the various models in the proposed dispersion scheme. (author)

  8. Vaporization of mercury from molten lead droplets doped with mercury: Pb/Hg source term experiment for the APT/SILC target

    International Nuclear Information System (INIS)

    Tutu, N.K.; Greene, G.A.

    1994-09-01

    Experiments were performed to measure the fraction of mercury inventory released when droplets of molten lead, doped with a known concentration of mercury, fall through a controlled environment. The temperature of molten droplets ranged from 335 C to 346 C, and the concentration of mercury in the droplets ranged from 0.2 mass % to 1.0 mass %. The environment consisted of an air stream, at a temperature nominally equal to the melt temperature, and moving vertically upwards at a velocity of 10 cm/s. Direct observations and chemical analysis showed that no mercury was released from the molten droplets. Based upon the experimental results, it is concluded that no mercury vapor is likely to be released from the potentially molten source rod material in the APT-SILC Neutron Source Array to the confinement atmosphere during a postulated Large Break Loss Of Coolant Accident scenario leading to the melting of a fraction of the source rods

  9. CORONAL SOURCES, ELEMENTAL FRACTIONATION, AND RELEASE MECHANISMS OF HEAVY ION DROPOUTS IN THE SOLAR WIND

    Energy Technology Data Exchange (ETDEWEB)

    Weberg, Micah J. [PhD Candidate in Space Science, Department of Atmospheric, Oceanic, and Space Sciences, University of Michigan, 2134A Space Research Building, 2455 Hayward Street, Ann Arbor, MI 48109-2143, USA. (United States); Lepri, Susan T. [Associate Professor, Department of Atmospheric, Oceanic, and Space Sciences, University of Michigan, 2429 Space Research Building, 2455 Hayward Street, Ann Arbor, MI 48109-2143, USA. (United States); Zurbuchen, Thomas H., E-mail: mjweberg@umich.edu, E-mail: slepri@umich.edu, E-mail: thomasz@umich.edu [Professor, Space Science and Aerospace Engineering, Associate Dean for Entrepreneurship Senior Counselor of Entrepreneurship Education, Department of Atmospheric, Oceanic, and Space Sciences, University of Michigan, 2431 Space Research Building, 2455 Hayward Street, Ann Arbor, MI 48109-2143, USA. (United States)

    2015-03-10

    The elemental abundances of heavy ions (masses larger than He) in the solar wind provide information about physical processes occurring in the corona. Additionally, the charge state distributions of these heavy ions are sensitive to the temperature profiles of their respective source regions in the corona. Heavy ion dropouts are a relatively new class of solar wind events identified by both elemental and ionic charge state distributions. We have shown that their origins lie in large, closed coronal loops where processes such as gravitational settling dominate and can cause a mass-dependent fractionation pattern. In this study we consider and attempt to answer three fundamental questions concerning heavy ion dropouts: (1) 'where are the source loops located in the large-scale corona?'; (2) 'how does the interplay between coronal processes influence the end elemental abundances?'; and (3) 'what are the most probable release mechanisms'? We begin by analyzing the temporal and spatial variability of heavy ion dropouts and their correlation with heliospheric plasma and magnetic structures. Next we investigate the ordering of the elements inside dropouts with respect to mass, ionic charge state, and first ionization potential. Finally, we discuss these results in the context of the prevailing solar wind theories and the processes they posit that may be responsible for the release of coronal plasma into interplanetary space.

  10. CORONAL SOURCES, ELEMENTAL FRACTIONATION, AND RELEASE MECHANISMS OF HEAVY ION DROPOUTS IN THE SOLAR WIND

    International Nuclear Information System (INIS)

    Weberg, Micah J.; Lepri, Susan T.; Zurbuchen, Thomas H.

    2015-01-01

    The elemental abundances of heavy ions (masses larger than He) in the solar wind provide information about physical processes occurring in the corona. Additionally, the charge state distributions of these heavy ions are sensitive to the temperature profiles of their respective source regions in the corona. Heavy ion dropouts are a relatively new class of solar wind events identified by both elemental and ionic charge state distributions. We have shown that their origins lie in large, closed coronal loops where processes such as gravitational settling dominate and can cause a mass-dependent fractionation pattern. In this study we consider and attempt to answer three fundamental questions concerning heavy ion dropouts: (1) 'where are the source loops located in the large-scale corona?'; (2) 'how does the interplay between coronal processes influence the end elemental abundances?'; and (3) 'what are the most probable release mechanisms'? We begin by analyzing the temporal and spatial variability of heavy ion dropouts and their correlation with heliospheric plasma and magnetic structures. Next we investigate the ordering of the elements inside dropouts with respect to mass, ionic charge state, and first ionization potential. Finally, we discuss these results in the context of the prevailing solar wind theories and the processes they posit that may be responsible for the release of coronal plasma into interplanetary space

  11. Multivesicular release underlies short term synaptic potentiation independent of release probability change in the supraoptic nucleus.

    Directory of Open Access Journals (Sweden)

    Michelle E Quinlan

    Full Text Available Magnocellular neurons of the supraoptic nucleus receive glutamatergic excitatory inputs that regulate the firing activity and hormone release from these neurons. A strong, brief activation of these excitatory inputs induces a lingering barrage of tetrodotoxin-resistant miniature EPSCs (mEPSCs that lasts for tens of minutes. This is known to accompany an immediate increase in large amplitude mEPSCs. However, it remains unknown how long this amplitude increase can last and whether it is simply a byproduct of greater release probability. Using in vitro patch clamp recording on acute rat brain slices, we found that a brief, high frequency stimulation (HFS of afferents induced a potentiation of mEPSC amplitude lasting up to 20 min. This amplitude potentiation did not correlate with changes in mEPSC frequency, suggesting that it does not reflect changes in presynaptic release probability. Nonetheless, neither postsynaptic calcium chelator nor the NMDA receptor antagonist blocked the potentiation. Together with the known calcium dependency of HFS-induced potentiation of mEPSCs, our results imply that mEPSC amplitude increase requires presynaptic calcium. Further analysis showed multimodal distribution of mEPSC amplitude, suggesting that large mEPSCs were due to multivesicular glutamate release, even at late post-HFS when the frequency is no longer elevated. In conclusion, high frequency activation of excitatory synapses induces lasting multivesicular release in the SON, which is independent of changes in release probability. This represents a novel form of synaptic plasticity that may contribute to prolonged excitatory tone necessary for generation of burst firing of magnocellular neurons.

  12. Iodine chemistry effect on source term assessments. A MELCOR 186 YT study of a PWR severe accident sequence

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Otero, Bernadette

    2009-01-01

    Level-2 Probabilistic Safety Analysis has demonstrated to be a powerful tool to give insights into multiple aspects concerning severe accidents: phenomena with the greatest potential to lead to containment failure, safety systems performance and, even, to identify any additional accident management that could mitigate the consequences of such an even, etc. A major result of level-2 PSA is iodine content in Source Term since it is the main responsible for the radiological impact during the first few days after a hypothetical severe accident. Iodine chemistry is known to considerably affect iodine behavior and although understanding has improved substantially since the early 90's, a thorough understanding is still missing and most PSA studies do not address it when assessing severe accident scenarios. This paper emphasizes the quantitative and qualitative significance of considering iodine chemistry in level-2 PSA estimates. To do so a cold leg break, low pressure severe accident sequence of an actual pressurized water reactor has been analyzed with the MELCOR 1.8.6 YT code. Two sets of calculations, with and without chemistry, have been carried out and compared. The study shows that iodine chemistry could result in an iodine release to environment about twice higher, most of which would consist of around 60% of iodine in gaseous form. From these results it is concluded that exploratory studies on the potential effect of iodine chemistry on source term estimates should be carried out. (author)

  13. 75 FR 70021 - South Dakota Prairie Winds Project; Partial Term Relinquishment and Release of Easement for Wind...

    Science.gov (United States)

    2010-11-16

    ... requested financing for the project from the RUS. PW SD1 has also submitted an application to the Service to...] South Dakota Prairie Winds Project; Partial Term Relinquishment and Release of Easement for Wind Energy... impact statement (FEIS) on the South Dakota Prairie Winds Project issued by the Department of Energy's...

  14. Glass dissolution at 20, 40, 70 and 90 C: Short-term effects of solution chemistry and long-term Na release

    International Nuclear Information System (INIS)

    Bakel, A.J.; Ebert, W.L.; Strachan, D.M.

    1996-01-01

    The corrosion behavior of a borosilicate glass containing 20 mass 5 Na 2 O was assessed using static dissolution tests. This glass (LD6-5412) is representative of high Na glasses that may be used to stabilize Hanford low-level radioactive waste. The normalized mass loss (NL) decreases as NL(Na) ∼ NL(B) > NL(Si) in 20 and 40 C for tests conducted at glass surface area to leachant volume (S/V) ratio of 10 m -1 , and decreases as NL(Na) > NL(B) ∼ NL(Si) in 90 C tests conducted at 10 m -1 and in all tests conducted at higher S/V. The difference in the corrosion behavior is probably caused by the influence of dissolved glass components in the leachates. The NL(Na) is greater than the NL(B) or NL(Si) in all the tests conducted. Results from long-term tests at 2,000 m -1 show that the preferential release of Na persists for longer than one year at all temperatures and indicate that Na is released from this glass by an ion exchange process

  15. Short-term effects of beta-amyloid25-35 peptide aggregates on transmitter release in neuromuscular synapses.

    Science.gov (United States)

    Garcia, Neus; Santafé, Manel M; Tomàs, Marta; Lanuza, Maria A; Tomàs, Josep

    2008-03-01

    The beta-amyloid (AB) peptide25-35 contains the functional domain of the AB precursor protein that is both required for neurotrophic effects in normal neural tissues and is involved in the neurotoxic effects in Alzheimer disease. We demonstrated the presence of the amyloid precursor protein/AB peptide in intramuscular axons, presynaptic motor nerve terminals, terminal and myelinating Schwann cells, and the postsynaptic and subsarcolemmal region in the Levator auris longus muscle of adult rats by immunocytochemistry. Using intracellular recording, we investigated possible short-term functional effects of the AB fragment (0.1-10 micromol/L) on acetylcholine release in adult and newborn motor end plates. We found no change in evoked, spontaneous transmitter release or resting membrane potential of the muscle cells. A previous block of the presynaptic muscarinic receptor subtypes and a previous block or stimulation of protein kinase C revealed no masked effect of the peptide on the regulation of transmitter release. The aggregated form of AB peptide25-35, however, interfered acutely with acetylcholine release (quantal content reduction) when synaptic activity was maintained by electric stimulation. The possible relevance of this inhibition of neurotransmission by AB peptide25-35 to the pathogenesis of Alzheimer remains to be determined.

  16. THE 31 DEG{sup 2} RELEASE OF THE STRIPE 82 X-RAY SURVEY: THE POINT SOURCE CATALOG

    Energy Technology Data Exchange (ETDEWEB)

    LaMassa, Stephanie M.; Urry, C. Megan; Ananna, Tonima; Civano, Francesca; Marchesi, Stefano; Pecoraro, Robert [Yale Center for Astronomy and Astrophysics, Physics Department, P.O. Box 208120, New Haven, CT 06520 (United States); Cappelluti, Nico; Comastri, Andrea; Brusa, Marcella [INAF-Osservatorio Astronomico di Bologna, via Ranzani 1, I-40127 Bologna (Italy); Böhringer, Hans; Chon, Gayoung [Max-Planck-Institut für extraterrestrische Physik, D-85748 Garching (Germany); Glikman, Eilat [Department of Physics, Middlebury College, Middlebury, VT 05753 (United States); Richards, Gordon [Department of Physics, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Cardamone, Carie [Department of Math and Science, Wheelock College, 200 Riverway, Boston, MA 02215 (United States); Farrah, Duncan [Department of Physics MC 0435, Virginia Polytechnic Institute and State University, 850 West Campus Drive, Blacksburg, VA 24061 (United States); Gilfanov, Marat [Max-Planck Institut für Astrophysik, Karl-Schwarzschild-Str. 1, Postfach 1317, D-85741 Garching (Germany); Green, Paul [Smithsonian Astrophysical Observatory, 60 Garden Street, Cambridge, MA 02138 (United States); Komossa, S. [Max-Planck-Institut für Radioastronomie, Auf dem Hügel 69, D-53121 Bonn (Germany); Lira, Paulina [Departamento de Astronomia, Universidad de Chile, Camino del Observatorio 1515, Santiago (Chile); Makler, Martin [Centro Brasileiro de Pesquisas Fisicas, Rua Dr Xavier Sigaud 150, Rio de Janeiro, RJ 22290-180 (Brazil); and others

    2016-02-01

    We release the next installment of the Stripe 82 X-ray survey point-source catalog, which currently covers 31.3 deg{sup 2} of the Sloan Digital Sky Survey (SDSS) Stripe 82 Legacy field. In total, 6181 unique X-ray sources are significantly detected with XMM-Newton (>5σ) and Chandra (>4.5σ). This catalog release includes data from XMM-Newton cycle AO 13, which approximately doubled the Stripe 82X survey area. The flux limits of the Stripe 82X survey are 8.7 × 10{sup −16} erg s{sup −1} cm{sup −2}, 4.7 × 10{sup −15} erg s{sup −1} cm{sup −2}, and 2.1 × 10{sup −15} erg s{sup −1} cm{sup −2} in the soft (0.5–2 keV), hard (2–10 keV), and full bands (0.5–10 keV), respectively, with approximate half-area survey flux limits of 5.4 × 10{sup −15} erg s{sup −1} cm{sup −2}, 2.9 × 10{sup −14} erg s{sup −1} cm{sup −2}, and 1.7 × 10{sup −14} erg s{sup −1} cm{sup −2}. We matched the X-ray source lists to available multi-wavelength catalogs, including updated matches to the previous release of the Stripe 82X survey; 88% of the sample is matched to a multi-wavelength counterpart. Due to the wide area of Stripe 82X and rich ancillary multi-wavelength data, including coadded SDSS photometry, mid-infrared WISE coverage, near-infrared coverage from UKIDSS and VISTA Hemisphere Survey, ultraviolet coverage from GALEX, radio coverage from FIRST, and far-infrared coverage from Herschel, as well as existing ∼30% optical spectroscopic completeness, we are beginning to uncover rare objects, such as obscured high-luminosity active galactic nuclei at high-redshift. The Stripe 82X point source catalog is a valuable data set for constraining how this population grows and evolves, as well as for studying how they interact with the galaxies in which they live.

  17. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  18. Consequences of severe radioactive releases to Nordic Marine environment

    Energy Technology Data Exchange (ETDEWEB)

    Iosjpe, M. [Norwegian Radiation Protection Authority (NRPA) (Norway); Isaksson, M. [Univ. of Gothenburg (Sweden); Joensen, H.P. [Froskaparsetur Foeroya. Faroe Islands, Torshavn (Denmark); Lahtinen, J. [Radiation and Nuclear Safety Authority (Finland); Logemann, K. [Univ. of Iceland (Iceland); Palsson, S.E. [Geislavarnir Rikisins (Iceland); Roos, P. [Technical Univ. of Denmark. Risoe DTU, Roskilde (Denmark); Suolanen, V. [Technical Research Centre of Finland (Finland)

    2013-02-15

    In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW{sub th} reactor size were used. Based on source term analyses, the

  19. Consequences of severe radioactive releases to Nordic Marine environment

    International Nuclear Information System (INIS)

    Iosjpe, M.; Isaksson, M.; Joensen, H.P.; Lahtinen, J.; Logemann, K.; Palsson, S.E.; Roos, P.; Suolanen, V.

    2013-02-01

    In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW th reactor size were used. Based on source term analyses, the chosen

  20. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.M.

    1994-09-01

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  1. Realistic minimum accident source terms - Evaluation, application, and risk acceptance

    International Nuclear Information System (INIS)

    Angelo, P. L.

    2009-01-01

    The evaluation, application, and risk acceptance for realistic minimum accident source terms can represent a complex and arduous undertaking. This effort poses a very high impact to design, construction cost, operations and maintenance, and integrated safety over the expected facility lifetime. At the 2005 Nuclear Criticality Safety Division (NCSD) Meeting in Knoxville Tenn., two papers were presented mat summarized the Y-12 effort that reduced the number of criticality accident alarm system (CAAS) detectors originally designed for the new Highly Enriched Uranium Materials Facility (HEUMF) from 258 to an eventual as-built number of 60. Part of that effort relied on determining a realistic minimum accident source term specific to the facility. Since that time, the rationale for an alternate minimum accident has been strengthened by an evaluation process that incorporates realism. A recent update to the HEUMF CAAS technical basis highlights the concepts presented here. (authors)

  2. Black Carbon and Sulfate Aerosols in the Arctic: Long-term Trends, Radiative Impacts, and Source Attributions

    Science.gov (United States)

    Wang, H.; Zhang, R.; Yang, Y.; Smith, S.; Rasch, P. J.

    2017-12-01

    The Arctic has warmed dramatically in recent decades. As one of the important short-lived climate forcers, aerosols affect the Arctic radiative budget directly by interfering radiation and indirectly by modifying clouds. Light-absorbing particles (e.g., black carbon) in snow/ice can reduce the surface albedo. The direct radiative impact of aerosols on the Arctic climate can be either warming or cooling, depending on their composition and location, which can further alter the poleward heat transport. Anthropogenic emissions, especially, BC and SO2, have changed drastically in low/mid-latitude source regions in the past few decades. Arctic surface observations at some locations show that BC and sulfate aerosols had a decreasing trend in the recent decades. In order to understand the impact of long-term emission changes on aerosols and their radiative effects, we use the Community Earth System Model (CESM) equipped with an explicit BC and sulfur source-tagging technique to quantify the source-receptor relationships and decadal trends of Arctic sulfate and BC and to identify variations in their atmospheric transport pathways from lower latitudes. The simulation was conducted for 36 years (1979-2014) with prescribed sea surface temperatures and sea ice concentrations. To minimize potential biases in modeled large-scale circulations, wind fields in the simulation are nudged toward an atmospheric reanalysis dataset, while atmospheric constituents including water vapor, clouds, and aerosols are allowed to evolve according to the model physics. Both anthropogenic and open fire emissions came from the newly released CMIP6 datasets, which show strong regional trends in BC and SO2 emissions during the simulation time period. Results show that emissions from East Asia and South Asia together have the largest contributions to Arctic sulfate and BC concentrations in the upper troposphere, which have an increasing trend. The strong decrease in emissions from Europe, Russia and

  3. Self-Replenishing Vascularized Fouling-Release Surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Howell, C; Vu, TL; Lin, JJ; Kolle, S; Juthani, N; Watson, E; Weaver, JC; Alvarenga, J; Aizenberg, J

    2014-08-13

    Inspired by the long-term effectiveness of living antifouling materials, we have developed a method for the self-replenishment of synthetic biofouling-release surfaces. These surfaces are created by either molding or directly embedding 3D vascular systems into polydimethylsiloxane (PDMS) and filling them with a silicone oil to generate a nontoxic oil-infused material. When replenished with silicone oil from an outside source, these materials are capable of self-lubrication and continuous renewal of the interfacial fouling-release layer. Under accelerated lubricant loss conditions, fully infused vascularized samples retained significantly more lubricant than equivalent nonvascularized controls. Tests of lubricant-infused PDMS in static cultures of the infectious bacteria Staphylococcus aureus and Escherichia coli as well as the green microalgae Botryococcus braunii, Chlamydomonas reinhardtii, Dunaliella sauna, and Nannochloropsis oculata showed a significant reduction in biofilm adhesion compared to PDMS and glass controls containing no lubricant. Further experiments on vascularized versus nonvascularized samples that had been subjected to accelerated lubricant evaporation conditions for up to 48 h showed significantly less biofilm adherence on the vascularized surfaces. These results demonstrate the ability of an embedded lubricant-filled vascular network to improve the longevity of fouling-release surfaces.

  4. A review of ''The potential effects of release of krypton-85''

    International Nuclear Information System (INIS)

    Saunders, C.P.R.

    1993-06-01

    A Report into the potential effects of the release of Krypton-85 from the Thermal Oxide Reprocessing Plant (THORP) at Sellafield, was commissioned by British Nuclear Fuels Ltd. The report dealt with the atmospheric impact of the release of Krypton-85 and considered specifically the areas of atmospheric electricity, ionisation, aerosols and conductivity, in terms of the releases of Krypton-85 from THORP. This present review has three objectives: to consider the atmospheric effects of the proposed discharges from THORP, to comment upon the magnitude of the postulated effects from Krypton-85 discharges compared with those from other sources of radioactivity in the atmosphere, and, to comment on the conclusions drawn by the authors of the report. This review was conducted under the aegis of the United Kingdom Inspectorate of Pollution. (author)

  5. A novel integrated approach for the hazardous radioactive dust source terms estimation in future nuclear fusion power plants.

    Science.gov (United States)

    Poggi, L A; Malizia, A; Ciparisse, J F; Gaudio, P

    2016-10-01

    An open issue still under investigation by several international entities working on the safety and security field for the foreseen nuclear fusion reactors is the estimation of source terms that are a hazard for the operators and public, and for the machine itself in terms of efficiency and integrity in case of severe accident scenarios. Source term estimation is a crucial key safety issue to be addressed in the future reactors safety assessments, and the estimates available at the time are not sufficiently satisfactory. The lack of neutronic data along with the insufficiently accurate methodologies used until now, calls for an integrated methodology for source term estimation that can provide predictions with an adequate accuracy. This work proposes a complete methodology to estimate dust source terms starting from a broad information gathering. The wide number of parameters that can influence dust source term production is reduced with statistical tools using a combination of screening, sensitivity analysis, and uncertainty analysis. Finally, a preliminary and simplified methodology for dust source term production prediction for future devices is presented.

  6. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  7. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

  8. Fluoride release and recharge behavior of a nano-filled resin-modified glass ionomer compared with that of other fluoride releasing materials.

    Science.gov (United States)

    Mitra, Sumita B; Oxman, Joe D; Falsafi, Afshin; Ton, Tiffany T

    2011-12-01

    To compare the long-term fluoride release kinetics of a novel nano-filled two-paste resin-modified glass-ionomer (RMGI), Ketac Nano (KN) with that of two powder-liquid resin-modified glass-ionomers, Fuji II LC (FLC) and Vitremer (VT) and one conventional glass-ionomer, Fuji IX (FIX). Fluoride release was measured in vitro using ion-selective electrodes. Kinetic analysis was done using regression analysis and compared with existing models for GIs and compomers. In a separate experiment the samples of KN and two conventional glass-ionomers, FIX and Ketac Molar (KM) were subjected to a treatment with external fluoride source (Oral-B Neutra-Foam) after 3 months of fluoride release and the recharge behavior studied for an additional 7-day period. The cumulative amount of fluoride released from KN, VT and FLC and the release profiles were statistically similar but greater than that for FIX at P coating of KN with its primer and of DY with its adhesive did not significantly alter the fluoride release behavior of the respective materials. The overall rate for KN was significantly higher than for the compomer DY. DY showed a linear rate of release vs. t and no burst effect as expected for compomers. The nanoionomer KN showed fluoride recharge behavior similar to the conventional glass ionomers FIX and KM. Thus, it was concluded that the new RMGI KN exhibits fluoride ion release behavior similar to typical conventional and RMGIs and that the primer does not impede the release of fluoride.

  9. 135Cs/137Cs isotopic composition of environmental samples across Europe: Environmental transport and source term emission applications

    International Nuclear Information System (INIS)

    Snow, Mathew S.; Snyder, Darin C.

    2016-01-01

    135 Cs/ 137 Cs isotopic analyses represent an important tool for studying the fate and transport of radiocesium in the environment; in this work the 135 Cs/ 137 Cs isotopic composition in environmental samples taken from across Europe is reported. Surface soil and vegetation samples from western Russia, Ukraine, Austria, and Hungary show consistent aged thermal fission product 135 Cs/ 137 Cs isotope ratios of 0.58 ± 0.01 (age corrected to 1/1/15), with the exception of one sample of soil-moss from Hungary which shows an elevated 135 Cs/ 137 Cs ratio of 1.78 ± 0.12. With the exception of the outlier sample from Hungary, surface soil/vegetation data are in quantitative agreement with values previously reported for soils within the Chernobyl exclusion zone, suggesting that radiocesium at these locations is primarily composed of homogenous airborne deposition from Chernobyl. Seawater samples taken from the Irish Sea show 135 Cs/ 137 Cs isotope ratios of 1.22 ± 0.11 (age corrected to 1/1/15), suggesting aged thermal fission product Cs discharged from Sellafield. The differences in 135 Cs/ 137 Cs isotope ratios between Sellafield, Chernobyl, and global nuclear weapons testing fallout indicate that 135 Cs/ 137 Cs isotope ratios can be utilized to discriminate between and track radiocesium transport from different nuclear production source terms, including major emission sources in Europe. - Highlights: • 135 Cs/ 137 Cs useful for tracking anthropogenic environmental radiocesium releases. • European surface soils/vegetation have uniform ratio consistent with Chernobyl. • 135 Cs/ 137 Cs in Irish sea represents thermal fission ratio distinct from Chernobyl. • Can distinguish between major source terms in Europe based on 135 Cs/ 137 Cs.

  10. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  11. Modified ensemble Kalman filter for nuclear accident atmospheric dispersion: prediction improved and source estimated.

    Science.gov (United States)

    Zhang, X L; Su, G F; Yuan, H Y; Chen, J G; Huang, Q Y

    2014-09-15

    Atmospheric dispersion models play an important role in nuclear power plant accident management. A reliable estimation of radioactive material distribution in short range (about 50 km) is in urgent need for population sheltering and evacuation planning. However, the meteorological data and the source term which greatly influence the accuracy of the atmospheric dispersion models are usually poorly known at the early phase of the emergency. In this study, a modified ensemble Kalman filter data assimilation method in conjunction with a Lagrangian puff-model is proposed to simultaneously improve the model prediction and reconstruct the source terms for short range atmospheric dispersion using the off-site environmental monitoring data. Four main uncertainty parameters are considered: source release rate, plume rise height, wind speed and wind direction. Twin experiments show that the method effectively improves the predicted concentration distribution, and the temporal profiles of source release rate and plume rise height are also successfully reconstructed. Moreover, the time lag in the response of ensemble Kalman filter is shortened. The method proposed here can be a useful tool not only in the nuclear power plant accident emergency management but also in other similar situation where hazardous material is released into the atmosphere. Copyright © 2014 Elsevier B.V. All rights reserved.

  12. The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

    International Nuclear Information System (INIS)

    Starflinger, J.; Scholtyssek, W.; Unger, H.

    1994-12-01

    For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

  13. Use of source term code package in the ELEBRA MX-850 system

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.; Goes, A.G.A.

    1988-12-01

    The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.) [pt

  14. Semi-implicit and fully implicit shock-capturing methods for hyperbolic conservation laws with stiff source terms

    International Nuclear Information System (INIS)

    Yee, H.C.; Shinn, J.L.

    1986-12-01

    Some numerical aspects of finite-difference algorithms for nonlinear multidimensional hyperbolic conservation laws with stiff nonhomogenous (source) terms are discussed. If the stiffness is entirely dominated by the source term, a semi-implicit shock-capturing method is proposed provided that the Jacobian of the source terms possesses certain properties. The proposed semi-implicit method can be viewed as a variant of the Bussing and Murman point-implicit scheme with a more appropriate numerical dissipation for the computation of strong shock waves. However, if the stiffness is not solely dominated by the source terms, a fully implicit method would be a better choice. The situation is complicated by problems that are higher than one dimension, and the presence of stiff source terms further complicates the solution procedures for alternating direction implicit (ADI) methods. Several alternatives are discussed. The primary motivation for constructing these schemes was to address thermally and chemically nonequilibrium flows in the hypersonic regime. Due to the unique structure of the eigenvalues and eigenvectors for fluid flows of this type, the computation can be simplified, thus providing a more efficient solution procedure than one might have anticipated

  15. Semi-implicit and fully implicit shock-capturing methods for hyperbolic conservation laws with stiff source terms

    International Nuclear Information System (INIS)

    Yee, H.C.; Shinn, J.L.

    1987-01-01

    Some numerical aspects of finite-difference algorithms for nonlinear multidimensional hyperbolic conservation laws with stiff nonhomogeneous (source) terms are discussed. If the stiffness is entirely dominated by the source term, a semi-implicit shock-capturing method is proposed provided that the Jacobian of the source terms possesses certain properties. The proposed semi-implicit method can be viewed as a variant of the Bussing and Murman point-implicit scheme with a more appropriate numerical dissipation for the computation of strong shock waves. However, if the stiffness is not solely dominated by the source terms, a fully implicit method would be a better choice. The situation is complicated by problems that are higher than one dimension, and the presence of stiff source terms further complicates the solution procedures for alternating direction implicit (ADI) methods. Several alternatives are discussed. The primary motivation for constructing these schemes was to address thermally and chemically nonequilibrium flows in the hypersonic regime. Due to the unique structure of the eigenvalues and eigenvectors for fluid flows of this type, the computation can be simplified, thus providing a more efficient solution procedure than one might have anticipated. 46 references

  16. Reconstruction of Atmospheric Tracer Releases with Optimal Resolution Features: Concentration Data Assimilation

    Science.gov (United States)

    Singh, Sarvesh Kumar; Turbelin, Gregory; Issartel, Jean-Pierre; Kumar, Pramod; Feiz, Amir Ali

    2015-04-01

    function which exhibits a priori information about the unknown releases apparent to the monitoring network. The properties of the weight function provide an optimal data resolution and model resolution to the retrieved source estimates. The retrieved source estimates are proved theoretically to be stable against the random measurement errors and their reliability can be interpreted in terms of the distribution of the weight functions. Further, the same framework can be extended for the identification of the point type releases by utilizing the maximum of the retrieved source estimates. The inversion technique has been evaluated with the several diffusion experiments, like, Idaho low wind diffusion experiment (1974), IIT Delhi tracer experiment (1991), European Tracer Experiment (1994), Fusion Field Trials (2007), etc. In case of point release experiments, the source parameters are mostly retrieved close to the true source parameters with least error. Primarily, the proposed technique overcomes two major difficulties incurred in the source reconstruction: (i) The initialization of the source parameters as required by the optimization based techniques. The converged solution depends on their initialization. (ii) The statistical knowledge about the measurement and background errors as required by the Bayesian inference based techniques. These are hypothetically assumed in case of no prior knowledge.

  17. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L.

    2007-01-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  18. Evaluation of Long-term Performance of Enhanced Anaerobic Source Zone Bioremediation using mass flux

    Science.gov (United States)

    Haluska, A.; Cho, J.; Hatzinger, P.; Annable, M. D.

    2017-12-01

    Chlorinated ethene DNAPL source zones in groundwater act as potential long term sources of contamination as they dissolve yielding concentrations well above MCLs, posing an on-going public health risk. Enhanced bioremediation has been applied to treat many source zones with significant promise, but long-term sustainability of this technology has not been thoroughly assessed. This study evaluated the long-term effectiveness of enhanced anaerobic source zone bioremediation at chloroethene contaminated sites to determine if the treatment prevented contaminant rebound and removed NAPL from the source zone. Long-term performance was evaluated based on achieving MCL-based contaminant mass fluxes in parent compound concentrations during different monitoring periods. Groundwater concertation versus time data was compiled for 6-sites and post-remedial contaminant mass flux data was then measured using passive flux meters at wells both within and down-gradient of the source zone. Post-remedial mass flux data was then combined with pre-remedial water quality data to estimate pre-remedial mass flux. This information was used to characterize a DNAPL dissolution source strength function, such as the Power Law Model and the Equilibrium Stream tube model. The six-sites characterized for this study were (1) Former Charleston Air Force Base, Charleston, SC; (2) Dover Air Force Base, Dover, DE; (3) Treasure Island Naval Station, San Francisco, CA; (4) Former Raritan Arsenal, Edison, NJ; (5) Naval Air Station, Jacksonville, FL; and, (6) Former Naval Air Station, Alameda, CA. Contaminant mass fluxes decreased for all the sites by the end of the post-treatment monitoring period and rebound was limited within the source zone. Post remedial source strength function estimates suggest that decreases in contaminant mass flux will continue to occur at these sites, but a mass flux based on MCL levels may never be exceeded. Thus, site clean-up goals should be evaluated as order

  19. Variational Iterative Refinement Source Term Estimation Algorithm Assessment for Rural and Urban Environments

    Science.gov (United States)

    Delle Monache, L.; Rodriguez, L. M.; Meech, S.; Hahn, D.; Betancourt, T.; Steinhoff, D.

    2016-12-01

    It is necessary to accurately estimate the initial source characteristics in the event of an accidental or intentional release of a Chemical, Biological, Radiological, or Nuclear (CBRN) agent into the atmosphere. The accurate estimation of the source characteristics are important because many times they are unknown and the Atmospheric Transport and Dispersion (AT&D) models rely heavily on these estimates to create hazard assessments. To correctly assess the source characteristics in an operational environment where time is critical, the National Center for Atmospheric Research (NCAR) has developed a Source Term Estimation (STE) method, known as the Variational Iterative Refinement STE algorithm (VIRSA). VIRSA consists of a combination of modeling systems. These systems include an AT&D model, its corresponding STE model, a Hybrid Lagrangian-Eulerian Plume Model (H-LEPM), and its mathematical adjoint model. In an operational scenario where we have information regarding the infrastructure of a city, the AT&D model used is the Urban Dispersion Model (UDM) and when using this model in VIRSA we refer to the system as uVIRSA. In all other scenarios where we do not have the city infrastructure information readily available, the AT&D model used is the Second-order Closure Integrated PUFF model (SCIPUFF) and the system is referred to as sVIRSA. VIRSA was originally developed using SCIPUFF 2.4 for the Defense Threat Reduction Agency and integrated into the Hazard Prediction and Assessment Capability and Joint Program for Information Systems Joint Effects Model. The results discussed here are the verification and validation of the upgraded system with SCIPUFF 3.0 and the newly implemented UDM capability. To verify uVIRSA and sVIRSA, synthetic concentration observation scenarios were created in urban and rural environments and the results of this verification are shown. Finally, we validate the STE performance of uVIRSA using scenarios from the Joint Urban 2003 (JU03

  20. Radon releases from Australian uranium mining and milling projects: assessing the UNSCEAR approach.

    Science.gov (United States)

    Mudd, Gavin M

    2008-02-01

    The release of radon gas and progeny from the mining and milling of uranium-bearing ores has long been recognised as a potential radiological health hazard. The standards for exposure to radon and progeny have decreased over time as the understanding of their health risk has improved. In recent years there has been debate on the long-term releases (10,000 years) of radon from uranium mining and milling sites, focusing on abandoned, operational and rehabilitated sites. The primary purpose has been estimates of the radiation exposure of both local and global populations. Although there has been an increasing number of radon release studies over recent years in the USA, Australia, Canada and elsewhere, a systematic evaluation of this work has yet to be published in the international literature. This paper presents a detailed compilation and analysis of Australian studies. In order to quantify radon sources, a review of data on uranium mining and milling wastes in Australia, as they influence radon releases, is presented. An extensive compilation of the available radon release data is then assembled for the various projects, including a comparison to predictions of radon behaviour where available. An analysis of cumulative radon releases is then developed and compared to the UNSCEAR approach. The implications for the various assessments of long-term releases of radon are discussed, including aspects such as the need for ongoing monitoring of rehabilitation at uranium mining and milling sites and life-cycle accounting.

  1. Radon releases from Australian uranium mining and milling projects: assessing the UNSCEAR approach

    International Nuclear Information System (INIS)

    Mudd, Gavin M.

    2008-01-01

    The release of radon gas and progeny from the mining and milling of uranium-bearing ores has long been recognised as a potential radiological health hazard. The standards for exposure to radon and progeny have decreased over time as the understanding of their health risk has improved. In recent years there has been debate on the long-term releases (10,000 years) of radon from uranium mining and milling sites, focusing on abandoned, operational and rehabilitated sites. The primary purpose has been estimates of the radiation exposure of both local and global populations. Although there has been an increasing number of radon release studies over recent years in the USA, Australia, Canada and elsewhere, a systematic evaluation of this work has yet to be published in the international literature. This paper presents a detailed compilation and analysis of Australian studies. In order to quantify radon sources, a review of data on uranium mining and milling wastes in Australia, as they influence radon releases, is presented. An extensive compilation of the available radon release data is then assembled for the various projects, including a comparison to predictions of radon behaviour where available. An analysis of cumulative radon releases is then developed and compared to the UNSCEAR approach. The implications for the various assessments of long-term releases of radon are discussed, including aspects such as the need for ongoing monitoring of rehabilitation at uranium mining and milling sites and life-cycle accounting

  2. Impact of human milk pasteurization on the kinetics of peptide release during in vitro dynamic term newborn digestion.

    Science.gov (United States)

    Deglaire, Amélie; De Oliveira, Samira C; Jardin, Julien; Briard-Bion, Valérie; Emily, Mathieu; Ménard, Olivia; Bourlieu, Claire; Dupont, Didier

    2016-07-01

    Holder pasteurization (62.5°C, 30 min) ensures sanitary quality of donor's human milk but also denatures beneficial proteins. Understanding whether this further impacts the kinetics of peptide release during gastrointestinal digestion of human milk was the aim of the present paper. Mature raw (RHM) or pasteurized (PHM) human milk were digested (RHM, n = 2; PHM, n = 3) by an in vitro dynamic system (term stage). Label-free quantitative peptidomics was performed on milk and digesta (ten time points). Ascending hierarchical clustering was conducted on "Pasteurization × Digestion time" interaction coefficients. Preproteolysis occurred in human milk (159 unique peptides; RHM: 91, PHM: 151), mostly on β-casein (88% of the endogenous peptides). The predicted cleavage number increased with pasteurization, potentially through plasmin activation (plasmin cleavages: RHM, 53; PHM, 76). During digestion, eight clusters resumed 1054 peptides from RHM and PHM, originating for 49% of them from β-casein. For seven clusters (57% of peptides), the kinetics of peptide release differed between RHM and PHM. The parent protein was significantly linked to the clustering (p-value = 1.4 E-09), with β-casein and lactoferrin associated to clusters in an opposite manner. Pasteurization impacted selectively gastric and intestinal kinetics of peptide release in term newborns, which may have further nutritional consequences. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  3. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  4. Release of fission products from irradiated SRP fuels at elevated temperature. Data report on the first stage of the SRP source term study

    International Nuclear Information System (INIS)

    Woodley, R.E.

    1986-06-01

    For a sound evaluation of the consequences of a hypothetical nuclear reactor accident, a knowledge of the extent of fission product release from the fuel at anticipated temperatures and atmosphere conditions is required. Measurements of fission product release have been performed with a variety of nuclear fuels under various conditions of temperature and atmosphere. While the use of data obtained on fuels similar to the fuel of interest may provide a reasonable estimate of release fractions, precise information of this nature can only be obtained from measurements employing specimens of the actual fuels used in the nuclear reactor under consideration. The two fuels of interest in the present study are an alloy, a dispersion of UAl 4 in an aluminum matrix, and a cermet, a dispersion of U 3 O 8 in an aluminum matrix. Both fuels are clad in aluminum

  5. Potential off-normal events and associated radiological source terms for the compact ignition tokamak: Fusion Safety Program

    International Nuclear Information System (INIS)

    Holland, D.F.; Lyon, R.E.

    1987-10-01

    The Compact Ignition Tokamak (CIT), the latest step in the United States program to develop the commercial application of fusion power, is designed as the first fusion device to achieve ignition conditions. It is to be constructed near Princeton, New Jersey on the site of the existing Tokamak Fusion Test Reactor (TFTR). To address the environmental impact and public safety concerns, a preliminary analysis was performed of potential off-normal radiological releases. Operational occurrences, natural phenomena, accidents with external origins, and accidents external to the PPPL site were considered as potential sources for off-normal events. Based on an initial screening, events were selected for preliminary analysis. Included in these events were tritium releases from the tritium delivery and recovery system, tritium releases from the torus, releases of activated nitrogen from the test cell or cryostat, seismic events, and shipping accidents. In each case, the design considerations related to the event were reviewed and the release scenarios discussed. Because of the complexity of some of the proposed safety systems, in some cases event trees were used to describe the accident scenarios. For each scenario, the probability was estimated as well as the release magnitude, isotope, chemical form, and release mode. 10 refs., 17 figs., 5 tabs

  6. Unsplit schemes for hyperbolic conservation laws with source terms in one space dimension

    International Nuclear Information System (INIS)

    Papalexandris, M.V.; Leonard, A.; Dimotakis, P.E.

    1997-01-01

    The present work is concerned with an application of the theory of characteristics to conservation laws with source terms in one space dimension, such as the Euler equations for reacting flows. Space-time paths are introduced on which the flow/chemistry equations decouple to a characteristic set of ODE's for the corresponding homogeneous laws, thus allowing the introduction of functions analogous to the Riemann invariants in classical theory. The geometry of these paths depends on the spatial gradients of the solution. This particular decomposition can be used in the design of efficient unsplit algorithms for the numerical integration of the equations. As a first step, these ideas are implemented for the case of a scalar conservation law with a nonlinear source term. The resulting algorithm belongs to the class of MUSCL-type, shock-capturing schemes. Its accuracy and robustness are checked through a series of tests. The stiffness of the source term is also studied. Then, the algorithm is generalized for a system of hyperbolic equations, namely the Euler equations for reacting flows. A numerical study of unstable detonations is performed. 57 refs

  7. Testing IH Instrumentation: Analysis of 1996-1998 Tank Ventilation Data in Terms of Characterizing a Transient Release

    International Nuclear Information System (INIS)

    Droppo, James G.

    2004-01-01

    An analysis is conducted of the 1996-1998 Hanford tank ventilation studies of average ventilation rates to help define characteristics of shorter term releases. This effort is being conducted as part of the design of tests of Industrial Hygiene's (IH) instrumentation ability to detect transient airborne plumes from tanks using current deployment strategies for tank operations. This analysis has improved our understanding of the variability of hourly average tank ventilation processes. However, the analysis was unable to discern the relative importance of emissions due to continuous releases and short-duration bursts of material. The key findings are as follows: (1) The ventilation of relatively well-sealed, passively ventilated tanks appears to be driven by a combination of pressure, buoyancy, and wind influences. The results of a best-fit analysis conducted with a single data set provide information on the hourly emission variability that IH instrumentation will need to detect. (2) Tank ventilation rates and tank emission rates are not the same. The studies found that the measured infiltration rates for a single tank are often a complex function of air exchanges between tanks and air exchanges with outdoor air. This situation greatly limits the usefulness of the ventilation data in defining vapor emission rates. (3) There is no evidence in the data to discern if the routine tank vapor releases occur over a short time (i.e., a puff) or over an extended time (i.e., continuous releases). Based on this analysis of the tank ventilation studies, it is also noted that (1) the hourly averaged emission peaks from the relatively well-sealed passively-vented tanks (such as U-103) are not a simple function of one meteorological parameter--but the peaks often are the result of the coincidence of temporal maximums in pressure, temperature, and wind influences and (2) a mechanistic combination modeling approach and/or field studies may be necessary to understand the short-term

  8. Challenges in defining a radiologic and hydrologic source term for underground nuclear test centers, Nevada Test Site, Nye County, Nevada

    International Nuclear Information System (INIS)

    Smith, D.K.

    1995-06-01

    The compilation of a radionuclide inventory for long-lived radioactive contaminants residual from nuclear testing provides a partial measure of the radiologic source term at the Nevada Test Site. The radiologic source term also includes potentially mobile short-lived radionuclides excluded from the inventory. The radiologic source term for tritium is known with accuracy and is equivalent to the hydrologic source term within the saturated zone. Definition of the total hydrologic source term for fission and activation products that have high activities for decades following underground testing involves knowledge and assumptions which are presently unavailable. Systematic investigation of the behavior of fission products, activation products and actinides under saturated or Partially saturated conditions is imperative to define a representative total hydrologic source term. This is particularly important given the heterogeneous distribution of radionuclides within testing centers. Data quality objectives which emphasize a combination of measurements and credible estimates of the hydrologic source term are a priority for near-field investigations at the Nevada Test Site

  9. Least-squares finite-element method for shallow-water equations with source terms

    Institute of Scientific and Technical Information of China (English)

    Shin-Jye Liang; Tai-Wen Hsu

    2009-01-01

    Numerical solution of shallow-water equations (SWE) has been a challenging task because of its nonlinear hyperbolic nature, admitting discontinuous solution, and the need to satisfy the C-property. The presence of source terms in momentum equations, such as the bottom slope and friction of bed, compounds the difficulties further. In this paper, a least-squares finite-element method for the space discretization and θ-method for the time integration is developed for the 2D non-conservative SWE including the source terms. Advantages of the method include: the source terms can be approximated easily with interpolation functions, no upwind scheme is needed, as well as the resulting system equations is symmetric and positive-definite, therefore, can be solved efficiently with the conjugate gradient method. The method is applied to steady and unsteady flows, subcritical and transcritical flow over a bump, 1D and 2D circular dam-break, wave past a circular cylinder, as well as wave past a hump. Computed results show good C-property, conservation property and compare well with exact solutions and other numerical results for flows with weak and mild gradient changes, but lead to inaccurate predictions for flows with strong gradient changes and discontinuities.

  10. Inverse kinetics method with source term for subcriticality measurements during criticality approach in the IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

    2009-01-01

    In reactor physics tests which are performed at the startup after refueling the commercial PWRs, it is important to monitor subcriticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor and it is possible to perform an online reactivity measurement based on the point reactor kinetics equations. This technique is successful applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels, the contribution of the neutron source must be taken into account and this implies the knowledge of a quantity proportional to the source strength, and then it should be determined. Some experiments have been performed in the IPEN/MB-01 Research Reactor for the determination of the Source Term, using the Least Square Inverse Kinetics Method (LSIKM). A digital reactivity meter which neglects the source term is used to calculate the reactivity and then the source term can be determined by the LSIKM. After determining the source term, its value can be added to the algorithm and the reactivity can be determined again, considering the source term. The new digital reactivity meter can be used now to monitor reactivity during the criticality approach and the measured value for the reactivity is more precise than the meter which neglects the source term. (author)

  11. Modelling the environmental transport of tritium in the vicinity of long term atmospheric and sub-surface sources. Report of the Tritium Working Group of the Biosphere Modelling and Assessment (BIOMASS) Programme, Theme 3

    International Nuclear Information System (INIS)

    2003-03-01

    The IAEA Programme on BIOsphere Modelling and ASSessment (BIOMASS) was launched in Vienna in October 1996. The programme was concerned with developing and improving capabilities to predict the transfer of radionuclides in the environment. The programme had three themes: Theme 1: Radioactive Waste Disposal. The objective was to develop the concept of a standard or reference biosphere for application to the assessment of the long term safety of repositories for radioactive waste. Theme 2: Environmental Releases. BIOMASS provided an international forum for activities aimed at increasing the confidence in methods and models for the assessment of radiation exposure related to environmental releases. Two Working Groups addressed issues concerned with the reconstruction of radiation doses received by people from past releases of radionuclides to the environment and the evaluation of the efficacy of remedial measures. Theme 3: Biosphere Processes. The aim of this Theme was to improve capabilities for modelling the transfer of radionuclides in particular parts of the biosphere identified as being of potential radiological significance and where there were gaps in modelling approaches. This topic was explored using a range of methods including reviews of the literature, model inter-comparison exercises and, where possible, model testing against independent sources of data. Three Working Groups were established to examine the modelling of: (1) long term tritium dispersion in the environment; (2) radionuclide uptake by fruits; and (3) radionuclide migration and accumulation in forest ecosystems. This report describes results of the studies undertaken by the Tritium Working Group under Theme 3. It discusses the six model test exercises (three model-model intercomparisons and three model-data exercises) and the twenty-month field sampling programme undertaken to investigate the environmental transport of tritium in the vicinity of long term atmospheric and sub-surface sources

  12. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations. Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report

  13. Unclassified Source Term and Radionuclide Data for Corrective Action Unit 98: Frenchman Flat Nevada Test Site, Nevada, Rev. No.: 0

    Energy Technology Data Exchange (ETDEWEB)

    Farnham, Irene

    2005-09-01

    Frenchman Flat is one of several areas of the Nevada Test Site (NTS) used for underground nuclear testing (Figure 1-1). These nuclear tests resulted in groundwater contamination in the vicinity of the underground test areas. As a result, the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) is currently conducting a corrective action investigation (CAI) of the Frenchman Flat underground test areas. Since 1996, the Nevada Division of Environmental Protection (NDEP) has regulated NNSA/NSO corrective actions through the ''Federal Facility Agreement and Consent Order'' ([FFACO], 1996). Appendix VI of the FFACO agreement, ''Corrective Action Strategy'', was revised on December 7, 2000, and describes the processes that will be used to complete corrective actions, including those in the Underground Test Area (UGTA) Project. The individual locations covered by the agreement are known as corrective action sites (CASs), which are grouped into corrective action units (CAUs). The UGTA CASs are grouped geographically into five CAUs: Frenchman Flat, Central Pahute Mesa, Western Pahute Mesa, Yucca Flat/Climax Mine, and Rainier Mesa/Shoshone Mountain (Figure 1-1). These CAUs have distinctly different contaminant source, geologic, and hydrogeologic characteristics related to their location (FFACO, 1996). The Frenchman Flat CAU consists of 10 CASs located in the northern part of Area 5 and the southern part of Area 11 (Figure 1-1). This report documents the evaluation of the information and data available on the unclassified source term and radionuclide contamination for Frenchman Flat, CAU 98. The methodology used to estimate hydrologic source terms (HSTs) for the Frenchman Flat CAU is also documented. The HST of an underground nuclear test is the portion of the total inventory of radionuclides that is released over time into the groundwater following the test. The total residual inventory

  14. Long-term dust aerosol production from natural sources in Iceland.

    Science.gov (United States)

    Dagsson-Waldhauserova, Pavla; Arnalds, Olafur; Olafsson, Haraldur

    2017-02-01

    Iceland is a volcanic island in the North Atlantic Ocean with maritime climate. In spite of moist climate, large areas are with limited vegetation cover where >40% of Iceland is classified with considerable to very severe erosion and 21% of Iceland is volcanic sandy deserts. Not only do natural emissions from these sources influenced by strong winds affect regional air quality in Iceland ("Reykjavik haze"), but dust particles are transported over the Atlantic ocean and Arctic Ocean >1000 km at times. The aim of this paper is to place Icelandic dust production area into international perspective, present long-term frequency of dust storm events in northeast Iceland, and estimate dust aerosol concentrations during reported dust events. Meteorological observations with dust presence codes and related visibility were used to identify the frequency and the long-term changes in dust production in northeast Iceland. There were annually 16.4 days on average with reported dust observations on weather stations within the northeastern erosion area, indicating extreme dust plume activity and erosion within the northeastern deserts, even though the area is covered with snow during the major part of winter. During the 2000s the highest occurrence of dust events in six decades was reported. We have measured saltation and Aeolian transport during dust/volcanic ash storms in Iceland, which give some of the most intense wind erosion events ever measured. Icelandic dust affects the ecosystems over much of Iceland and causes regional haze. It is likely to affect the ecosystems of the oceans around Iceland, and it brings dust that lowers the albedo of the Icelandic glaciers, increasing melt-off due to global warming. The study indicates that Icelandic dust may contribute to the Arctic air pollution. Long-term records of meteorological dust observations from Northeast Iceland indicate the frequency of dust events from Icelandic deserts. The research involves a 60-year period and

  15. Phosphate release upon short- and long-term flooding of fen meadows depends on land use history and pH.

    NARCIS (Netherlands)

    Beltman, Boudewijn; van der Ven, P.J.M.; Verhoeven, Jos; Sarneel, J

    2014-01-01

    Flooding of acidified and desiccated fen meadows is a management approach for mitigating loss of plant species as well as a short-term measure to prevent flooding in urban areas. Studies have shown that flooding events can cause extreme P release from soils. We questioned whether the occurrence of

  16. Conceptual model for deriving the repository source term

    International Nuclear Information System (INIS)

    Alexander, D.H.; Apted, M.J.; Liebetrau, A.M.; Doctor, P.G.; Williford, R.E.; Van Luik, A.E.

    1984-11-01

    Part of a strategy for evaluating the compliance of geologic repositories with federal regulations is a modeling approach that would provide realistic release estimates for a particular configuration of the engineered-barrier system. The objective is to avoid worst-case bounding assumptions that are physically impossible or excessively conservative and to obtain probabilistic estimates of (1) the penetration time for metal barriers and (2) radionuclide-release rates for individually simulated waste packages after penetration has occurred. The conceptual model described in this paper will assume that release rates are explicitly related to such time-dependent processes as mass transfer, dissolution and precipitation, radionuclide decay, and variations in the geochemical environment. The conceptual model will take into account the reduction in the rates of waste-form dissolution and metal corrosion due to a buildup of chemical reaction products. The sorptive properties of the metal-barrier corrosion products in proximity to the waste form surface will also be included. Cumulative releases from the engineered-barrier system will be calculated by summing the releases from a probabilistically generated population of individual waste packages. 14 refs., 7 figs

  17. Conceptual model for deriving the repository source term

    International Nuclear Information System (INIS)

    Alexander, D.H.; Apted, M.J.; Liebetrau, A.M.; Van Luik, A.E.; Williford, R.E.; Doctor, P.G.; Pacific Northwest Lab., Richland, WA; Roy F. Weston, Inc./Rogers and Assoc. Engineering Corp., Rockville, MD)

    1984-01-01

    Part of a strategy for evaluating the compliance of geologic repositories with Federal regulations is a modeling approach that would provide realistic release estimates for a particular configuration of the engineered-barrier system. The objective is to avoid worst-case bounding assumptions that are physically impossible or excessively conservative and to obtain probabilitistic estimates of (1) the penetration time for metal barriers and (2) radionuclide-release rates for individually simulated waste packages after penetration has occurred. The conceptual model described in this paper will assume that release rates are explicitly related to such time-dependent processes as mass transfer, dissolution and precipitation, radionuclide decay, and variations in the geochemical environment. The conceptual model will take into account the reduction in the rates of waste-form dissolution and metal corrosion due to a buildup of chemical reaction products. The sorptive properties of the metal-barrier corrosion products in proximity to the waste form surface will also be included. Cumulative released from the engineered-barrier system will be calculated by summing the releases from a probabilistically generated population of individual waste packages. 14 refs., 7 figs

  18. Effect of MK-801 on methamphetamine-induced dopaminergic neurotoxicity: long-term attenuation of methamphetamine-induced dopamine release

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Eun; Kim, Yu Ri; Hwang, Se Hwan [Sungkyunkwan Univ., School of Medicine, Seoul (Korea, Republic of)

    2001-08-01

    Repeated administration of methamphetamine (METH) produces high extracellular levels of dopamine (DA) and subsequent striatal DA terminal damage. The effect of MK-801, a noncompetitive N-methyl-D-aspartate receptor antagonist, on METH-induced changes in DA transporter (DAT) and DA release evoked by an acute METH challenge was evaluated in rodent striatum using [{sup 3}H] WIN 38,428 ex vivo auto-radiography and in vivo microdialysis. Four injections of METH (10 mg/kg, i.p.), each given 2 h apart, produced 71% decrease in DAT levels in mouse striatum 3 d after administration. Pretreatment with MK-801 (2.5 g/kg, i.p.) 15 min before each of the four METH injections protected completely against striatal DAT depletions. Four injections of MK-801 alone did not significantly change striatal DAT levels. Striatal DA release evoked by an acute METH challenge (4mg/kg, i.p.) at 3 d after repeated administration of METH in rats was decreased but significant compared with controls, which was attenuated by repeated pretreatment with MK-801. Also, repeated injections of MK-801 alone attenuated acute METH-induced striatal DA release 3 d after administration. These results suggest that repeated administration of MK-801 may exert a preventive effect against METH-induced DA terminal injury through long-term attenuation of DA release induced by METH and other stimuli.

  19. Effect of MK-801 on methamphetamine-induced dopaminergic neurotoxicity: long-term attenuation of methamphetamine-induced dopamine release

    International Nuclear Information System (INIS)

    Kim, Sang Eun; Kim, Yu Ri; Hwang, Se Hwan

    2001-01-01

    Repeated administration of methamphetamine (METH) produces high extracellular levels of dopamine (DA) and subsequent striatal DA terminal damage. The effect of MK-801, a noncompetitive N-methyl-D-aspartate receptor antagonist, on METH-induced changes in DA transporter (DAT) and DA release evoked by an acute METH challenge was evaluated in rodent striatum using [ 3 H] WIN 38,428 ex vivo auto-radiography and in vivo microdialysis. Four injections of METH (10 mg/kg, i.p.), each given 2 h apart, produced 71% decrease in DAT levels in mouse striatum 3 d after administration. Pretreatment with MK-801 (2.5 g/kg, i.p.) 15 min before each of the four METH injections protected completely against striatal DAT depletions. Four injections of MK-801 alone did not significantly change striatal DAT levels. Striatal DA release evoked by an acute METH challenge (4mg/kg, i.p.) at 3 d after repeated administration of METH in rats was decreased but significant compared with controls, which was attenuated by repeated pretreatment with MK-801. Also, repeated injections of MK-801 alone attenuated acute METH-induced striatal DA release 3 d after administration. These results suggest that repeated administration of MK-801 may exert a preventive effect against METH-induced DA terminal injury through long-term attenuation of DA release induced by METH and other stimuli

  20. Source-term characterisation and solid speciation of plutonium at the Semipalatinsk NTS, Kazakhstan.

    Science.gov (United States)

    Nápoles, H Jiménez; León Vintró, L; Mitchell, P I; Omarova, A; Burkitbayev, M; Priest, N D; Artemyev, O; Lukashenko, S

    2004-01-01

    New data on the concentrations of key fission/activation products and transuranium nuclides in samples of soil and water from the Semipalatinsk Nuclear Test Site are presented and interpreted. Sampling was carried out at Ground Zero, Lake Balapan, the Tel'kem craters and reference locations within the test site boundary well removed from localised sources. Radionuclide ratios have been used to characterise the source term(s) at each of these sites. The geochemical partitioning of plutonium has also been examined and it is shown that the bulk of the plutonium contamination at most of the sites examined is in a highly refractory, non-labile form.

  1. Source-term characterisation and solid speciation of plutonium at the Semipalatinsk NTS, Kazakhstan

    Energy Technology Data Exchange (ETDEWEB)

    Napoles, H.J.H. Jimenez; Leon Vintro, L. E-mail: luis.leon@ucd.ie; Mitchell, P.I.; Omarova, A.; Burkitbayev, M.; Priest, N.D.; Artemyev, O.; Lukashenko, S

    2004-09-01

    New data on the concentrations of key fission/activation products and transuranium nuclides in samples of soil and water from the Semipalatinsk Nuclear Test Site are presented and interpreted. Sampling was carried out at Ground Zero, Lake Balapan, the Tel'kem craters and reference locations within the test site boundary well removed from localised sources. Radionuclide ratios have been used to characterise the source term(s) at each of these sites. The geochemical partitioning of plutonium has also been examined and it is shown that the bulk of the plutonium contamination at most of the sites examined is in a highly refractory, non-labile form.

  2. Fission product source term research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1985-01-01

    The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed

  3. Taking inventory on VOC releases from Amoco's Yorktown refinery

    International Nuclear Information System (INIS)

    Klee, H.H. Jr.; Schmitt, R.E.; Harrass, M.C.; Podar, M.K.

    1996-01-01

    Amoco's Yorktown, Virginia, refinery is a 35-year-old, 53,000 bbl/day facility that manufacturers gasoline, heating oil, liquid petroleum gas, sulfur, and coke. In a cooperative and voluntary effort, Amoco Corporation and the US Environmental Protection Agency conducted a joint project to study pollution prevention opportunities at an operating industrial facility. Source reduction efforts--key to pollution prevention strategies--require knowledge of specific sources of releases. However, data on releases from individual process units are limited in favor of data to monitor existing end-of-pipe pollution control requirements. The study's sampling program sought to portray the distribution of releases within the refinery, their management within the refinery, and ultimate releases leaving the refinery. Subsequent tests of blowdown stack and fugitive emissions further improved total release estimates. The initial study estimated that the refinery generates about 25,000 metric tons (t)/year of potential pollutants. Of these, about half are released from the refinery as airborne, waterborne, or land-disposed releases. Airborne releases comprise the majority of releases by mass, about 12,000 t/year. Most of the airborne releases are volatile organic compound hydrocarbons. The inventory sampling project and subsequent work identified differences with Toxic Release Inventory (TRI) values and standard emission factors (AP-42). The inventory and other data provided an opportunity to consider options for, and limitations of, specific pollution prevention or source reduction strategies

  4. Hubble Space Telescope: a Vision to 2020 and Beyond: The Hubble Source Catalog

    Science.gov (United States)

    Strolger, Louis-Gregory

    2016-01-01

    The Hubble Source Catalog (HSC) is an initiative centered on what science would be enabled by a master catalog of all the sources HST has imaged over its lifetime. The first version of this catalog was released in early 2015, and included approximately 30 million sources from archived direct imaging with WFPC2, ACS (through 2011), and WFC3 (to 2014). Version 2, scheduled for release in early 2016, will feed off the Hubble Legacy Archive DR9 release, updating the ACS sources with more detections, and more direct imaging, through to mid-2015. This talk will overview the properties and goals of the HSC in terms of its source detection, object resolution, confusion limits, and overall astrometric and photometric precision. I will also discuss the connections to other MAST activities (e.g., the Discovery Portal interface), to STScI and user products (e.g., the Spectroscopic Catalog and High-Level Science Products), and to community resources (e.g., Pan-STARRS, SDSS, and eventually GAIA). The HSC successfully amalgamates the diverse observations with HST, and despite the limitations in uniformity on the sky, will be an important reference for JWST, LSST, and other future telescopes.

  5. Proposal for implementation of alternative source term in the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Bazan L, A.; Lopez L, M.; Vargas A, A.; Cardenas J, J. B.

    2009-10-01

    In 2010 the nuclear power plant of Laguna Verde will implement the extended power upbeat in both units of the plant. Agree with methodology of NEDC-33004P-A, (constant pressure power up rate), and the source term of core, for accidents evaluations, were increased in proportion to the ratio of power level. This means that for the case of a design basis accident of loss of coolant an increase of power of 15% originated an increase of 15% in dose to main control room. Using the method of NEDC-33004P-A to extended power upbeat conditions was determined that the dose value to main control room is very near to regulatory limit established by SRP 6.4. By the above and in order to recover the margin, the nuclear power plant of Laguna Verde will calculate an alternative source term following the criteria established in RG 1.183 (alternative radiological source term for evaluating DBA at nuclear power reactor). This approach also have a more realistic dose value using the criterion of 10-CFR-50.67, in addition is predicted to get the benefit of additional operational flexibilities. This paper present the proposal of implementing the alternative source term in Laguna Verde. (Author)

  6. Assessment on health and energy sources

    International Nuclear Information System (INIS)

    Acket, C.; Yvon, M.

    2013-01-01

    After having recalled some issues related to the prevention of environmental health risks and mentioned in the preparation of the debate on energy transition in France, this document gathers actual objective elements for an assessment of health impact of the different energy sources. It discusses the impacts on health (mortality, sicknesses and diseases) of fossil fuels (coal and its wastes, gas), of renewable energies, of nuclear energy. For this last one, the document outlines the lack of documentation for various topics, discusses some results published on the dose impact of nuclear operation, and comment the issue of waste storage. It also recalls the main accidents (Three Mile Island, Chernobyl, and Fukushima) and some of the known and assessed impacts. The third part proposes comparisons between the different energy sources in terms of deadly accidents, of pollution and greenhouse effect (current and late mortality), of released radioactivity (release sources and collective dose). In conclusion, the authors outline that the impact on health of environmental risks must be one of the essential issues for the definition of energy policy, and discuss the resulting implications. Various data are provided in appendix: energy in France and in the world, origins of radioactivity

  7. Ongoing environmental monitoring and assessment of the long-term impacts of the February 2014 radiological release from the waste isolation pilot plant.

    Science.gov (United States)

    Thakur, Punam; Runyon, Tim

    2018-04-09

    Three years ago, the Waste Isolation Pilot Plant (WIPP) experienced its first minor accident involving a radiological release. Late in the evening on February 14, 2014, a waste container in the repository underwent a chemical reaction that caused the container to overheat and breach, releasing its contents into the underground. Following a lengthy recovery process, the facility recently resumed waste disposal operations. The accident released significant levels of radioactivity into the disposal room and adjacent exhaust drifts, and although no one was present in the underground at the time of the release, a total of 22 workers tested positive for very low level of radiation, presumably from some of the radioactive material that was released above ground through a small leak in the HEPA filtration system. The dominant radionuclides released were 241 Am and 239 + 240 Pu in a ratio that matched the content of the drum from Los Alamos National Laboratory (LANL) that was eventually identified as the breached container. From the air particulate monitoring and plume modeling, it was concluded that the dose, at the nearest location accessible to the general public, from this radiation release event would have been less than 0.01 mSv (< 1 mrem/year). This level is well below the 0.1 mSv/year (10 mrem/year) regulatory limit for DOE facilities established by the US Environmental Protection Agency (EPA).While no long-term impacts to public health or the environment are expected as a result of the WIPP radiation release, the limited ventilation and residual contamination levels in the underground are still a concern and pose a major challenge for the full recovery of WIPP. This article provides an up-to-date overview of environmental monitoring results through the WIPP recovery and an estimate of the long-term impacts of the accident on the natural and human environment.

  8. Review of the accident source terms for aluminide fuel: Application to the BR2 reactor

    International Nuclear Information System (INIS)

    Joppen, F.

    2005-01-01

    A major safety review of the BR2, a material test reactor, is to be conducted for the year 2006. One of the subjects selected for the safety review is the definition of source terms for emergency planning and in particular the development of accident scenarios. For nuclear power plants the behaviour of fuel under accident conditions is a well studied object. In case of non-power reactors this basic knowledge is rather scarce. The usefulness of information from power plant fuels is limited due to the differences in fuel type, power level and thermohydraulical conditions. First investigation indicates that using data from power plant fuel leads to an overestimation of the source terms. Further research on this subject could be very useful for the research reactor community, in order to define more realistic source terms and to improve the emergency preparedness. (author)

  9. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  10. The long-term relationships among China's energy consumption sources and adjustments to its renewable energy policy

    International Nuclear Information System (INIS)

    Zou Gaolu

    2012-01-01

    To reduce its consumption of coal and oil in its primary energy consumption, China promotes the development of renewable energy resources. I have analysed the long-term relationship among China's primary energy consumption sources. Changes in coal consumption lead those in the consumption of other energy sources in the long term. Coal and oil fuels substitute for each other equally. The long-term elasticities of China's coal consumption relative to its hydroelectricity consumption were greater than one and nearly equal during the two sample periods. Therefore, increased hydroelectricity consumption did not imply a reduction in coal consumption. China holds abundant hydroelectricity, wind and, solar energy potential. China must prevent an excessive escalation of its economy and resultant energy demand to realise a meaningful substitution of coal with hydroelectricity. Moreover, China must develop and use wind and solar energy sources. Natural gas can be a good substitute for coal, given its moderate price growth and affordable price levels. - Highlights: ► Coal consumption changes lead those of other energy sources in the long term. ► Coal and oil fuels substitute for each other equally. ► Increased hydroelectricity consumption has not meant lower coal consumption. ► Wind, solar and natural gas are China's promising energy sources.

  11. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  12. Observations of the release of non-methane hydrocarbons from fractured shale.

    Science.gov (United States)

    Sommariva, Roberto; Blake, Robert S; Cuss, Robert J; Cordell, Rebecca L; Harrington, Jon F; White, Iain R; Monks, Paul S

    2014-01-01

    The organic content of shale has become of commercial interest as a source of hydrocarbons, owing to the development of hydraulic fracturing ("fracking"). While the main focus is on the extraction of methane, shale also contains significant amounts of non-methane hydrocarbons (NMHCs). We describe the first real-time observations of the release of NMHCs from a fractured shale. Samples from the Bowland-Hodder formation (England) were analyzed under different conditions using mass spectrometry, with the objective of understanding the dynamic process of gas release upon fracturing of the shale. A wide range of NMHCs (alkanes, cycloalkanes, aromatics, and bicyclic hydrocarbons) are released at parts per million or parts per billion level with temperature- and humidity-dependent release rates, which can be rationalized in terms of the physicochemical characteristics of different hydrocarbon classes. Our results indicate that higher energy inputs (i.e., temperatures) significantly increase the amount of NMHCs released from shale, while humidity tends to suppress it; additionally, a large fraction of the gas is released within the first hour after the shale has been fractured. These findings suggest that other hydrocarbons of commercial interest may be extracted from shale and open the possibility to optimize the "fracking" process, improving gas yields and reducing environmental impacts.

  13. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  14. Impact assessment of radionuclides released to environment from the European Spallation Source

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, S.; Andersson, K. [Technical University of Denmark (Denmark); Ene, D. [European Spallation Soure AB - ESS, Lund (Sweden)

    2014-07-01

    The European Spallation Source (ESS) is a large science and technology infrastructure project currently under construction in Lund, Sweden, with operation planned by 2019. The facility design and construction includes a linear proton accelerator, a heavy-metal target station, neutron instruments, laboratories, and a data management and software development centre. During operation the ESS will produce a wide range of radionuclides via spallation and activation processes. Radiological assessments are needed to ensure that operational discharges and releases from potential incidents/accidents are within acceptable limits. The spectrum of radionuclides produced at ESS is quite different from that produced in nuclear power plants and assessment work has therefore been challenged by lack of information on less well-known radionuclides. Traditional assessment methodologies have been applied focusing on releases to air and public sewer systems and calculating radiation doses to representative persons living in and near Lund close to the ESS site. Exposure pathways considered include external radiation from radionuclides in air, external radiation from radionuclides deposited on ground and skin, inhalation of radionuclides and ingestion of locally produced contaminated food. Atmospheric dispersion has been simulated with the Gaussian plume model which is considered adequate within a few kilometres. Effects of release height have been investigated and site specific values of other parameters such as wind speed, wind direction, rain fall etc. have been used. Contamination of food has been calculated from the (ECOSYS) food dose model used in the RODOS and ARGOS decision support systems. The food dose model does not contain specific data for a number of ESS relevant radionuclides, e.g. {sup 7}Be, {sup 32}P and {sup 35}S. The data required include mobility of these isotopes, soil-to-plant concentration ratios and equilibrium transfer factors of daily intake by ingestion of meat

  15. Polyester Textiles as a Source of Microplastics from Households: A Mechanistic Study to Understand Microfiber Release During Washing.

    Science.gov (United States)

    Hernandez, Edgar; Nowack, Bernd; Mitrano, Denise M

    2017-06-20

    Microplastic fibers make up a large proportion of microplastics found in the environment, especially in urban areas. There is good reason to consider synthetic textiles a major source of microplastic fibers, and it will not diminish since the use of synthetic fabrics, especially polyester, continues to increase. In this study we provide quantitative data regarding the size and mass of microplastic fibers released from synthetic (polyester) textiles during simulated home washing under controlled laboratory conditions. Consideration of fabric structure and washing conditions (use of detergents, temperature, wash duration, and sequential washings) allowed us to study the propensity of fiber shedding in a mechanistic way. Thousands of individual fibers were measured (number, length) from each wash solution to provide a robust data set on which to draw conclusions. Among all the variables tested, the use of detergent appeared to affect the total mass of fibers released the most, yet the detergent composition (liquid or powder) or overdosing of detergent did not significantly influence microplastic release. Despite different release quantities due to the addition of a surfactant (approximately 0.025 and 0.1 mg fibers/g textile washed, without and with detergent, respectively), the overall microplastic fiber length profile remained similar regardless of wash condition or fabric structure, with the vast majority of fibers ranging between 100 and 800 μm in length irrespective of wash cycle number. This indicates that the fiber staple length and/or debris encapsulated inside the fabric from the yarn spinning could be directly responsible for releasing stray fibers. This study serves as a first look toward understanding the physical properties of the textile itself to better understand the mechanisms of fiber shedding in the context of microplastic fiber release into laundry wash water.

  16. Volatile organic compounds released from Microcystis flos-aquae under nitrogen sources and their toxic effects on Chlorella vulgaris.

    Science.gov (United States)

    Xu, Qinghuan; Yang, Lin; Yang, Wangting; Bai, Yan; Hou, Ping; Zhao, Jingxian; Zhou, Lv; Zuo, Zhaojiang

    2017-01-01

    Eutrophication promotes massive growth of cyanobacteria and algal blooms, which can poison other algae and reduce biodiversity. To investigate the differences in multiple nitrogen (N) sources in eutrophicated water on the emission of volatile organic compounds (VOCs) from cyanobacteria, and their toxic effects on other algal growth, we analyzed VOCs emitted from Microcystis flos-aquae with different types and concentrations of nitrogen, and determined the effects under Normal-N and Non-N conditions on Chlorella vulgaris. M. flos-aquae released 27, 22, 20, 27, 19, 25 and 17 compounds, respectively, with NaNO 3 , NaNO 2 , NH 4 Cl, urea, Ser, Lys and Arg as the sole N source. With the reduction in N amount, the emission of VOCs was increased markedly, and the most VOCs were found under Non-N condition. C. vulgaris cell propagation, photosynthetic pigment and Fv/Fm declined significantly following exposure to M. flos-aquae VOCs under Non-N condition, but not under Normal-N condition. When C. vulgaris cells were treated with two terpenoids, eucalyptol and limonene, the inhibitory effects were enhanced with increasing concentrations. Therefore, multiple N sources in eutrophicated water induce different VOC emissions from cyanobacteria, and reduction in N can cause nutrient competition, which can result in emissions of more VOCs. Those VOCs released from M. flos-aquae cells under Non-N for nutrient competition can inhibit other algal growth. Among those VOCs, eucalyptol and limonene are the major toxic agents. Copyright © 2016 Elsevier Inc. All rights reserved.

  17. Analysis of safety information for nuclear power plants and development of source term estimation program

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Choi, Seong Soo; Park, Jin Hee

    1999-12-01

    Current CARE(Computerized Advisory System for Radiological Emergency) in KINS(Korea Institute of Nuclear Safety) has no STES(Source Term Estimation System) which links between SIDS(Safety Information Display System) and FADAS(Following Accident Dose Assessment System). So in this study, STES is under development. STES system is the system that estimates the source term based on the safety information provided by SIDS. Estimated source term is given to FADAS as an input for estimation of environmental effect of radiation. Through this first year project STES for the Kori 3,4 and Younggwang 1,2 has been developed. Since there is no CARE for Wolsong(PHWR) plants yet, CARE for Wolsong is under construction. The safety parameters are selected and the safety information display screens and the alarm logic for plant status change are developed for Wolsong Unit 2 based on the design documents for CANDU plants

  18. The ALPS project release 2.0: open source software for strongly correlated systems

    International Nuclear Information System (INIS)

    Bauer, B; Gamper, L; Gukelberger, J; Hehn, A; Isakov, S V; Ma, P N; Mates, P; Carr, L D; Evertz, H G; Feiguin, A; Freire, J; Koop, D; Fuchs, S; Gull, E; Guertler, S; Igarashi, R; Matsuo, H; Parcollet, O; Pawłowski, G; Picon, J D

    2011-01-01

    We present release 2.0 of the ALPS (Algorithms and Libraries for Physics Simulations) project, an open source software project to develop libraries and application programs for the simulation of strongly correlated quantum lattice models such as quantum magnets, lattice bosons, and strongly correlated fermion systems. The code development is centered on common XML and HDF5 data formats, libraries to simplify and speed up code development, common evaluation and plotting tools, and simulation programs. The programs enable non-experts to start carrying out serial or parallel numerical simulations by providing basic implementations of the important algorithms for quantum lattice models: classical and quantum Monte Carlo (QMC) using non-local updates, extended ensemble simulations, exact and full diagonalization (ED), the density matrix renormalization group (DMRG) both in a static version and a dynamic time-evolving block decimation (TEBD) code, and quantum Monte Carlo solvers for dynamical mean field theory (DMFT). The ALPS libraries provide a powerful framework for programmers to develop their own applications, which, for instance, greatly simplify the steps of porting a serial code onto a parallel, distributed memory machine. Major changes in release 2.0 include the use of HDF5 for binary data, evaluation tools in Python, support for the Windows operating system, the use of CMake as build system and binary installation packages for Mac OS X and Windows, and integration with the VisTrails workflow provenance tool. The software is available from our web server at http://alps.comp-phys.org/

  19. Modelling for radiological and radioecological consequences of an accidental radionuclide release at Sea

    International Nuclear Information System (INIS)

    Paschoa, A.S.

    1997-01-01

    Full text: Scenarios concerning accidental releases of radionuclides into water bodies can be found in the open literature, mostly in connection with nuclear power plants located either onshore or inland. However, meager attention has been given to nuclear reactors used as energy sources for propulsion at sea, which are also subject to accidents. Such potential accidents may involve the loss of part of the reactor core to the surrounding water body. In addition of the initial instantaneous releases, one can estimate delayed source terms based on the rate at which radionuclides are dissolved or leached from the solidified material, like part of the core or structural materials in contact with water. Most of such solidified material might be a mixture of uranium, zirconium, iron, calcium, silica, fission and activation products, and transuranium elements as oxides, forming a glassy type solid. Transport models were used to calculate radionuclide concentrations in water resulting from short and delayed source terms. Oceanographic data used in the calculations were taken either from the open literature or from unclassified reports of the Brazilian Navy, being, however, as generic as possible. Time-dependent concentration functions for radionuclides in aquatic food following an accidental release reflect the net result of intake and elimination processes. However, to avoid the complexities of multiple parameters involved in such processes, the model accounts only for trophic transfer of radionuclides, and yet avoids the necessity of analyzing the details of each transfer step used to determine fish, crustacea, molluscs and seaweed accumulation. Swimming and other aquatic sports are not included in the model used for dose calculations because of theirs relatively low importance in comparison with the pathways concerning ingestion of aquatic food

  20. A computer code to estimate accidental fire and radioactive airborne releases in nuclear fuel cycle facilities: User's manual for FIRIN

    International Nuclear Information System (INIS)

    Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.

    1989-02-01

    This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs

  1. Oxygenated gasoline release in the unsaturated zone - Part 1: Source zone behavior.

    Science.gov (United States)

    Freitas, Juliana G; Barker, James F

    2011-11-01

    Oxygenates present in gasoline, such as ethanol and MTBE, are a concern in subsurface contamination related to accidental spills. While gasoline hydrocarbon compounds have low solubility, MTBE and ethanol are more soluble, ethanol being completely miscible with water. Consequently, their fate in the subsurface is likely to differ from that of gasoline. To evaluate the fate of gasoline containing oxygenates following a release in the unsaturated zone shielded from rainfall/recharge, a controlled field test was performed at Canadian Forces Base Borden, in Ontario. 200L of a mixture composed of gasoline with 10% ethanol and 4.5% MTBE was released in the unsaturated zone, into a trench 20cm deep, about 32cm above the water table. Based on soil cores, most of the ethanol was retained in the source, above the capillary fringe, and remained there for more than 100 days. Ethanol partitioned from the gasoline to the unsaturated pore-water and was retained, despite the thin unsaturated zone at the site (~35cm from the top of the capillary fringe to ground surface). Due to its lower solubility, most of the MTBE remained within the NAPL as it infiltrated deeper into the unsaturated zone and accumulated with the gasoline on top of the depressed capillary fringe. Only minor changes in the distribution of ethanol were noted following oscillations in the water table. Two methods to estimate the capacity of the unsaturated zone to retain ethanol are explored. It is clear that conceptual models for sites impacted by ethanol-fuels must consider the unsaturated zone. Copyright © 2011 Elsevier B.V. All rights reserved.

  2. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A.

    1990-08-01

    This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs

  3. Modeling of release of radionuclides from an engineered disposal facility for shallow-land disposal of low-level radioactive wastes

    International Nuclear Information System (INIS)

    Matsuzuru, H.; Suzuki, A.

    1989-01-01

    The computer code, ENBAR-1, for the simulation of radionuclide releases from an engineered disposal facility has been developed to evaluate the source term for subsequent migration of radionuclides in and through a natural barrier. The system considered here is that a waste package (waste form and container) is placed, together with backfill materials, into a concrete pit as a disposal unit for shallow-land disposal of low-level radioactive wastes. The code developed includes the following modules: water penetration into a concrete pit, corrosion of a drum as a container, leaching of radionuclides from a waste form, migration of radionuclides in backfill materials, release of radionuclides from the pit. The code has the advantage of its simplicity of operation and presentation while still allowing comprehensive evaluation of each element of an engineered disposal facility to be treated. The performance and source term of the facility might be readily estimated with a few key parameters to define the problem

  4. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Energy Technology Data Exchange (ETDEWEB)

    Pappas, Charalampos, E-mail: chpappas@ipta.demokritos.gr; Ikonomopoulos, Andreas; Sfetsos, Athanasios; Andronopoulos, Spyros; Varvayanni, Melpomeni; Catsaros, Nicolas

    2014-07-01

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed.

  5. Source term and activation calculations for the new TR-FLEX cyclotron for medical applications at HZDR

    Energy Technology Data Exchange (ETDEWEB)

    Konheiser, Joerg [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety; Ferrari, A. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. of Radiation Physics; Magin, A. [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany); Naumann, B. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Dept. of Radiation Protection and Safety; Mueller, S.E.

    2017-06-01

    The neutron source terms for a proton beam hitting an {sup 18}O-enriched water target were calculated with the radiation transport programs MCNP6 and FLUKA and were compared to source terms for exclusive {sup 18}O(p,n){sup 18}F production. To validate the radiation fields obtained in the simulations, an experimental program has been started using activation samples originally used in reactor dosimetry.

  6. Dissolution and Release of Gaseous Nitrogen (N2, N2O) in the Source Region of the Yellow River

    Science.gov (United States)

    Zhang, L.; Xia, X.; Wang, J.

    2017-12-01

    Nitrogen is an important biogenic element. The migration and transformation of nitrogen in rivers is an important process affecting global nitrogen cycling and greenhouse gas emissions. However, there is a lack of research on nitrogen removal and greenhouse gas emission characteristics of high altitude rivers. In this work, the spatial and temporal variations of dissolved nitrogen (N2 and N2O) concentrations, saturation, and release flux as well as their responses to environmental factors were studied in the Yellow River source area, a typical high altitude river. The results showed that the dissolved concentrations of N2 and N2O in the rivers were 8.24-137.75 μmol.L-1 and 2.57-31.94 nmol.L-1, respectively. N2 and N2O saturation were greater than 100% for all the sampling sites, indicating that the river is a release source for atmosphere N2 and N2O. Correspondingly, the fluxes of N2 and N2O from river water to atmosphere were 24.12-1606.57 mmol (m2.d) -1 and 12.96-276.81 μmol (m2.d) -1, respectively. Generally, the dissolution concentration and release flux of N2 and N2O in July were larger than that in May. The concentrations of N2 and N2O in river water were related to the environmental factors, and the dissolved concentration of N2 in the surface water was significantly positively correlated with water temperature, NH4+-N and total inorganic nitrogen (DIN) (p<0.01). The dissolved concentration of N2O was significantly positively correlated with the content of suspended particulates, DO, and DIN (p<0.01). Thus, DIN is a key factor in the process of N2 and N2O formation. This study can help to understand the nitrogen cycling in high-altitude rivers and provide basic data for a comprehensive assessment of global river nitrogen loss. Key Words: Source Region of the Yellow River; Gaseous Nitrogen; Nitrogen loss; High altitude river

  7. RELOS.MOD2: a code system for the determination of instationary fission product releases from molten pools

    International Nuclear Information System (INIS)

    Kortz, Ch.; Koch, M.K.; Unger, H.; Funke, F.

    1999-01-01

    For the assessment of molten corium pool source terms, a mechanistic model has been developed to describe the transport of fission products from liquid corium pool surfaces into a colder gas atmosphere. Modelling is based on an approach for diffusive and convective transport processes coupled with thermochemical equilibrium considerations enabling detailed speciation analyses of the fission products released. Both have been implemented into the code system RELOS.MOD2. RELOS.MOD2 sensitivity calculations on possible effects of anticipated uncertainties in the thermo-chemical data on the fission product release predictions are presented. (author)

  8. Phosphate Release Upon Long- and Short -Term Flooding of Fen Meadows Depends on Land Use History and Soil pH

    NARCIS (Netherlands)

    Beltman, B.; Van Der Ven, P. J. M.; Verhoeven, J. T. A.; Sarneel, J. M.

    2014-01-01

    Flooding of acidified and desiccated fen meadows is a management approach for mitigating loss of plant species as well as a short-term measure to prevent flooding in urban areas. Studies have shown that flooding events can cause extreme P release from soils. We questioned whether the occurrence of

  9. Long term population dose due to radon (Rn-222) released from uranium mill tailings

    International Nuclear Information System (INIS)

    Chambers, D.B.; Lower, L.M.; Stager, R.H.

    2001-01-01

    The results of a study undertaken by the European Commission on the external costs (environmental and social) of various energy production systems is likely to be influential in determining how the European Union will develop its energy supply systems. The estimated costs for nuclear power from the study will be based on the findings of the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), with the costs being dominated by the estimated long term (10,000 y) population doses due to radon (Rn-222) released from mill tailings. UNSCEAR developed a central estimate of 150 person-Sv per GW y and a range of 1 to 1,000 person-Sv per GW y. However, the generic data available to and being used by UNSCEAR are dated and are not appropriate for the current and planned future conditions in the uranium production industry, with the result that the estimated external costs of nuclear power (specifically, the doses due to radon emitted from mill tailings) are overestimated. The Uranium Institute sponsored a study to estimate long term population doses based on the most recent 1993 UNSCEAR methodology, but using data that would be more appropriate to the current major uranium production facilities. Site-specific information obtained from the owners/operators and the Uranium Institute included: present and proposed tailings management plans; tailings volumes and areas; ore grades and reserves; measurements and estimates of radon emission rates; and population densities. Tailings at closed facilities that no longer contribute to uranium production were not evaluated since it was assumed that these radon sources need not be considered in evaluating the external costs of current and future nuclear power production. Based on the same approach as UNSCEAR, but using a more sophisticated air dispersion model, and more site-specific data relative to existing sites and proposed tailings management practices, radon emission rates and population densities (that

  10. An adaptive Bayesian inference algorithm to estimate the parameters of a hazardous atmospheric release

    Science.gov (United States)

    Rajaona, Harizo; Septier, François; Armand, Patrick; Delignon, Yves; Olry, Christophe; Albergel, Armand; Moussafir, Jacques

    2015-12-01

    In the eventuality of an accidental or intentional atmospheric release, the reconstruction of the source term using measurements from a set of sensors is an important and challenging inverse problem. A rapid and accurate estimation of the source allows faster and more efficient action for first-response teams, in addition to providing better damage assessment. This paper presents a Bayesian probabilistic approach to estimate the location and the temporal emission profile of a pointwise source. The release rate is evaluated analytically by using a Gaussian assumption on its prior distribution, and is enhanced with a positivity constraint to improve the estimation. The source location is obtained by the means of an advanced iterative Monte-Carlo technique called Adaptive Multiple Importance Sampling (AMIS), which uses a recycling process at each iteration to accelerate its convergence. The proposed methodology is tested using synthetic and real concentration data in the framework of the Fusion Field Trials 2007 (FFT-07) experiment. The quality of the obtained results is comparable to those coming from the Markov Chain Monte Carlo (MCMC) algorithm, a popular Bayesian method used for source estimation. Moreover, the adaptive processing of the AMIS provides a better sampling efficiency by reusing all the generated samples.

  11. Distribution and source of 129I, 239,240Pu, 137Cs in the environment of Lithuania

    DEFF Research Database (Denmark)

    Ezerinskis, Z.; Hou, Xiaolin; Druteikiene, R.

    2016-01-01

    ratio revealed that the source of 129I in Lithuania is global fallout in most cases though several sampling sites shows a possible impact of reprocessing releases. Estimated 129I/131I ratio in soil samples from the southern part of Lithuania shows negligible input of the Chernobyl fallout....... No correlation of the 137Cs and Pu isotopes with 129I was observed, indicating their different sources terms. Results demonstrate uneven distribution of these radionuclides in the Lithuanian territory and several sources of contamination i.e. Chernobyl accident, reprocessing releases and global fallout....

  12. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  13. Use of source term uncoupled in radionuclide migration equations

    International Nuclear Information System (INIS)

    Silveira, Claudia Siqueira da; Lima, Zelmo Rodrigues de; Alvim, Antonio Carlos Marques

    2008-01-01

    Final repositories of high-level radioactive waste have been considered in deep, low permeability and stable geological formations. A common problem found is the migration modeling of radionuclides in a fractured rock. In this work, the physical system adopted consists of the rock matrix containing a single planar fracture situated in water saturated porous rock. The partial differential equations that describe the radionuclide transport were discretized using finite differences techniques, of which the following methods were adopted: Explicit Euler, Implicit Euler and Crank-Nicholson. For each one of these methods, the advective term was discretized with the following numerical schemes: backward differences, centered differences and forward differences. We make a comparison to determine which temporal and space discretization has the best result in relation to a reference solution. The obtained results show that the Explicit Euler Method with forward discretization in the advective term has a good accuracy. Next, with the objective of improving the answer of the Implicit Euler and Crank-Nicholson Methods it was accomplished a source term uncouplement, the diffusive flux. The obtained results were considered satisfactory by comparison with previous studies. (author)

  14. Models for environmental impact assessments of releases of radioactive substances from CERN facilities

    CERN Document Server

    Vojtyla, P

    2005-01-01

    The document describes generic models for environmental impact assessments of releases of radioactive substances from CERN facilities. Except for few models developed in the Safety Commission, the models are based on the 1997 Swiss directive HSK-R-41 and on the 2001 IAEA Safety Report No. 19. The writing style is descriptive, facilitating the practical implementation of the models at CERN. There are four scenarios assumed for airborne releases: (1) short-term releases for release limit calculations, (2) actual short-term releases, (3) short-term releases during incidents/accidents, and (4) chronic long-term releases during the normal operation of a facility. For water releases, two scenarios are considered: (1) a release into a river, and (2) a release into a water treatment plant. The document shall be understood as a reference for specific environmental studies involving radioactive releases and as a recommendation of the Safety Commission.

  15. Bacterial self-defense antibiotics release from organic-inorganic hybrid multilayer films for long-term anti-adhesion and biofilm inhibition properties.

    Science.gov (United States)

    Xu, Qingwen; Li, Xi; Jin, Yingying; Sun, Lin; Ding, Xiaoxu; Liang, Lin; Wang, Lei; Nan, Kaihui; Ji, Jian; Chen, Hao; Wang, Bailiang

    2017-12-14

    Implant-associated bacterial infections pose serious medical and financial issues due to the colonization and proliferation of pathogens on the surface of the implant. The as-prepared traditional antibacterial surfaces can neither resist bacterial adhesion nor inhibit the development of biofilm over the long term. Herein, novel (montmorillonite/poly-l-lysine-gentamicin sulfate) 8 ((MMT/PLL-GS) 8 ) organic-inorganic hybrid multilayer films were developed to combine enzymatic degradation PLL for on-demand self-defense antibiotics release. Small molecule GS was loaded into the multilayer films during self-assembly and the multilayer films showed pH-dependent and linear growth behavior. The chymotrypsin- (CMS) and bacterial infections-responsive film degradation led to the peeling of the films and GS release. Enzyme-responsive GS release exhibited CMS concentration dependence as measured by the size of the inhibition zone and SEM images. Notably, the obtained antibacterial films showed highly efficient bactericidal activity which killed more than 99.9% of S. aureus in 12 h. Even after 3 d of incubation in S. aureus, E. coli or S. epidermidis solutions, the multilayer films exhibited inhibition zones of more than 1.5 mm in size. Both in vitro and in vivo antibacterial tests indicated good cell compatibility, and anti-inflammatory, and long-term bacterial anti-adhesion and biofilm inhibition properties.

  16. Numerical modelling and parametric study of the atmospheric dispersion after radionuclide releases: the Chernobyl accident and the Algeciras incident. Comparison with observation data

    International Nuclear Information System (INIS)

    Minier, Y.; Mathieu, A.; Quelo, D.; Sportisse, B.; Isnard, O.; Krysta, M.; Bocquet, M.

    2006-01-01

    Full text: The attempts of modelling the release following upon the Chernobyl accident and the Algeciras incident are reported. Computing power and observation database are used for sensitivity and parametric studies. The meteorological mesoscale model MM5 is nudged with the ERA-40 reanalysis to simulate the meteorological conditions used by the dispersion model, POLAIR3D. In case of the Chernobyl accident the points of interest are many: the representativity of the meteorological simulations is evaluated using observations with a special focus on precipitation events. The radionuclide dispersion, the dry deposition and scavenging simulated by POLAIR3D are compared with European measurements of activities and depositions. Results of the sensitivity studies are done to evaluate the impact of the deposition parameterizations and source-term characteristics (height of release, quantities). The time dynamic of the contaminated cloud is also investigated with regard to the arrival time on different countries. Similarly, for the Algeciras release, sensitivity to the meteorological fields, source term and depletion processes are analyzed. For the available activity concentrations in the air, data-model comparisons are performed. (author)

  17. Short-term X-ray variability of the globular cluster source 4U 1820 - 30 (NGC 6624)

    Science.gov (United States)

    Stella, L.; Kahn, S. M.; Grindlay, J. E.

    1984-01-01

    Analytical techniques for improved identification of the temporal and spectral variability properties of globular cluster and galactic bulge X-ray sources are described in terms of their application to a large set of observations of the source 4U 1820 - 30 in the globular cluster NGC 6624. The autocorrelation function, cross-correlations, time skewness function, erratic periodicities, and pulse trains are examined. The results are discussed in terms of current models with particular emphasis on recent accretion disk models. It is concluded that the analyzed observations provide the first evidence for shot-noise variability in a globular cluster X-ray source.

  18. Influence of Chemistry on source term assessment

    International Nuclear Information System (INIS)

    Herranz Puebla, L.E.; Lopez Diez, I.; Rodriguez Maroto, J.J.; Martinez Lopez-Alcorocho, A.

    1991-01-01

    The major goal of a phenomenology analysis of containment during a severe accident situation can be splitedd into the following ones: to know the containment response to the different loads and to predict accurately the fission product and aerosol behavior. In this report, the main results coming from the study of a hypothetical accident scenario, based on LA-4 experiment of LACE project, are presented. In order to do it, several codes have been coupled: CONTEMPT4/MOD5 (thermalhydraulics), NAUA/MOD5 (aerosol physics) and IODE (iodine chemistry). 12 refs. It has been demonstrated the impossibility of assessing with confidence the Source Term if the chemical conduct of some radionuclides is not taken into account. In particular, the influence on the iodine retention efficiency of the sump of variables such as pH has been proven. (Author). 12 refs

  19. The ALICE Software Release Validation cluster

    International Nuclear Information System (INIS)

    Berzano, D; Krzewicki, M

    2015-01-01

    One of the most important steps of software lifecycle is Quality Assurance: this process comprehends both automatic tests and manual reviews, and all of them must pass successfully before the software is approved for production. Some tests, such as source code static analysis, are executed on a single dedicated service: in High Energy Physics, a full simulation and reconstruction chain on a distributed computing environment, backed with a sample “golden” dataset, is also necessary for the quality sign off. The ALICE experiment uses dedicated and virtualized computing infrastructures for the Release Validation in order not to taint the production environment (i.e. CVMFS and the Grid) with non-validated software and validation jobs: the ALICE Release Validation cluster is a disposable virtual cluster appliance based on CernVM and the Virtual Analysis Facility, capable of deploying on demand, and with a single command, a dedicated virtual HTCondor cluster with an automatically scalable number of virtual workers on any cloud supporting the standard EC2 interface. Input and output data are externally stored on EOS, and a dedicated CVMFS service is used to provide the software to be validated. We will show how the Release Validation Cluster deployment and disposal are completely transparent for the Release Manager, who simply triggers the validation from the ALICE build system's web interface. CernVM 3, based entirely on CVMFS, permits to boot any snapshot of the operating system in time: we will show how this allows us to certify each ALICE software release for an exact CernVM snapshot, addressing the problem of Long Term Data Preservation by ensuring a consistent environment for software execution and data reprocessing in the future. (paper)

  20. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Trial Calculation. Work Plan

    International Nuclear Information System (INIS)

    Grabaskas, David; Bucknor, Matthew; Jerden, James; Brunett, Acacia J.

    2016-01-01

    The overall objective of the SFR Regulatory Technology Development Plan (RTDP) effort is to identify and address potential impediments to the SFR regulatory licensing process. In FY14, an analysis by Argonne identified the development of an SFR-specific MST methodology as an existing licensing gap with high regulatory importance and a potentially long lead-time to closure. This work was followed by an initial examination of the current state-of-knowledge regarding SFR source term development (ANLART-3), which reported several potential gaps. Among these were the potential inadequacies of current computational tools to properly model and assess the transport and retention of radionuclides during a metal fuel pool-type SFR core damage incident. The objective of the current work is to determine the adequacy of existing computational tools, and the associated knowledge database, for the calculation of an SFR MST. To accomplish this task, a trial MST calculation will be performed using available computational tools to establish their limitations with regard to relevant radionuclide release/retention/transport phenomena. The application of existing modeling tools will provide a definitive test to assess their suitability for an SFR MST calculation, while also identifying potential gaps in the current knowledge base and providing insight into open issues regarding regulatory criteria/requirements. The findings of this analysis will assist in determining future research and development needs.

  1. Evaluation of applicability of alternative source terms to operating nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Lim, S. N.; Park, Y. S.; Nam, K. M.; Song, D. B.; Bae, Y. J.; Lee, Y. J.; Jung, C. Y.

    2002-01-01

    In 1995 and 2000, NRC issued NUREG-1465 and Regulatory Guide 1.183 with respect to Alternative Source Terms(AST) replacing the existing source terms of TID-14844 and Regulatory Guide 1.4, 1.25, and 1.77 for radiological Design Basis Accidents(DBA) analysis. In 1990, ICRP published ICRP Pub. 60 which represents new recommendations on dose criteria and concepts. In Korea, alternative source terms were used for evaluation of effective doses for design basis accidents of Advanced Power Reactor(APR1400) using the computer program developed by an overseas company. Recently, DBADOSE, new computer program for DBA analysis incorporating AST and effective dose concept was developed by KHNP and KOPEC, and reanalysis applying AST to operating nuclear power plants, Kori units 3 and 4 in Korea using DBADOSE has been performed. As the results of this analysis, it was concluded that some conservative variables or operation procedures of operating plants could be mitigated or simplified by virtue of increased safety margin and consequently, economical and operational benefits ensue. In this paper, methodologies and results of Kori 3 and 4 DBA reanalysis and sensitivity analysis for mitigation of main design variables are introduced

  2. Alkali Release from Typical Danish Aggregates to Potential ASR Reactive Concrete

    DEFF Research Database (Denmark)

    Thomsen, Hans Christian Brolin; Grelk, Bent; Barbosa, Ricardo Antonio

    Alkali-silica reaction (ASR) in concrete is a well-known deterioration mechanism affecting the long term durability of Danish concrete structures. Deleterious ASR cracking can be significantly reduced or prevented by limiting the total alkali content of concrete under a certain threshold limit......, which in Denmark is recommended to 3 kg/m3 Na2Oeq.. However, this threshold limit does not account for the possible internal contribution of alkali to the concrete pore solution by release from aggregates or external contributions from varies sources. This study indicates that certain Danish aggregates...... are capable of releasing more than 0.46 kg/m3 Na2Oeq. at 13 weeks of exposure in laboratory test which may increase the risk for deleterious cracking due to an increase in alkali content in the concrete....

  3. Characterization of unsaturated fatty acid sustained-release microspheres for long-term algal inhibition.

    Science.gov (United States)

    Ni, Lixiao; Jie, Xiaoting; Wang, Peifang; Li, Shiyin; Hu, Shuzhen; Li, Yiping; Li, Yong; Acharya, Kumud

    2015-02-01

    The unsaturated fatty acid (linoleic acid) sustained-release microspheres were prepared with linoleic acid (LA) using alginate-chitosan microcapsule technology. These LA sustained-release microspheres had a high encapsulation efficiency (up to 62%) tested by high performance liquid chromatography with a photo diode array. The dry microspheres were characterized by a scanning electron microscope, X-ray diffraction measurement, dynamic thermogravimetric analysis and Fourier transform infrared spectral analysis. The results of characterization showed that the microspheres had good thermal stability (decomposition temperature of 236°C), stable and temperature independent release properties (release time of more than 40 d). Compared to direct dosing of LA, LA sustained-released microspheres could inhibit Microcystis aeruginosa growth to the non-growth state. The results of this study suggested that the LA sustained-release microspheres may be a potential candidate for algal inhibition. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Long term leaching of chlorinated solvents from source zones in low permeability settings with fractures

    DEFF Research Database (Denmark)

    Bjerg, Poul Løgstrup; Chambon, Julie Claire Claudia; Troldborg, Mads

    2008-01-01

    spreads to the low permeability matrix by diffusion. This results in a long term source of contamination due to back-diffusion. Leaching from such sources is further complicated by microbial degradation under anaerobic conditions to sequentially form the daughter products trichloroethylene, cis...

  5. Measurements of instant-release source terms for 137Cs, 90Sr, 99Tc, 129I and 14C in used CANDU fuels

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, S.

    1996-01-01

    Combined gap and grain-boundary inventories of 137 Cs, 129 I, 90 Sr, 99 Tc and 14 C were measured in 15 used CANDU fuel elements by leaching crushed fuel samples. A good correlation between the combined gap and grain-boundary inventories of 137 Cs and 129 I was found, suggesting that these fission products exhibit similar behavior in CANDU fuel. The expected correlation between combined gap and grain-boundary inventories of 137 Cs and 129 I with calculated fission-gas release to the gap and grain boundaries could only be confirmed for lower power fuels ( 90 Sr were higher than expected and showed no correlation with calculated fission-gas release. No values for the combined gap and grain-boundary inventories of 99 Tc were obtained because 99 Tc in used fuel samples is very insoluble and appears to require oxidation prior to dissolution. Combined gap and grain-boundary inventories of 14 C appeared to be independent of fuel power or burnup. (orig.)

  6. Chernobyl radiocesium in freshwater fish: Long-term dynamics and sources of variation

    Energy Technology Data Exchange (ETDEWEB)

    Sundbom, M [Uppsala Univ., Dept. of Limnology, Uppsala (Sweden)

    2002-04-01

    The aim of this thesis was to investigate both the long-term temporal pattern and sources of individual variation for radiocesium in freshwater fish. The basis for the study is time series of {sup 137}Cs activity concentrations in fish from three lakes in the area North-west of Uppsala, Sweden that received considerable amounts of {sup 137}Cs from Chernobyl in may 1986. The lakes were Lake Ekholmssjoen, Lake Flatsjoen and Lake Siggeforasjoen, all small forest lakes, but with different morphometrical and chemical characteristics. The data were collected regularly, usually several times per year, during 1986-2000, using consistent methods. More than 7600 fish individuals from 7 species covering wide size ranges and feeding habits were analysed for {sup 137}Cs. For each fish was the length, weight, sex, and often the stomach contend recorded. The evaluation on long-term trends were based on data from all three lakes, while the study on sources of variation evaluated data from Lake Flatsjoen only. (au)

  7. Chernobyl radiocesium in freshwater fish: Long-term dynamics and sources of variation

    International Nuclear Information System (INIS)

    Sundbom, M.

    2002-01-01

    The aim of this thesis was to investigate both the long-term temporal pattern and sources of individual variation for radiocesium in freshwater fish. The basis for the study is time series of 137 Cs activity concentrations in fish from three lakes in the area North-west of Uppsala, Sweden that received considerable amounts of 137 Cs from Chernobyl in may 1986. The lakes were Lake Ekholmssjoen, Lake Flatsjoen and Lake Siggeforasjoen, all small forest lakes, but with different morphometrical and chemical characteristics. The data were collected regularly, usually several times per year, during 1986-2000, using consistent methods. More than 7600 fish individuals from 7 species covering wide size ranges and feeding habits were analysed for 137 Cs. For each fish was the length, weight, sex, and often the stomach contend recorded. The evaluation on long-term trends were based on data from all three lakes, while the study on sources of variation evaluated data from Lake Flatsjoen only. (au)

  8. Results of dose calculations for NET accidental and normal operation releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1992-08-01

    This report documents conditions, data and results of dose calculations for accidental and normal operation releases of tritium and activation products, performed within the NET subtask SEP2.2 ('NET-Benchmark') of the European Fusion Technology Programme. For accidental releases, the computer codes UFOTRI and COSYMA for assessing the radiological consequences, have been applied for both deterministic and probabilistic calculations. The influence on dose estimates of different release times (2 minutes / 1 hour), two release heights (10 m / 150 m), two chemical forms of tritium (HT/HTO), and two different model approaches for the deposition velocity of HTO on soil was investigated. The dose calculations for normal operation effluents were performed using the tritium model of the German regulatory guidelines, parts of the advanced dose assessment model NORMTRI still under development, and the statistical atmospheric dispersion model ISOLA. Accidental and normal operation source terms were defined as follows: 10g (3.7 10 15 Bq) for accidental tritium releases, 10 Ci/day (3.7 10 11 Bq/day) for tritium releases during normal operation and unit releases of 10 9 Bq for accidental releases of activation products and fission products. (orig./HP) [de

  9. Basic repository source term and data sheet report: Davis Canyon

    International Nuclear Information System (INIS)

    1988-01-01

    This report is one of series describing studies undertaken in support of the US Department of Energy Civilian Radioactive Waste Management (CRWM) Program. This study contains the derivation of values for environmental source terms and resources consumed for a CRWM repository. Estimates include heavy construction equipment; support equipment; shaft-sinking equipment; transportation equipment; and consumption of fuel, water electricity, and natural gas. Data are presented for construction and operation at an assumed site in Davis Canyon, Utah. 6 tabs

  10. Implementation of a new gamma spectrometer on the MERARG loop: Application to the volatile fission products release measurement

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, S.; Gleizes, B.; Pontillon, Y.; Hanus, E.; Ducros, G. [CEA, DEN, DEC, SA3C, F-13108, Saint Paul lez Durance, (France); Roure, C. [CEA, DEN, DTN, SMTA, F-13108, Saint Paul lez Durance, (France)

    2015-07-01

    The MERARG facility initially aims at the annealing of irradiated fuel samples to study the gaseous fission products release kinetics. In order to complete the evaluation of the source term potentially released during accidental situation, the MERARG experimental circuit has been enhanced with a new gamma spectrometer. This one is directly sighting the fuel and is devoted to the fission products release kinetics. Because of the specificities of the fuel measurements, it has been dimensioned and designed to match the specific requirements. The acquisition chain and the collimation system have been optimized for this purpose and a first set of two experiments have shown the good functioning of this new spectrometry facility. (authors)

  11. Algorithms and analytical solutions for rapidly approximating long-term dispersion from line and area sources

    Science.gov (United States)

    Barrett, Steven R. H.; Britter, Rex E.

    Predicting long-term mean pollutant concentrations in the vicinity of airports, roads and other industrial sources are frequently of concern in regulatory and public health contexts. Many emissions are represented geometrically as ground-level line or area sources. Well developed modelling tools such as AERMOD and ADMS are able to model dispersion from finite (i.e. non-point) sources with considerable accuracy, drawing upon an up-to-date understanding of boundary layer behaviour. Due to mathematical difficulties associated with line and area sources, computationally expensive numerical integration schemes have been developed. For example, some models decompose area sources into a large number of line sources orthogonal to the mean wind direction, for which an analytical (Gaussian) solution exists. Models also employ a time-series approach, which involves computing mean pollutant concentrations for every hour over one or more years of meteorological data. This can give rise to computer runtimes of several days for assessment of a site. While this may be acceptable for assessment of a single industrial complex, airport, etc., this level of computational cost precludes national or international policy assessments at the level of detail available with dispersion modelling. In this paper, we extend previous work [S.R.H. Barrett, R.E. Britter, 2008. Development of algorithms and approximations for rapid operational air quality modelling. Atmospheric Environment 42 (2008) 8105-8111] to line and area sources. We introduce approximations which allow for the development of new analytical solutions for long-term mean dispersion from line and area sources, based on hypergeometric functions. We describe how these solutions can be parameterized from a single point source run from an existing advanced dispersion model, thereby accounting for all processes modelled in the more costly algorithms. The parameterization method combined with the analytical solutions for long-term mean

  12. Impaired basal glucose effectiveness but unaltered fasting glucose release and gluconeogenesis during short-term hypercortisolemia in healthy subjects

    DEFF Research Database (Denmark)

    Nielsen, Michael F; Caumo, Andrea; Chandramouli, Visvanathan

    2004-01-01

    Excess cortisol has been demonstrated to impair hepatic and extrahepatic insulin action. To determine whether glucose effectiveness and, in terms of endogenous glucose release (EGR), gluconeogenesis, also are altered by hypercortisolemia, eight healthy subjects were studied after overnight infusion...... resistance. Postabsorptive glucose production (P = 0.64) and the fractional....... Hepatic GE was lower during cortisol than during saline infusion (2.39 +/- 0.24 vs. 3.82 +/- 0.51 ml.kg-1.min-1; P

  13. Development of dose calculation program (DBADOSE) incorporating alternative source term due to design basis accident

    International Nuclear Information System (INIS)

    Bae, Young Jig; Nam, Ki Mun; Lee, Yu Jong; Chung, Chan Young

    2003-01-01

    Source terms presented in TID-14844 and Regulatory Guide 1.4 have been used for radiological analysis of design basis accidents for licensing existing pressurized water reactor (PWR). However, more realistic and physically-based source term based on results of study and experiments for about 30 years after the publication of TID-14844 was developed and presented in NUREG-1465 published by U.S NRC in 1995. In addition, ICRP has revised dose concepts and criteria through the publication of ICRP-9, 26, 60 and recommended effective dose concepts rather than critical organ concept since the publication of ICRP-26. Accordingly, multipurpose computer program called DBADOSE incorporating alternative source terms in NUREG-1465 and effective dose concepts in ICRP-60 was developed. Comparison of results of DBADOSE with those of POSTDBA and STARDOSE was performed and verified and no significant difference and inaccuracy were found. DBADOSE will be used to evaluate accidental doses for licensing application according to the domestic laws that are expected to be revised in the near future

  14. EIA new releases

    International Nuclear Information System (INIS)

    1994-09-01

    This report is a compliation of news releases from the Energy Information Administration. The september-october report includes articles on energy conservation, energy consumption in commercial buildings, and a short term energy model for a personal computer

  15. Low-molecular-weight organoiodine and organobromine compounds released by polar macroalgae--the influence of abiotic factors.

    Science.gov (United States)

    Laturnus, F; Giese, B; Wiencke, C; Adams, F C

    2000-01-01

    The influence of temperature, light, salinity and nutrient availability on the release of volatile halogenated hydrocarbons was investigated in the Antarctic red macroalgal species Gymnogongrus antarcticus Skottsberg. Compared to standard culture condition, an increase in the release rates of iodocompounds was generally found for the exposure of the alga to altered environmental conditions. Macroalgae exhibited higher release rates after adaptation for two months to the changed factors, than after short-term exposure. Monitoring the release rates during a 24 h incubation period (8.25 h light, 15.75 h darkness) showed that changes between light and dark periods had no influence on the release of volatile halocarbons. Compounds like bromoform and 1-iodobutane exhibited constant release rates during the 24 h period. The formation mechanisms and biological role of volatile organohalogens are discussed. Although marine macroalgae are not considered to be the major source of biogenically-produced volatile organohalogens, they contribute significantly to the bromine and iodine cycles in the environment. Under possible environmental changes like global warming and uncontrolled entrophication of the oceans their significance may be increase.

  16. Assessment of Radioactive Liquid Effluents Release at IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    Bessa Nisti, Marcelo; Godoy dos Santos, Adir Janete

    2008-01-01

    A continuous effluent monitoring program has been established at IPEN's plant in order to allow an environmental impact assessment due to radioactive liquid effluent discharge to sanitary system. Representative samples of radioactive liquid effluents are analyzed by using high resolution gamma spectroscopy and instrumental neutron activation analysis, facing to Brazilian radioprotection regulatory rules. The results are consolidating yearly in the Institute source-term. In this paper, results of the source-term are presented, concerning to years 2004, 2005 and 2006. The total activity discharged was 8.5xl0 8 Bq, 5.7x10 8 Bq and 2.7xl0 8 Bq, respectively. As the release is strongly dependent on the total amount of the effluent and on the dilution factor, special attention is needed in order to obtain the correct value of that last one. The estimated inside plant dilution factor, considering the recent facilities and the reshaping of the sewerage system was 80, 180 and 130, for period of 2004, 2005 and 2006 discharged liquid radioactive effluent

  17. The Impact of Pollution Prevention on Toxic Environmental Releases from U.S. Manufacturing Facilities.

    Science.gov (United States)

    Ranson, Matthew; Cox, Brendan; Keenan, Cheryl; Teitelbaum, Daniel

    2015-11-03

    Between 1991 and 2012, the facilities that reported to the U.S. Environmental Protection Agency's Toxic Release Inventory (TRI) Program conducted 370,000 source reduction projects. We use this data set to conduct the first quasi-experimental retrospective evaluation of how implementing a source reduction (pollution prevention) project affects the quantity of toxic chemicals released to the environment by an average industrial facility. We use a differences-in-differences methodology, which measures how implementing a source reduction project affects a facility's releases of targeted chemicals, relative to releases of (a) other untargeted chemicals from the same facility, or (b) the same chemical from other facilities in the same industry. We find that the average source reduction project causes a 9-16% decrease in releases of targeted chemicals in the year of implementation. Source reduction techniques vary in effectiveness: for example, raw material modification causes a large decrease in releases, while inventory control has no detectable effect. Our analysis suggests that in aggregate, the source reduction projects carried out in the U.S. since 1991 have prevented between 5 and 14 billion pounds of toxic releases.

  18. Source Term Characteristics Analysis for Structural Components in PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Choi, Heui Joo; Cho, Dong Keun [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core under different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.32x1015 Bequerels, 238 Watts, 4.32x109 m3 water, respectively, at 10 years after discharge. Those values correspond to 0.6 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 25{approx}50 % and 35{approx}40 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important

  19. Productivity of irrigated beans due to sources of stabilized nitrogen fertilizer and controlled release

    Directory of Open Access Journals (Sweden)

    Tatiely Gomes Bernardes

    2015-12-01

    Full Text Available ABSTRACT New nitrogen fertilizers are available in the market actually, however, does not have results on the efficiency of the Cerrado conditions. With that objective of this study was to evaluate the effect of urea including stabilized and controlled release urea on yield of irrigated common beans (Phaseolus vulgaris L in no-tillage system. The experiment was conducted in the winter crop, at Embrapa Arroz e Feijão, in Santo Antônio de Goiás, State of Goiás, Brazil. The experimental design was randomized blocks, with five replicates. Treatments consisted of five N sources (urea, urea + NBPT, urea + polymer, ammonium sulphate, and ammonium nitrate and a control (without N being applied 20 kg ha-1 of N at sowing and 80 kg ha-1 onf N in topdressing. We evaluated the chlorophyll content in leaves of common beans, the leaf N content and dry mass weight (MSPA in the flowering of common beans, the number of pods per plant, number of grains per pod, mass of 100 grains, grain yield and final stand of the common beans. The sources of nitrogen fertilizer did not influence, leaf N content, the mass of MSPA and the relative chlorophyll index of common beans. The use of polymerized urea and urea with urease inhibitor, did not produce increases in the number of grains per pod, number of pods per plant, mass of 100 grains and common beans yield compared to traditional sources of N, urea, ammonium sulfate and ammonium nitrate.

  20. Analysis of the source term in the Chernobyl-4 accident

    International Nuclear Information System (INIS)

    Alonso, A.; Lopez Montero, J.V.; Pinedo Garrido, P.

    1990-01-01

    The report presents the analysis of the Chernobyl accident and of the phenomena with major influence on the source term, including the chemical effects of materials dumped over the reactor, carried out by the Chair of Nuclear Technology at Madrid University under a contract with the CEC. It also includes the comparison of the ratio (Cs-137/Cs-134) between measurements performed by Soviet authorities and countries belonging to the Community and OECD area. Chapter II contains a summary of both isotope measurements (Cs-134 and Cs-137), and their ratios, in samples of air, water, soil and agricultural and animal products collected by the Soviets in their report presented in Vienna (1986). Chapter III reports on the inventories of cesium isotopes in the core, while Chapter IV analyses the transient, especially the fuel temperature reached, as a way to deduce the mechanisms which took place in the cesium escape. The cesium source term is analyzed in Chapter V. Normal conditions have been considered, as well as the transient and the post-accidental period, including the effects of deposited materials. The conclusion of this study is that Chernobyl accidental sequence is specific of the RBMK type of reactors, and that in the Western world, basic research on fuel behaviour for reactivity transients has already been carried out

  1. Source term calculations - Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Johansson, L.L.

    1998-02-01

    This project was performed within the fifth and final phase of sub-project RAK-2.1 of the Nordic Co-operative Reactor Safety Program, NKS.RAK-2.1 has also included studies of reflooding of degraded core, recriticality and late phase melt progression. Earlier source term calculations for Swedish nuclear power plants are based on the integral code MAAP. A need was recognised to compare these calculations with calculations done with mechanistic codes. In the present work SCDAP/RELAP5 and CONTAIN were used. Only limited results could be obtained within the frame of RAK-2.1, since many problems were encountered using the SCDAP/RELAP5 code. The main obstacle was the extremely long execution times of the MOD3.1 version, but also some dubious fission product calculations. However, some interesting results were obtained for the studied sequence, a total loss of AC power. The report describes the modelling approach for SCDAP/RELAP5 and CONTAIN, and discusses results for the transient including the event of a surge line creep rupture. The study will probably be completed later, providing that an improved SCDAP/RELAP5 code version becomes available. (au) becomes available. (au)

  2. Basic repository source term and data sheet report: Lavender Canyon

    International Nuclear Information System (INIS)

    1988-01-01

    This report is one of a series describing studies undertaken in support of the US Department of Energy Civilian Radioactive Waste Management (CRWM) Program. This study contains the derivation of values for environmental source terms and resources consumed for a CRWM repository. Estimates include heavy construction equipment; support equipment; shaft-sinking equipment; transportation equipment; and consumption of fuel, water, electricity, and natural gas. Data are presented for construction and operation at an assumed site in Lavender Canyon, Utah. 3 refs; 6 tabs

  3. Assessment of Radionuclides Release from Inshas LILW Disposal Facility Under Normal and Unusual Operational Conditions

    International Nuclear Information System (INIS)

    Zaki, A.A.

    2008-01-01

    Disposing of low and intermediate radioactive waste (LILW) is a big concern for Egypt due to the accumulated waste as a result of past fifty years of peaceful nuclear applications. Assessment of radionuclides release from Inshas LILW disposal facility under normal and unusual operational conditions is very important in order to apply for operation license of the facility. Aqueous release of radionuclides from this disposal facility is controlled by water flow, access of the water to the wasteform, release of the radionuclides from the wasteform, and transport to the disposal facility boundary. In this work, the release of 137 Cs , 6C o, and 90 Sr radionuclides from the Inshas disposal facility was studied under the change of operational conditions. The release of these radio contaminants from the source term to the unsaturated and saturated zones , to groundwater were studied. It was found that the concentration of radionuclides in a groundwater well located 150 m away from the Inshas disposal facility is less than the maximum permissible concentration in groundwater in both cases

  4. Real-time modeling of complex atmospheric releases in urban areas

    International Nuclear Information System (INIS)

    Baskett, R.L.; Ellis, J.S.; Sullivan, T.J.

    1994-08-01

    If a nuclear installation in or near an urban area has a venting, fire, or explosion, airborne radioactivity becomes the major concern. Dispersion models are the immediate tool for estimating the dose and contamination. Responses in urban areas depend on knowledge of the amount of the release, representative meteorological data, and the ability of the dispersion model to simulate the complex flows as modified by terrain or local wind conditions. A centralized dispersion modeling system can produce realistic assessments of radiological accidents anywhere in a country within several minutes if it is computer-automated. The system requires source-term, terrain, mapping and dose-factor databases, real-time meteorological data acquisition, three-dimensional atmospheric transport and dispersion models, and experienced staff. Experience with past responses in urban areas by the Atmospheric Release Advisory Capability (ARAC) program at Lawrence Livermore National Laboratory illustrate the challenges for three-dimensional dispersion models

  5. Real-time modelling of complex atmospheric releases in urban areas

    International Nuclear Information System (INIS)

    Baskett, R.L.; Ellis, J.S.; Sullivan, T.J.

    2000-01-01

    If a nuclear installation in or near an urban area has a venting, fire, or explosion, airborne radioactivity becomes the major concern. Dispersion models are the immediate tool for estimating the dose and contamination. Responses in urban areas depend on knowledge of the amount of the release, representative meteorological data, and the ability of the dispersion model to simulate the complex flows as modified by terrain or local wind conditions. A centralised dispersion modelling system can produce realistic assessments of radiological accidents anywhere in a country within several minutes if it is computer-automated. The system requires source-term, terrain, mapping and dose-factor databases, real-time meteorological data acquisition, three-dimensional atmospheric transport and dispersion models, and experienced staff. Experience with past responses in urban areas by the Atmospheric Release Advisory Capability (ARAC) program at Lawrence Livermore National Laboratory illustrate the challenges for three-dimensional dispersion models. (author)

  6. Assessment of atmospherically-released radionuclides using the computerized radiological risk investigation system

    International Nuclear Information System (INIS)

    Nelson, C.B.; Sjoreen, A.L.; Miller, C.W.; Baes, C.F. III.

    1986-01-01

    For radionuclides, the standards are in terms of an annual dose, and the regulations require assurance that no member of the general public receives a dose in excess of that standard. Thus, spatial variations in the population around an emission source must be considered. Furthermore, for most chemical pollutants the standards are written in terms of an air concentration while for radionuclides other pathways of exposure, e.g., uptake of the airborne emissions by terrestrial food chains must also be considered. The remainder of this paper discusses the computer codes that make up the CRRIS and how they are used to perform an assessment of the health impacts on man of radionuclides released to the atmosphere

  7. Larval salivary glands are a source of primer and releaser pheromone in honey bee ( Apis mellifera L.)

    Science.gov (United States)

    Conte, Yves Le; Bécard, Jean-Marc; Costagliola, Guy; de Vaublanc, Gérard; Maâtaoui, Mohamed El; Crauser, Didier; Plettner, Erika; Slessor, Keith N.

    2006-05-01

    A brood pheromone identified in honeybee larvae has primer and releaser pheromone effects on adult bees. Using gas chromatography-mass spectrometry (GC-MS) to evaluate fatty acid esters—the pheromonal compounds—in different parts of the larvae, we have localized the source of the esters as the larval salivary glands. A histochemical study describes the glands and confirms the presence of lipids in the glands. Epithelial cells of the gland likely secrete the fatty acids into the lumen of the gland. These results demonstrate the salivary glands to be a reservoir of esters, components of brood pheromone, in honeybee larvae.

  8. A study on source term assessment and waste disposal requirement of decontamination and decommissioning for the TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Joo Ho; Lee, Kyung JIn; Lee, Jae Min; Choi, Gyu Seup; Shin, Byoung Sun [Kyunghee Univ., Seoul (Korea, Republic of)

    1999-08-15

    The objective and necessity of the project : TRIGA is the first nuclear facility that decide to decommission and decontamination in our nation. As we estimate the expected life of nuclear power generation at 30 or 40 years, the decommissioning business should be conducted around 2010, and the development of regulatory technique supporting it should be developed previously. From a view of decommissioning and decontamination, the research reactor is just small in scale but it include all decommissioning and decontamination conditions. So, the rules by regulatory authority with decommissioning will be a guide for nuclear power plant in the future. The basis of regulatory technique required when decommissioning the research reactor are the radiological safety security and the data for it. The source term is very important condition not only for security of worker but for evaluating how we dispose the waste is appropriate for conducting the middle store and the procedure after it when the final disposal is considered. The content and the scope in this report contain the procedure of conducting the assessment of the source term which is most important in understanding the general concept of the decommissioning procedure of the decommissioning and decontamination of TRIGA research reactor. That is, the sampling and measuring method is presented as how to measure the volume of the radioactivity of the nuclear facilities. And also, the criterion of classifying the waste occurred in other countries and the site release criteria which is the final step of decommissioning and decontamination presented through MARSSIM. Finally, the program to be applicable through comparing the methods of our nation and other countries ones is presented as plan for disposal of the waste in the decommissioning.

  9. Standardization of iridium-192 coiled source in terms of air kerma output

    Energy Technology Data Exchange (ETDEWEB)

    Shanta, A; Unnikrishnan, K; Tripathi, U B; Kannan, A; Iyer, P S [Bhabha Atomic Research Centre, Bombay (India)

    1996-08-01

    ICRU (1985) recommended that the output of gamma ray brachytherapy sources should be specified in terms of reference air kerma rate, defined as the kerma rate to air in air at a reference distance of 1 meter, perpendicular to the long axis of the source, corrected for air attenuation and scattering. As these measurements are difficult to carry out in the routine clinical use, it is the common practice to calibrate the re-entrant ionization chamber with respect to open air measurements and use the re-entrant chamber for routine measurements. This paper reports on the measurements carried out to correlate the nominal activity and air kerma rate of {sup 192}Ir wire sources supplied by the Board of Radiation and Isotope Technology, Department of Atomic Energy. (author). 3 refs, 1 tab.

  10. Standardization of iridium-192 coiled source in terms of air kerma output

    International Nuclear Information System (INIS)

    Shanta, A.; Unnikrishnan, K.; Tripathi, U.B.; Kannan, A.; Iyer, P.S.

    1996-01-01

    ICRU (1985) recommended that the output of gamma ray brachytherapy sources should be specified in terms of reference air kerma rate, defined as the kerma rate to air in air at a reference distance of 1 meter, perpendicular to the long axis of the source, corrected for air attenuation and scattering. As these measurements are difficult to carry out in the routine clinical use, it is the common practice to calibrate the re-entrant ionization chamber with respect to open air measurements and use the re-entrant chamber for routine measurements. This paper reports on the measurements carried out to correlate the nominal activity and air kerma rate of 192 Ir wire sources supplied by the Board of Radiation and Isotope Technology, Department of Atomic Energy. (author). 3 refs, 1 tab

  11. Long-term program up to fiscal 1993 of electric power source development

    International Nuclear Information System (INIS)

    Kawakami, Shin-ichi

    1984-01-01

    The long-term, ten years, program up to fiscal 1993 of electric power source development, determined by the Government aims at stable power supply and the expansion of utilization of petroleum-substitute energy. The annual growth in the gross national product (GNP) during the ten years was taken as about 4 %. So, the total electric power demand in fiscal 1993 is scheduled to be 731,000 million kwh, about 34 % up from 547,000 million kwh in fiscal 1983. The structure of electric power sources at the end of fiscal 1993 will be hydraulic 19.7 %, thermal 58.3 %, and nuclear 21.9 %. The development of electric power sources to be initiated in fiscal 1984 is hydraulic 500 MW, thermal 2,000 MW, and nuclear 6,000 MW. (Mori, K.)

  12. Modeling of cumulative release on long term leaching behaviour of selected oil sludge from crude oil terminal and petroleum refining plant

    International Nuclear Information System (INIS)

    Mohd Fadzil, S.; Khoo, K.S.; Sarmani, S.; Majid, A.Ab.; Hamzah, A.

    2013-01-01

    Management of oil sludge containing environmentally toxic elements is a major problem in crude oil processing industry. Oil sludge samples from the petroleum refinery plant in Melaka and crude oil terminal in Sarawak were analysed. The aim of present work is to study long term leaching behaviour of arsenic (As), cobalt (Co), chromium (Cr) and zinc (Zn) from oil sludge. Tank leaching test was carried out and the samples were analysed using inductively coupled plasma-mass spectrometry (ICP-MS). The results were studied using LeachXS software to plot the graphs of elements concentration in order to study the leaching behaviour of toxic elements in oil sludge. The long term leaching (100 years) modeling was calculated using equations referred to National Institute of Public Health and the Environment Bilthoven (RIVM) and the results were plotted for cumulative release in different areas of oil sludge. Tank leaching test of the oil sludge samples from petroleum refinery plant in Melaka showed concentrations of As, Co, Cr and Zn ranging from 0.205 to 1.102, 0.031-0.454, 0.016-0.086 and 0.409-8.238 mg/l, respectively while the concentrations of As, Co, Cr and Zn in oil sludge samples from crude oil terminal in Sarawak were in the range of 0.002-0.089, 0.001-0.033, 0.006-1.016 and 0.100-2.744 mg/l, respectively. On the other hand, results on cumulative release from the modeling of long term leaching (100 years) showed that As, Co, Cr and Zn concentrations were proportional to the quantity of oil sludge. In conclusion, during extrapolation of release of toxic elements using the data in the laboratory, several other factors were taken into account to suit environmental conditions such as soil moisture, the negative logarithm of the effective diffusion coefficient (pD e ) and temperature, while the long-term behaviour of As, Co, Cr and Zn was proportional to the quantity of oil sludge to be disposed off. (author)

  13. Updated on effluents releases of the CEA nuclear fuel cycle facilities - 1995 to 2010 period

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Nelson Luiz Dias [Centro Tecnologico da Marinha em Sao Paulo (CTMSP) Sao Paulo, SP (Brazil)

    2011-07-01

    The environmental impact assessment of the Centro Experimental Aramar (CEA) facilities has been presented in a former work, based on the measured effluent releases data, for the period from 1995 to 2007. This work shows the update up to 2010. The effluents releases to the environment result from the routine operation of CEA nuclear fuel cycle facilities (LEI - Isotopic Enrichment Laboratory, USIDE - Pilot Plant for Industrial Verification of Uranium Enrichment and LABMAT - Nuclear Materials Laboratory). Basically, this work presents the radioactive release source terms, as described at the CEA Effluent Report sent to the National Commission for Nuclear Energy (CNEN) each semester, and a historical assessment of the critical group annual doses from 1995 up to 2010. The assessed doses are compared to the maximum dose constraint as well as to the exemption level specified by CNEN. (author)

  14. Updated on effluents releases of the CEA nuclear fuel cycle facilities - 1995 to 2010 period

    International Nuclear Information System (INIS)

    Ferreira, Nelson Luiz Dias

    2011-01-01

    The environmental impact assessment of the Centro Experimental Aramar (CEA) facilities has been presented in a former work, based on the measured effluent releases data, for the period from 1995 to 2007. This work shows the update up to 2010. The effluents releases to the environment result from the routine operation of CEA nuclear fuel cycle facilities (LEI - Isotopic Enrichment Laboratory, USIDE - Pilot Plant for Industrial Verification of Uranium Enrichment and LABMAT - Nuclear Materials Laboratory). Basically, this work presents the radioactive release source terms, as described at the CEA Effluent Report sent to the National Commission for Nuclear Energy (CNEN) each semester, and a historical assessment of the critical group annual doses from 1995 up to 2010. The assessed doses are compared to the maximum dose constraint as well as to the exemption level specified by CNEN. (author)

  15. Radioactivity source terms for underground engineering application

    Energy Technology Data Exchange (ETDEWEB)

    Tewes, H A [Lawrence Radiation Laboratory, Livermore, CA (United States)

    1969-07-01

    The constraints on nuclide production are usually very similar in any underground engineering application of nuclear explosives. However, in some applications the end product could be contaminated unless the proper nuclear device is used. This fact can be illustrated from two underground engineering experiments-Gasbuggy and Sloop. In the Gasbuggy experiment, appreciable tritium has been shown to be present in the gas currently being produced. However, in future gas stimulation applications (as distinct from experiments), a minimum production of tritium by the explosive is desirable since product contamination by this nuclide may place severe limitations on the use of the tritiated gas. In Sloop, where production of copper is the goal of the experiment, product contamination would not be caused by tritium but could result from other nuclides: Thus, gas stimulation could require the use of fission explosives while the lower cost per kiloton of thermonuclear explosives could make them attractive for ore-crushing applications. Because of this consideration, radionuclide production calculations must be made for both fission and for thermonuclear explosives in the underground environment. Such activation calculations materials of construction are performed in a manner similar to that described in another paper, but radionuclide production in the environment must be computed using both fission neutron and 14-MeV neutron sources in order to treat the 'source term' problem realistically. In making such computations, parameter studies including the effects of environmental temperature, neutron shielding, and rock types have been carried out. Results indicate the importance of carefully evaluating the radionuclide production for each individual underground engineering application. (author)

  16. Radioactivity source terms for underground engineering application

    International Nuclear Information System (INIS)

    Tewes, H.A.

    1969-01-01

    The constraints on nuclide production are usually very similar in any underground engineering application of nuclear explosives. However, in some applications the end product could be contaminated unless the proper nuclear device is used. This fact can be illustrated from two underground engineering experiments-Gasbuggy and Sloop. In the Gasbuggy experiment, appreciable tritium has been shown to be present in the gas currently being produced. However, in future gas stimulation applications (as distinct from experiments), a minimum production of tritium by the explosive is desirable since product contamination by this nuclide may place severe limitations on the use of the tritiated gas. In Sloop, where production of copper is the goal of the experiment, product contamination would not be caused by tritium but could result from other nuclides: Thus, gas stimulation could require the use of fission explosives while the lower cost per kiloton of thermonuclear explosives could make them attractive for ore-crushing applications. Because of this consideration, radionuclide production calculations must be made for both fission and for thermonuclear explosives in the underground environment. Such activation calculations materials of construction are performed in a manner similar to that described in another paper, but radionuclide production in the environment must be computed using both fission neutron and 14-MeV neutron sources in order to treat the 'source term' problem realistically. In making such computations, parameter studies including the effects of environmental temperature, neutron shielding, and rock types have been carried out. Results indicate the importance of carefully evaluating the radionuclide production for each individual underground engineering application. (author)

  17. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    International Nuclear Information System (INIS)

    Ewing, Rodney C.

    2004-01-01

    Spent nuclear fuel, essentially U 2 , accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO 2 in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term

  18. Finite volume schemes with equilibrium type discretization of source terms for scalar conservation laws

    International Nuclear Information System (INIS)

    Botchorishvili, Ramaz; Pironneau, Olivier

    2003-01-01

    We develop here a new class of finite volume schemes on unstructured meshes for scalar conservation laws with stiff source terms. The schemes are of equilibrium type, hence with uniform bounds on approximate solutions, valid in cell entropy inequalities and exact for some equilibrium states. Convergence is investigated in the framework of kinetic schemes. Numerical tests show high computational efficiency and a significant advantage over standard cell centered discretization of source terms. Equilibrium type schemes produce accurate results even on test problems for which the standard approach fails. For some numerical tests they exhibit exponential type convergence rate. In two of our numerical tests an equilibrium type scheme with 441 nodes on a triangular mesh is more accurate than a standard scheme with 5000 2 grid points

  19. Short term memory may be the depletion of the readily releasable pool of presynaptic neurotransmitter vesicles of a metastable long term memory trace pattern.

    Science.gov (United States)

    Tarnow, Eugen

    2009-09-01

    The Tagging/Retagging model of short term memory was introduced earlier (Tarnow in Cogn Neurodyn 2(4):347-353, 2008) to explain the linear relationship between response time and correct response probability for word recall and recognition: At the initial stimulus presentation the words displayed tag the corresponding long term memory locations. The tagging process is linear in time and takes about one second to reach a tagging level of 100%. After stimulus presentation the tagging level decays logarithmically with time to 50% after 14 s and to 20% after 220 s. If a probe word is reintroduced the tagging level has to return to 100% for the word to be properly identified, which leads to a delay in response time. This delay is proportional to the tagging loss. The tagging level is directly related to the probability of correct word recall and recognition. Evidence presented suggests that the tagging level is the level of depletion of the Readily Releasable Pool (RRP) of neurotransmitter vesicles at presynaptic terminals. The evidence includes the initial linear relationship between tagging level and time as well as the subsequent logarithmic decay of the tagging level. The activation of a short term memory may thus be the depletion of RRP (exocytosis) and short term memory decay may be the ensuing recycling of the neurotransmitter vesicles (endocytosis). The pattern of depleted presynaptic terminals corresponds to the long term memory trace.

  20. Predicting hydrocarbon release from soil

    International Nuclear Information System (INIS)

    Poppendieck, D.; Loehr, R.C.

    2002-01-01

    'Full text:' The remediation of hazardous chemicals from soils can be a lengthy and costly process. As a result, recent regulatory initiatives have focused on risk-based corrective action (RBCA) approaches. Such approaches attempt to identify the amount of chemical that can be left at a site with contaminated soil and still be protective of human health and the environment. For hydrocarbons in soils to pose risk to human heath and the environment, the hydrocarbons must be released from the soil and accessible to microorganisms, earthworms, or other higher level organisms. The sorption of hydrocarbons to soil can reduce the availability of the hydrocarbon to receptors. Typically in soils and sediments, there is an initial fast release of a hydrocarbon from the soil to the aqueous phase followed by a slower release of the remaining hydrocarbon to the aqueous phase. The rate and extent of slow release can influence aqueous hydrocarbon concentrations and the fate and transport of hydrocarbons in the subsurface. Once the fast fraction of the chemical has been removed from the soil, the remaining fraction of a chemical may desorb at a rate that natural mechanisms can attenuate the released hydrocarbon. Hence, active remediation may be needed only until the fast fraction has been removed. However, the fast fraction is a soil and chemical specific parameter. This presentation will present a tier I type protocol that has been developed to quickly estimate the fraction of hydrocarbons that are readily released from the soil matrix to the aqueous phase. Previous research in our laboratory and elsewhere has used long-term desorption (four months) studies to determine the readily released fraction. This research shows that a single short-term (less than two weeks) batch extraction procedure provides a good estimate of the fast released fraction derived from long-term experiments. This procedure can be used as a tool to rapidly evaluate the release and bioavailability of