WorldWideScience

Sample records for accident source term

  1. 10 CFR 50.67 - Accident source term.

    Science.gov (United States)

    2010-01-01

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

  2. 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.

  3. 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.

  4. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  9. Review of Past Nuclear Accidents: Source Terms and Recorded Gamma-Ray Spectra

    OpenAIRE

    Sanderson, D.C.W.; Cresswell, A.; Allyson, J.D.; McConville, P.

    1997-01-01

    Airborne gamma ray spectrometry using high volume scintillation detectors, optionally in conjunction with Ge detectors, has potential for making rapid environmental measurements in response to nuclear accidents. A literature search on past nuclear accidents has been conducted to define the source terms which have been experienced so far. Selected gamma ray spectra recorded after past accidents have also been collated to examine the complexity of observed behaviour.

  10. Project on Transfer Mechanism of Radioactive Source Term Under Severe Accident

    Institute of Scientific and Technical Information of China (English)

    SUN; Xue-ting; JI; Song-tao; CHEN; Lin-lin

    2012-01-01

    <正>The "Transfer mechanism of radioactive source term under severe accident" is a sub-project of the research program of "Mechanism and phenomenology of severe accident". An aerosol transfer mechanism experimental facility is built to simulate the passive containment cooling system (PCCS) of advanced pressurizer reactors to research effects to the transfer process of fission products under severe accident. An advanced CFD method is also utilized to research the effects. The objective of this project is to understand

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee Min [Department of Engineering and System Science, and Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec II, Kung Fu Road, Hsinchu, Taiwan (China)], E-mail: mlee@mail.ess.nthu.edu.tw; Ko, Y.-C. [Department of Engineering and System Science, and Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec II, Kung Fu Road, Hsinchu, Taiwan, ROC (China)

    2008-04-15

    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.

  12. Advanced sodium fast reactor accident source terms : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    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 eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions 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.

  13. 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.

  14. CFD Analysis of Migration Mechanism of Source Term Under Severe Accident

    Institute of Scientific and Technical Information of China (English)

    CHEN; Lin-lin; SUN; Xue-ting; JI; Song-tao

    2013-01-01

    The analysis of the migration of source term under severe accident is one of the important aspects of‘Studies on Migration Mechanism of the Source Term under Severe Accident’,which is a significant task of the National Large Advanced PWR Research Program.This research aims at building up a method for analyzing fission product behavior in the containment with CFD code.The effect of PCCS(Passive

  15. Development of the source term PIRT based on findings during Fukushima Daiichi NPPs accident

    Energy Technology Data Exchange (ETDEWEB)

    Suehiro, Shoichi, E-mail: suehiro-shouichi@tepsys.co.jp [TEPCO SYSTEMS Co., 2-37-28 Eitai, Koto-Ku, Tokyo 135-0034 (Japan); Sugimoto, Jun [Kyoto University, Yoshida Sakyo, Kyoto 606-8501 (Japan); Hidaka, Akihide [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Okada, Hidetoshi [The Institute of Applied Energy, 14-2 Nishi-Shimbashi 1-Chome, Minato-ku, Tokyo 105-0003 (Japan); Mizokami, Shinya [Tokyo Electric Power Company, 1-3 Uchisaiwai-cho 1-Chome, Chiyoda-ku, Tokyo 100-8560 (Japan); Okamoto, Koji [The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan)

    2015-05-15

    Highlights: • We developed the source term PIRT based on findings during the Fukushima accident. • The FoM is the masses or fractions of radionuclides released into the environment. • 68 phenomena were identified as influencing to the FoM. • Radionuclide release from molten fuel had the highest score in the early phase. • MCCI, iodine chemistry, and chemical form had the highest score in the later phase. - Abstract: Research Expert Committee on Evaluation of Severe Accident of AESJ (Atomic Energy Society of Japan) has developed thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and source term (ST) PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aim 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 analytical codes. The ST PIRT is divided into 3 phases for time domain and 9 categories for spatial domain. The 68 phenomena have been extracted and the importance from the viewpoint of the source term has been ranked through brainstorming and discussions among experts. The present paper describes the developed ST PIRT list and summarizes the high ranked phenomena in each phase.

  16. Revaporisation of fission product deposits in the primary circuit and its impact on accident source term

    OpenAIRE

    BOTTOMLEY Paul; KNEBEL KEVIN; VAN WINCKEL Stefaan; HASTE Tim; Souvi, Sidi,; AUVINEN Ari; KALILAINEN J.; KÄRKELÄ Teemu

    2014-01-01

    Chemical revaporisation or physical resuspension of fission product deposits from the primary circuit is now recognised to be a major source term in the late phase of severe fuel degradation in a severe nuclear accident. These results come from tests carried out under different experimental projects in the European Commission (EC) Framework Programmes. These include the revaporisation tests carried out at the Transuranium Institute (ITU), Karlsruhe under the Fourth Framework Programme, the Ph...

  17. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  18. Inversion method of source term in nuclear accident based on Gaussian puff model

    International Nuclear Information System (INIS)

    The inverse problem of source terms information estimation in nuclear accident is important for emergency response. In this study a review of data assimilation applied on atmospheric dispersion is given. For the atmospheric dispersion model is nonlinear and with model errors, ensemble Kalman filter is adopted for data assimilation. The dispersion consequences is described by Gaussian puff model, and the source term emission rate and release height is estimated real-time. To determine the best first guess parameters' value and errors, more than 10 twin experiments have been carried on. The results show that the ensemble Kalman filter can be applied successfully to estimate the source term information when there are one or two unknown parameters, the estimated accuracy is related to first guess value, and is impacted by the standard deviation of perturbation. To reduce the estimation error, first guess value setting to the half to two times of true value is recommended. (author)

  19. Severe accident source terms for a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • This study analyzes offsite doses for characteristic SFR scenarios. • Models to calculate the source term for an SFR were developed for this work. • Environmental releases are small due to effectiveness of retention mechanisms. • NRC’s Quantitative Health Objectives are satisfied with high margins. - Abstract: In order to support the demonstration of a risk-informed approach to the design optimization of a sodium-cooled fast reactor (SFR), it was necessary to make realistic estimates of the consequences of severe accident scenarios. This paper describes the database, models, and assumptions used to estimate the offsite consequences of characteristic severe accident scenarios. As required for comparison with the NRC’s technology neutral framework limit curve, the offsite dose at one mile from the plant boundary is calculated using conservative meteorology. The reference plant design is a 1000 MWt pool-type design with metallic fuel. Because an integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis, it was necessary to write a computer code that would assess release of radionuclides from the fuel and transport within the reactor primary system and to link those analyses with results from existing computer codes that assess the dynamic response of the reactor, containment thermal–hydraulics, and radionuclide transport processes within the containment. The analyses indicate that the offsite source terms for SFR severe accident scenarios tend to be small because of the low melting temperature of the fuel, likelihood of significant retention of fission products within the sodium pool, augmentation of containment deposition processes by interaction with sodium oxide aerosols, and small driving force for release from the containment to the environment. A number of major sources of modeling uncertainty are identified as requiring further development effort. An integrated modeling capability, similar to the

  20. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  1. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    International Nuclear Information System (INIS)

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences

  2. 77 FR 19740 - Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident

    Science.gov (United States)

    2012-04-02

    ... COMMISSION Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident AGENCY... that provide water sources for emergency core cooling, containment heat removal, or containment... Guide (RG) 1.82, ``Water Sources for Long- Term Recirculation Cooling Following a...

  3. 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.

  4. Methods to prevent the source term of methyl lodide during a core melt accident

    Energy Technology Data Exchange (ETDEWEB)

    Karhu, A. [VTT Energy (Finland)

    1999-11-01

    The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)

  5. Simulation of THAI HD-12 test with the Accident Source Term Evaluation Code (ASTEC)

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, Thimo; Koch, Marco K. [Bochum Univ. (DE). Chair of Energy Systems and Energy Economics (LEE)

    2010-05-15

    test facilities in which vertical up- and downward directed combustion was investigated [2]. The HD-12 experiment is one of the open post test calculations in the ''International Standard Problem on Hydrogen Combustion (ISP-49)''. Within the ISP-49 experiments are simulated by 25 participants of 10 countries (Czech Republic, France, Germany, Hungary, Netherlands, Poland, Finland, Korea, Slovenia and UK) representing 16 organizations. The simulations are subdivided into an open and a blind phase. The objective of the ISP-49 is to check the validity of different computer codes for simulation of hydrogen combustion with a wide range of possible conditions representing different regimes. In the following the simulation of the HD-12 experiment with the Accident Source Term Evaluation Code (ASTEC) is described. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ang, M.L. [NNC Ltd., Knutsford (United Kingdom); Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Kersting, E.J.; Friederichs, H.G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany); Lee, R.Y. [Nuclear Regulatory Commission, Washington, DC (United States); Meyer-Heine, A. [CEA Centre d`Etudes de Cadarache, Saint Paul Lez Durance (France); Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States); Soda, K. [Japan Atomic Energy Research Inst., Tokyo (Japan); Sweet, D. [AEA Technology, Winfrith (United Kingdom)

    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.

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

    International Nuclear Information System (INIS)

    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

  8. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    International Nuclear Information System (INIS)

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m2, fluence 3 MW-y/m2, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass

  9. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    Energy Technology Data Exchange (ETDEWEB)

    Cambi, G. [Bologna Univ. (Italy). Physics Dept.; Cepraga, D.G. [ENEA, Bologna (Italy). Innovation Dept.; Di Pace, L. [ENEA, Frascati (Italy). Fusion Sector CR di Frascati; Porfiri, M.T. [ENEA, Frascati (Italy). Fusion Dept.

    1995-12-31

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m{sup 2}, fluence 3 MW-y/m{sup 2}, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass.

  10. Water simulation experiments on the instantaneous source term of a severe breeder reactor accident

    International Nuclear Information System (INIS)

    FAUST is an experimental program to give contributions to the assessment of the instantaneous source term in case of an LMFBR loss-of-flow accident with expanding fuel or sodium vapor. In the FAUST 1a-series, experiments with discharge of a gas-particle mixture (nitrogen from 0.3 to 2.0 MPa with iron or nickel powder of different particle size) from a 1.45 liter source into a water pool cylinder of 28.8 cm diameter and 1 m height by rupture disks were performed at different pool height (0.90 cm). The system was closed, i.e. no openings were provided in the cover plate. Important measuring instruments were high-speed cameras, pressure transducers and magnets for article trapping in the cover gas. The most important quantity to be determined was the retention factor RF, defined as the ratio of the amount of particles discharged to the amount trapped in the cover gas. Furthermore, the expansion characteristics of the bubble, the correlated cover gas phenomena, the oscillation period and the entrainment were considered. In most cases, particle release stayed below detection limit, which corresponds to RF > 104. For the 1B series, using the same source, a larger pool vessel (63 cm diameter, 60 cm height) was installed and a cover plate with two openings of 4 cm diameter to simulate leaks. The discharge pressure was varied from 0.002 to 4 MPa. Other experimental parameters were pool height (0.50 cm), particles size (1 to 100 μm), and leak size. A release of airborne particles was found only at very low discharge pressure. At high pressure, major amounts of water were released, whereas the release of particles remained below detection limit (retention factor > 104). The oscillation period was of the order of 80 msec for 1A and 50 msec for 1B. Approximative calculations have shown that the large particle absorption may be explained by impaction during the bubble oscillations. (orig.)

  11. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community

  12. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses; Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Hora, S.C. [Hawaii Univ., Hilo, HI (United States)

    1993-12-01

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community.

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

    International Nuclear Information System (INIS)

    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

  14. 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

  15. 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.

  16. 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

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

    International Nuclear Information System (INIS)

    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

  18. Simplified approach for reconstructing the atmospheric source term for Fukushima Daiichi nuclear power plant accident using scanty meteorological data

    International Nuclear Information System (INIS)

    Highlights: • Estimation of source terms for I-131 and Cs-137 for Fukushima Daiichi NPP. • Simplified Gaussian puff based atmospheric dispersion model is used. • Good agreement of estimated values as compared to that given by NISA, TEPCO and IRSN. - Abstract: The atmospheric source term for the Fukushima Daiichi nuclear power plant accident in March 2011 has been estimated by a Gaussian puff based atmospheric dispersion model. The scanty meteorological data available at irregular time intervals are utilized to demonstrate the utility of such data along with a simplified modeling approach to derive useful information. The source terms for I-131 and Cs-137 have been estimated as a function of time from the observed values of activity concentration in the air and deposited activity on the ground. The model results suggest that during 12th March 2011–16th March 2011, 9.29 × 1016 Bq of I-131 and 6.15 × 1015 Bq of Cs-137 might have got released to the environment

  19. Source term and radiological consequence evaluation for nuclear accidents using a 'hand type' methodology

    International Nuclear Information System (INIS)

    In the last decades, hand type calculations have been replaced by computerized solutions, which are much more accurate, but the preparation of an input to run the code can be a time consuming process and can require a laborious work. This is why, a place for hand calculation based on nomograms still exist in some areas. An example is emergency response to an accidental release of radioactive contaminants when the health of persons close to the accident site might be at risk. In this case, results from computerized accident consequences assessment models may be delayed due to the equipment malfunction or the time required developing minimal input files and performing the calculations (typically more than five minutes). A simple nomogram (developed using computerized dispersion model calculations) can provide dispersion and dose estimates within a minute. The paper presents the methodology used for these 'hand type' calculation and the nomograms, figures and tables used to evaluate the dose to an individual close to the release point. In order to illustrate the use of methodology, a hypothetical severe accident scenario involving 14-MW INR-TRIGA research reactor was considered. (authors)

  20. Source term assessment, containment atmosphere control systems, and accident consequences. Report to CSNI by an OECD/NEA Group of experts

    International Nuclear Information System (INIS)

    CSNI Report 135 summarizes the results of the work performed by CSNI's Principal Working Group No. 4 on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 to 1986. This document contains the latest information on some important topics relating to source terms, accident consequence assessment, and containment atmospheric control systems. It consists of five parts: (1) a Foreword and Executive Summary prepared by PWG4's Chairman; (2) a Report on the Technical Status of the Source Term; (3) a Report on the Technical Status of Filtration and Containment Atmosphere Control Systems for Nuclear Reactors in the Event of a Severe Accident; (4) a Report on the Technical Status of Reactor Accident Consequence Assessment; (5) a list of members of PWG4

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

    International Nuclear Information System (INIS)

    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

  2. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kress, T. S. [comp.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  3. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    International Nuclear Information System (INIS)

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  4. 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.

  5. An inverse modeling method to assess the source term of the Fukushima nuclear power plant accident using gamma dose rate observations

    Directory of Open Access Journals (Sweden)

    O. Saunier

    2013-06-01

    Full Text Available The Chernobyl nuclear accident and more recently the Fukushima accident highlighted that the largest source of error on consequences assessment is the source term including the time evolution of the release rate and its distribution between radioisotopes. Inverse modeling methods, which combine environmental measurements and atmospheric dispersion models, have proven efficient in assessing source term due to an accidental situation (Gudiksen, 1989; Krysta and Bocquet, 2007; Stohl et al., 2012a; Winiarek et al., 2012. Most existing approaches are designed to use air sampling measurements (Winiarek et al., 2012 and some of them also use deposition measurements (Stohl et al., 2012a; Winiarek et al., 2013 but none of them uses dose rate measurements. However, it is the most widespread measurement system, and in the event of a nuclear accident, these data constitute the main source of measurements of the plume and radioactive fallout during releases. This paper proposes a method to use dose rate measurements as part of an inverse modeling approach to assess source terms. The method is proven efficient and reliable when applied to the accident at the Fukushima Daiichi nuclear power plant (FD-NPP. The emissions for the eight main isotopes 133Xe, 134Cs, 136Cs, 137Cs, 137mBa, 131I, 132I and 132Te have been assessed. Accordingly, 103 PBq of 131I, 35.5 PBq of 132I, 15.5 PBq of 137Cs and 12 100 PBq of noble gases were released. The events at FD-NPP (such as venting, explosions, etc. known to have caused atmospheric releases are well identified in the retrieved source term. The estimated source term is validated by comparing simulations of atmospheric dispersion and deposition with environmental observations. The result is that the model-measurement agreement for all of the monitoring locations is correct for 80% of simulated dose rates that are within a factor of 2 of the observed values. Changes in dose rates over time have been overall properly

  6. Passive ALWR source term

    International Nuclear Information System (INIS)

    The purpose of this report is to provide technical support for the physically-based source term which is proposed as the licensing design basis fission product release from a major core accident for the Passive Advanced Light Water Reactor (ALWR) in Volume 3, Section 5 of the ALWR Requirements Document. A substantial body of new research motivated by the Three Mile Island (TMI) accident is maturing, and the ALWR Requirements Document provides an opportunity to incorporate this experience in an updated source term. This update will provide a more rational basis for Passive ALWR accident mitigation system designs, particularly where the designs afford opportunities for improvement and innovation. Great attention has been paid to accident prevention in the ALWR Requirements Document which will reduce the likelihood of core damage by an order of magnitude or more compared to earlier LWR designs. Nonetheless, for defense-in-depth the Passive ALWR source term is based on evaluation of a core damage event. Selection of this core damage event and the associated quantification of the fission product release were done in a conservative, yet physically-based manner so as to provide significant margin to the expected releases, given an ALWR accident, while avoiding non-physical assumptions which could produce mitigation system designs not well-suited to the important accidents. The physically-based source term presented in this report is intended for use in ALWR design basis analysis defining the radiological environment for plant systems and equipment and evaluating the offsite dose for emergency planning considerations. 100 refs., 18 figs., 44 tabs

  7. Review of plutonium aerosol source-term in nuclear accident%核事故条件下钚气溶胶源项研究综述

    Institute of Scientific and Technical Information of China (English)

    刘文杰; 胡八一; 李庆忠

    2011-01-01

    The theoretical and experimental evaluations of plutonium aerosol source-term in nuclear accident are summarized and reviewed in this paper. Hie content of this paper include oxidation mechanism of plutonium, paniculate distribution of plutonium aerosol, aerosol release fraction (ARF) and respirable fraction (RF) of radioactive aerosolization during nuclear accident. The source-term data of three kinds of nuclear accidents which are explosive detonation, static combustion and dynamic combustion have been investigated. The latter two accidental scenes tend to imitate stockpile fire accidents and air transportation disasters wherein the aircrafts and missiles with nuclear devices run up against unexpected fire, crash or in-flight breakup respectively . It is indicated that the aerosolization mechanisms of static combustion and the dynamic combustion without plutonium droplets sparking and explosion are all derived from the oxide particles spilled from the plutonium surface. The size distributions of smaller aerosol particles dispersed with updraft and biggish ones deposited in the soil during static combustion have been measured individually. After that, the full-scale distribution could be obtained in accordance with the combination of the two parts. The investigation of full-scale particu-late distribution in the dynamic combustion scene is unreachable due to the restriction of experimental conditions. The aerosolization distribution in the explosive detonation scene is listed via field test data of Operation Roller Coaster. The source-term of static combustion is lower than the dynamic combustion without sparking and explosion of plutonium droplets, which is based on the thermo dissipation and the loss of oxygen around the plutonium solidity. The dynamic combustion with plutonium droplets sparking and explosion leads to vapor venting and higher source-term. During the explosive detonation the reaction of plutonium oxidation is severest and leads to the highest

  8. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    International Nuclear Information System (INIS)

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

  9. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

  10. 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.

  11. 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)

    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

  12. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    International Nuclear Information System (INIS)

    JNES is developing severe accident analysis codes in order to apply to the probability safety analysis (PSA) for a typical fast breeder reactor (FBR). AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary, and the discharge rate to the environment of fission products (FP). This report summarizes analysis results using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass scenario (CVBP) and the containment failure scenario by hydrogen deflagration or detonation. The coolant temperature of the primary system and the secondary system in the PLOHS sequence increases at the almost same temperature, and the creep damage to the reactor coolant boundary will become remarkable if coolant temperature exceeds about 1,100 K. In the CVBP scenario, when an intermediate heat exchanger is ruptured by creep and the boundary of the secondary system is failed, the path from the primary system to environment is formed. Then, the reactor vessel (RV) is failed and sodium in the primary coolant system releases into the reactor vessel room (RV room). Sodium of high temperature which fell in the RV room damages the floor liner, and generates hydrogen by a reaction with concrete. In addition the reactor core is exposed into atmosphere and the core temperature increases with decay heat and then volatile FP and non-volatile FP are released to the environment through the secondary system from the primary system. In the non-CVBP scenario which the intermediate heat exchanger does not fail by creep, core debris falls into the RV room after reactor vessel failure or evaporation of sodium coolant molten. FPs released from the reactor vessel are retained in the RV room, the primary system room, the containment dome and so on. The hydrogen generated by sodium-concrete reaction and

  13. Source term estimation using air concentration measurements and a Lagrangian dispersion model - Experiments with pseudo and real cesium-137 observations from the Fukushima nuclear accident

    Science.gov (United States)

    Chai, Tianfeng; Draxler, Roland; Stein, Ariel

    2015-04-01

    A transfer coefficient matrix (TCM) was created in a previous study using a Lagrangian dispersion model to provide plume predictions under different emission scenarios. The TCM estimates the contribution of each emission period to all sampling locations and can be used to estimate source terms by adjusting emission rates to match the model prediction with the measurements. In this paper, the TCM is used to formulate a cost functional that measures the differences between the model predictions and the actual air concentration measurements. The cost functional also includes a background term which adds the differences between a first guess and the updated emission estimates. Uncertainties of the measurements, as well as those for the first guess of source terms are both considered in the cost functional. In addition, a penalty term is added to create a smooth temporal change in the release rate. The method is first tested with pseudo observations generated using the Hybrid Single Particle Lagrangian Integrated Trajectory (HYSPLIT) model at the same location and time as the actual observations. The inverse estimation system is able to accurately recover the release rates and performs better than a direct solution using singular value decomposition (SVD). It is found that computing ln(c) differences between model and observations is better than using the original concentration c differences in the cost functional. The inverse estimation results are not sensitive to artificially introduced observational errors or different first guesses. To further test the method, daily average cesium-137 air concentration measurements around the globe from the Fukushima nuclear accident are used to estimate the release of the radionuclide. Compared with the latest estimates by Katata et al. (2014), the recovered release rates successfully capture the main temporal variations. When using subsets of the measured data, the inverse estimation method still manages to identify most of the

  14. Atmospheric discharge and dispersion of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant accident, 2; Verification of the source term and analysis of regional-scale atmospheric dispersion

    OpenAIRE

    寺田 宏明; 堅田 元喜; 茅野 政道; 永井 晴康

    2012-01-01

    Regional-scale atmospheric dispersion simulations were carried out to verify source term of 131I and 137Cs estimated by our previous studies and analyze the atmospheric dispersion during the Fukushima Dai-ichi Nuclear Power Plant accident with measurements of daily and monthly surface depositions over land in Eastern Japan from March 12 to April 30, 2011. The prediction accuracy of daily surface deposition by using the refined source term was mostly within a factor of 10 without apparent bias...

  15. 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)

    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

  16. Detailed source term estimation of atmospheric release during the Fukushima Dai-ichi nuclear power plant accident by coupling atmospheric and oceanic dispersion models

    Science.gov (United States)

    Katata, Genki; Chino, Masamichi; Terada, Hiroaki; Kobayashi, Takuya; Ota, Masakazu; Nagai, Haruyasu; Kajino, Mizuo

    2014-05-01

    Temporal variations of release amounts of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) accident and their dispersion process are essential to evaluate the environmental impacts and resultant radiological doses to the public. Here, we estimated a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data and coupling atmospheric and oceanic dispersion simulations by WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN developed by the authors. New schemes for wet, dry, and fog depositions of radioactive iodine gas (I2 and CH3I) and other particles (I-131, Te-132, Cs-137, and Cs-134) were incorporated into WSPEEDI-II. The deposition calculated by WSPEEDI-II was used as input data of ocean dispersion calculations by SEA-GEARN. The reverse estimation method based on the simulation by both models assuming unit release rate (1 Bq h-1) was adopted to estimate the source term at the FNPP1 using air dose rate, and air sea surface concentrations. The results suggested that the major release of radionuclides from the FNPP1 occurred in the following periods during March 2011: afternoon on the 12th when the venting and hydrogen explosion occurred at Unit 1, morning on the 13th after the venting event at Unit 3, midnight on the 14th when several openings of SRV (steam relief valve) were conducted at Unit 2, morning and night on the 15th, and morning on the 16th. The modified WSPEEDI-II using the newly estimated source term well reproduced local and regional patterns of air dose rate and surface deposition of I-131 and Cs-137 obtained by airborne observations. Our dispersion simulations also revealed that the highest radioactive contamination areas around FNPP1 were created from 15th to 16th March by complicated interactions among rainfall (wet deposition), plume movements, and phase properties (gas or particle) of I-131 and release rates

  17. 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 1: Sections 1-9

    International Nuclear Information System (INIS)

    This report documents the methodology, computational framework, and results of facility accident analyses performed for 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. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. 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. 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

  18. 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 1: Sections 1-9

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

    1995-04-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for 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. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. 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. 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.

  19. An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Bixler, Nathan E.; Wagner, Kenneth Charles.

    2014-03-01

    A methodology for using the MELCOR code with the Latin Hypercube Sampling method was developed to estimate uncertainty in various predicted quantities such as hydrogen generation or release of fission products under severe accident conditions. In this case, the emphasis was on estimating the range of hydrogen sources in station blackout conditions in the Sequoyah Ice Condenser plant, taking into account uncertainties in the modeled physics known to affect hydrogen generation. The method uses user-specified likelihood distributions for uncertain model parameters, which may include uncertainties of a stochastic nature, to produce a collection of code calculations, or realizations, characterizing the range of possible outcomes. Forty MELCOR code realizations of Sequoyah were conducted that included 10 uncertain parameters, producing a range of in-vessel hydrogen quantities. The range of total hydrogen produced was approximately 583kg 131kg. Sensitivity analyses revealed expected trends with respected to the parameters of greatest importance, however, considerable scatter in results when plotted against any of the uncertain parameters was observed, with no parameter manifesting dominant effects on hydrogen generation. It is concluded that, with respect to the physics parameters investigated, in order to further reduce predicted hydrogen uncertainty, it would be necessary to reduce all physics parameter uncertainties similarly, bearing in mind that some parameters are inherently uncertain within a range. It is suspected that some residual uncertainty associated with modeling complex, coupled and synergistic phenomena, is an inherent aspect of complex systems and cannot be reduced to point value estimates. The probabilistic analyses such as the one demonstrated in this work are important to properly characterize response of complex systems such as severe accident progression in nuclear power plants.

  20. Source terms in relation to air cleaning

    International Nuclear Information System (INIS)

    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. 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

    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.

    2015-01-01

    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 (I2 and CH3I) 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 cumulative

  2. 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

    Directory of Open Access Journals (Sweden)

    G. Katata

    2014-06-01

    Full Text Available 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

  3. 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

  4. Long term cooling analysis after Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    The objective of this study is to analyze of the long term cooling after Fukushima Daiichi accident by RELAP5mode3.3 code and to check the validity of the cooling method. In order to simulate the cooling conditions in Fukushima plants after accident, the model is nodalized on the assumption of the existence of steam/liquid leak position from RPV/PCV and the variety of debris distribution in RPV/PCV. As a result, we estimated the debris distribution in RPV by referring plant parameter such as reactor pressure and temperature. In addition, we performed the analysis of the loss of injection water accident for the current cooling system installed in Fukushima Daiichi cite after the earthquake. In this case, we develop simplified nodalization of RPV to analyze temperature behavior of reactor structural materials by using the radiation heat transfer model. (author)

  5. Source term evaluation for accident conditions

    International Nuclear Information System (INIS)

    The Symposium presentations were divided into 5 sessions devoted to the following topics: in-vessel-release (12 papers), retention in the primary circuit (9 papers), ex-vessel release (6 papers), retention in containment (8 papers) and release from the plant (9 papers). In addition, a poster session was held (8 papers) as well as two panel discussions on the following subjects: containment challenges and regulatory implications. The Proceedings contain all the introductory summaries, papers and posters that were presented during the Symposium. A separate abstract was prepared for each of these papers

  6. Influence of Chemistry on source term assessment

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. MELCOR code source term characteristics for fast SBO scenario of OPR1000 plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    Off site consequence analysis in Level 3 PSA is mainly affected by source terms release characteristics of nuclear plant. The severe accidents analysis codes for quantifying the source terms release characteristics, such as MELCOR or MAAP, could be available to provide the detailed information of these characteristics to assess offsite consequence. To utilize these characteristics from severe accident analysis codes, MELCOR code was used in a specific severe accident scenario, i.e., fast station black out (SBO) for OPR1000 plant.

  9. Effects of source term characteristics on Off Site consequence

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    Off site consequence analysis in Level 3 PSA is mainly affected by source term release characteristics of nuclear plant. The severe accident analysis codes for quantifying the source term release characteristics such as MELCOR and MAAP provide detailed information of these characteristics to assess the off site consequence. The aforementioned characteristics, however, have not been considered in the consequence analysis of domestic plants because of large uncertainty in these characteristics so far. Recently, the USNRC SOARCA report showed an approach to utilize detailed source term characteristics provided by MELCOR code to quantify the off site consequence more realistically. Main purpose of this study is to assess effects of the MELCOR source term characteristics on the off site consequence analysis of a domestic nuclear power plant, in a similar fashion to the SOARCA approach. Among many features characterizing source term, the most important one is to determine initial and boundary conditions of atmospheric dispersion such as:- Release amounts of source term - Release time and duration Moreover, plumes features (i.e., radiation clouds) affect atmospheric dispersion that shapes plume characteristics as follows: - Initial dimension of plumes - Plume rise - Deposition of radioactive materials during dispersion Although the current severe accident codes have some limitation in providing the entire source term release characteristics needed in the consequence analysis, the essential information for these features could be obtained from these codes. It is noted that the typical approaches, which generate source term information for the consequence analysis from the severe accident codes, should require a technical manipulation by the experts of consequence analysis. The present effort focused on an identification of insights to utilize source term characteristics of the severe accident codes.

  10. Unconventional sources of plant information for accident management

    International Nuclear Information System (INIS)

    An essential element to accident management is having as clear a picture as is practical of the plant status and thus of the accident and its progress. Effective, appropriate decisions to control and mitigate an accident are dependent on making this assessment of the accident. The objective of this paper is to stimulate consideration of unconventional plant information sources through discussion of specific examples. A plant's condition during an accident can be characterized by plant parameters such as temperatures and pressures and by plant system operational status. For example, core damage is associated with increasing temperatures, pressures, and radiation levels in many different systems and plant areas. Reg. Guide 1.97 instrumentation exists to provide information to allow operators to take specified manual actions (Type A), to indicate whether plant safety functions are being accomplished (Type B), to indicate the potential for breach of barriers to fission product release (Type C), to indicate operability of individual safety systems (Type D), and to indicate the magnitude of radioactive material releases (Type E). Reg. Guide 1.97 instrument range requirements, with the exception of pressure instruments, address conditions up to design basis conditions. Pressure instrument range requirements exceed design basis conditions. During a severe accident, some instruments may not see conditions beyond their design basis. Effective accident management includes the ability to establish a consistent picture of the accident by accumulating information from as many sources as is practical. Operability of systems and components, and non-safety related temperature, radiation, pressure, and water-level indication can be used to directly indicate, measure, or infer plant parameters which confirm, augment or replace those otherwise available. Innovative uses of information sources thus serve to increase the diversity and flexibility of accident data available. Both the

  11. Source term analyses for the Muehleberg nuclear power plant

    International Nuclear Information System (INIS)

    In the study presented here, the source terms for six accident scenarios at the Muehleberg nuclear power station were investigated; namely two low pressure incidents, a high pressure incident, a fire, an earthquake and a plane crash. figs., tabs., 44 refs

  12. Thermochemical considerarations in source term evaluation

    International Nuclear Information System (INIS)

    The calculation of the release of fission products from degraded fuel in a light water reactor core uncovery accident is the first step in determining the overall radiological source term. It is the aim of costly experiments to improve our knowledge about the release behavior of the relevant fission products. Since this depends greatly on their chemistry, a thermodynamic evaluation about compound formation and vaporisation in a fuel-fission product-coolant system should precede such tests. It shows how the volatility of these products may change with test conditions. It will need more reduction of the steam atmosphere to get a noticable release of barium and strontium than to have europium show up. It is very unlikely that ruthenium is significantly released even in a nonreduced steam environment. Molybdenum will be released with the cesium in oxidising and slightly reducing atmospheres. Boron has an effect on the iodine and cesium chemistry. This, however, depends greatly on test or accident conditions. It is practically nonexistent at high steam pressures. Low oxygen potential and high boron content of the atmosphere increase the effect. (orig.)

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

    International Nuclear Information System (INIS)

    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)

  14. 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)

  15. 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

    OpenAIRE

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

    2015-01-01

    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 measure...

  16. 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

    OpenAIRE

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

    2014-01-01

    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 WSP...

  17. Literature study of source term research for PWRs

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Source term calculations for assessing radiation dose to equipment

    International Nuclear Information System (INIS)

    This study examines results of analyses performed with the Source Term Code Package to develop updated source terms using NUREG-0956 methods. The updated source terms are to be used to assess the adequacy of current regulatory source terms used as the basis for equipment qualification. Time-dependent locational distributions of radionuclides within a containment following a severe accident have been developed. The Surry reactor has been selected in this study as representative of PWR containment designs. Similarly, the Peach Bottom reactor has been used to examine radionuclide distributions in boiling water reactors. The time-dependent inventory of each key radionuclide is provided in terms of its activity in curies. The data are to be used by Sandia National Laboratories to perform shielding analyses to estimate radiation dose to equipment in each containment design. See NUREG/CR-5175, ''Beta and Gamma Dose Calculations for PWR and BWR Containments.'' 6 refs., 11 tabs

  20. Selected source term topics. Report to CSNI by an OECD/NEA Group of experts

    International Nuclear Information System (INIS)

    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. 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.

  2. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  3. Estimated long term health effects of the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Cardis, E. [International Agency for Research on Cancer, Lyon (France)

    1996-07-01

    Apart from the dramatic increase in thyroid cancer in those exposed as children, there is no evidence to date of a major public health impact as a result of radiation exposure due to the Chernobyl accident in the three most affected countries (Belarus, Russia, and Ukraine). Although some increases in the frequency of cancer in exposed populations have been reported ,these results are difficult to interpret, mainly because of differences in the intensity and method of follow-up between exposed populations and the general population with which they are compared. If the experience of the survivors of the atomic bombing of Japan and of other exposed populations is applicable, the major radiological impact of the accident will be cases of cancer. The total lifetime numbers of excess cancers will be greatest among the `liquidators` (emergency and recovery workers) and among the residents of `contaminated` territories, of the order of 2000 to 2500 among each group (the size of the exposed populations is 200,000 liquidators and 3,700,000 residents of `contaminated` areas). These increases would be difficult to detect epidemiologically against an expected background number of 41500 and 433000 cases of cancer respectively among the two groups. The exposures for populations due to the Chernobyl accident are different in type and pattern from those of the survivors of the atomic bombing of Japan. Thus predictions derived from studies of these populations are uncertain. The extent of the incidence of thyroid cancer was not envisaged. Since only ten years have lapsed since the accident, continued monitoring of the health of the population is essential to assess the public health impact.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  6. Evaluation of radiation source term for the DUPIC fuel core

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Park, Chang Je

    2004-12-01

    The radiation source term of the DUPIC fuel CANDU reactor was estimated for the total and gap inventories of fission products. The calculation was performed by the ELESTRES code using fuel burnup and linear power distribution obtained from the reactor physics calculations. The radiation source term represented by the fission products gap inventory was 46912 TBq for the 1/4 DUPIC fuel core, while it was 75448 TBq for the natural uranium core. Such a reduction of the radiation source term for the DUPIC fuel core can be attributed to the lower linear power of the DUPIC fuel bundle caused by the flattened power distribution of the DUPIC fuel core which adopts a 2-bundle shift refueling scheme. It is therefore expected that the consequence of the loss of coolant accident for the DUPIC fuel core could be weak when compared to the natural uranium core from the viewpoint of radiation doses to the public.

  7. 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)

    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

  8. Modified ensemble Kalman filter for nuclear accident atmospheric dispersion: Prediction improved and source estimated

    International Nuclear Information System (INIS)

    Highlights: • A modified ensemble Kalmen filter data assimilation method is proposed. • The method can consider four main uncertain parameters in the puff model. • The prediction of radioactive material atmospheric dispersion is improved. • The source release rate and plume rise height are successfully reconstructed. • It can shorten the time lag in the response of ensemble Kalmen filter. - Abstract: 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

  9. Running the source term code package in Elebra MX-850

    International Nuclear Information System (INIS)

    The source term package (STCP) is one of the main tools applied in calculations of behavior of fission products from nuclear power plants. It is a set of computer codes to assist the calculations of the radioactive materials leaving from the metallic containment of power reactors to the environment during a severe reactor accident. The original version of STCP runs in SDC computer systems, but as it has been written in FORTRAN 77, is possible run it in others systems such as IBM, Burroughs, Elebra, etc. The Elebra MX-8500 version of STCP contains 5 codes:March 3, Trapmelt, Tcca, Vanessa and Nava. The example presented in this report has taken into consideration a small LOCA accident into a PWR type reactor. (M.I.)

  10. Light water reactor severe accident seminar. Seminar presentation manual

    International Nuclear Information System (INIS)

    The topics covered in this manual on LWR severe accidents were: Evolution of Source Term Definition and Analysis, Current Position on Severe Accident Phenomena, Current Position on Fission Product Behavior, Overview of Software Models Used in Severe Accident Analysis, Overview of Plant Specific Source Terms and Their Impact on Risk, Current Applications of Severe Accident Analysis, and Future plans

  11. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  12. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  13. HTGR Mechanistic Source Terms White Paper

    Energy Technology Data Exchange (ETDEWEB)

    Wayne Moe

    2010-07-01

    The primary purposes of this white paper are: (1) to describe the proposed approach for developing event specific mechanistic source terms for HTGR design and licensing, (2) to describe the technology development programs required to validate the design methods used to predict these mechanistic source terms and (3) to obtain agreement from the NRC that, subject to appropriate validation through the technology development program, the approach for developing event specific mechanistic source terms is acceptable

  14. Calculation and analysis of radioactive source term in PWR assemblies

    International Nuclear Information System (INIS)

    Background: When fission occurs in fuel of reactor core, it produces a large amount of radioactive materials, which may cause harm to the environment and human health. Purpose: The radioactive materials in fuel could provide input data for shielding design of reactor coolant radioactive source term, analysis of accident source term and radioactive consequence assessment. Methods: The calculation of radioactive source in fuel was studied for pressurized water reactor: the calculation methods and models were established using ORIGEN-S, and the difference of nuclides radioactivity under different burnup was also studied. The effect of different versions of ENDF/B cross-section database on the calculation results was analyzed, so as to provide a basis for the calculation of radioactive source in fuel. Results: The results showed that the method established by ORIGEN-ARP was more suitable for calculating radioactive source term in fuel assemblies and the different versions of ENDF/B database had a great impact on radioactivity calculation. Conclusion: Based on the ENDF/B-VII database, using ORIGEN-ARP to calculate radioactive source term in fuel assemblies could not only improve efficiency, but also improve the calculation accuracy. (authors)

  15. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    OpenAIRE

    Andrej Prošek; Leon Cizelj

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  16. Emergency drinking water treatment during source water pollution accidents in China: origin analysis, framework and technologies.

    Science.gov (United States)

    Zhang, Xiao-Jian; Chen, Chao; Lin, Peng-Fei; Hou, Ai-Xin; Niu, Zhang-Bin; Wang, Jun

    2011-01-01

    China has suffered frequent source water contamination accidents in the past decade, which has resulted in severe consequences to the water supply of millions of residents. The origins of typical cases of contamination are discussed in this paper as well as the emergency response to these accidents. In general, excessive pursuit of rapid industrialization and the unreasonable location of factories are responsible for the increasing frequency of accidental pollution events. Moreover, insufficient attention to environmental protection and rudimentary emergency response capability has exacerbated the consequences of such accidents. These environmental accidents triggered or accelerated the promulgation of stricter environmental protection policy and the shift from economic development mode to a more sustainable direction, which should be regarded as the turning point of environmental protection in China. To guarantee water security, China is trying to establish a rapid and effective emergency response framework, build up the capability of early accident detection, and develop efficient technologies to remove contaminants from water. PMID:21133359

  17. Strengthening long term control over radioactive sources

    International Nuclear Information System (INIS)

    The traditional focus of the regulation of radioactive sources is the protection of workers and the public from the misuse of sources and from accidents. Security measures were also a concern, but with the principal aim of preventing petty theft or accidental loss. Our concern, of course, is that a high risk radioactive source might be married with conventional explosives and used in a radiological dispersal device (RDD). Means must be found to protect the public from the use of high risk radioactive sources in an RDD. The task may appear daunting at first because of the widespread use of radioactive sources throughout the world. Compounding the problem is the fact that there also is a general lack of effective domestic controls on even high risk radioactive sources. The IAEA has noted that more than 100 countries lack effective control over radiation sources because most do not have the required infrastructure. The US Nuclear Regulatory Commission (NRC), like its counterparts in other countries, has found that modification of our regulatory programme to account for the terrorist threat is necessary. Although the work on this problem is still under way, some of the components are underlined that are believed to be the elements of an effective regulatory programme to counteract the RDD threat. The aim is a programme that achieves an appropriate balance of safety, security, public benefit and economic feasibility. The main objectives covered in this presentation cover the following topics: Categorization; Security measures; Imports/exports; Disposal; Orphan sources; Emergency response

  18. Estimation Of Source Term For Indian PHWRS (KAPS) As Part Of PSA Level-2 Study

    International Nuclear Information System (INIS)

    Source Term (ST) is generally known as the amount of the radio-nuclides(fission products along with activation and Actinides) that can be released from a nuclear power plant in an accident. The ST can be more accurately defined as the quantity, timing, composition, chemical and physical form of radio-nuclides. The amount of radio-nuclides is a fundamental parameter to estimate the consequences of an accident on individuals and environment. A quantitative estimation of the ST is of importance for assessing the effectiveness of safety design features and for the planning of post accident emergency measures in the public domain. The PSA Level-1 study for IPHWRs(KAPS) was completed in 2002 and an attempt was made to estimate the ST for different accident scenarios as part of PSA Level-2 study. The scope of this paper is limited to estimate the ST for Indian Pressurized Heavy Water Reactors (IPHWRs) in accident conditions. (authors)

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

    International Nuclear Information System (INIS)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. The long-term problems of contaminated land: Sources, impacts and countermeasures

    International Nuclear Information System (INIS)

    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'')

  2. 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'').

  3. The source term experiments project deposition sample characterization: Interim report

    International Nuclear Information System (INIS)

    A series of four experiments aimed at characterizing the radiological source term associated with postulated severe light water reactor accidents has been conducted at Argonne National Laboratory's TREAT Facility. The experiments were designed to provide dta regarding the physicochemical properties, near the point of origin, of the biologically important volatile fission products released early in such accidents. The experimental vehicles were equipped to capture representative fission products released from fuel rods undergoing severe cladding degradation in a steam environment. Test conditions of pressure, fuel heatup rate, and steam flow were selected to simulate conditions predicted for hypothetical reactor accident sequences. One of the main components of the experimental vehicle's aerosol characterization system, common to all four tests, was a sample tree. It served to suspend coupons composed of a variety of materials, some typical of reactor structures, into the fission-product-laden steam flow. These coupons collected particulates and condensing vapors. Coupons frome ach of the four tests have been examined using scanning electron microscopy and associated energy dispersive x-ray analysis. The results of these initial sample examinations are presented. They are preceeded by a brief description of the test series and the experimental vehicle. Also included is a discussion of planned posttest examinations of other aerosol characterization system components and the test fuel as well as further examinations of the sample tree coupons. Results of the additional examinations thermal-hydraulic data, and interpretation of the information for each test will be included in future reports

  4. Development of Reference Source Terms for EU-APR1400

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. 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 team 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, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

  6. The rehabilitation strategies in agriculture in the long term after the Chernobyl NPP accident

    International Nuclear Information System (INIS)

    The experience gained in the aftermath of the severe radiation accidents shows that in the case of large-scaled radionuclide contamination the limitation of internal radiation doses to people by means of restoration of agricultural lands is more realistic than reduction of levels of external irradiation. Therefore, the problems connected with the optimal restoration strategies of agricultural land subjected to radioactive contamination after the Chernobyl accident are of crucial importance. The justification of the approach for the estimation of the effectiveness of countermeasure strategies in the long term after the Chernobyl accident, based on the classification of farms by contamination density and risk of the exceeding of radiological standards, restricting the use of agricultural products, is presented. For each class of the farms the ranking of rehabilitation options and the time periods when their application would be of importance are given. Comparative analysis of the rehabilitation strategies, which are different in their effectiveness and cost, is provided. (author)

  7. Source Term Analysis in Severe Accident Induced by Large Break Loss of Coolant Accident Coincident With Ship Blackout for Ship Reactor%船用堆大破口失水叠加全船断电严重事故源项分析

    Institute of Scientific and Technical Information of China (English)

    张彦招; 张帆; 赵新文; 郑映峰

    2013-01-01

    以某船用压水堆为研究对象,采用M ELCOR程序建立事故分析模型,研究大破口失水事故叠加全船断电严重事故下放射性裂变产物的行为,着重分析了惰性气体和CsI的释放、迁移、滞留特点及在堆舱内的分布。结果表明,83.12%惰性气体从堆芯释放出来,并主要存在于堆舱的气空间;83.08%的CsI从堆芯释放出来,其中,72.66%滞留在堆坑熔融物与一回路内,27.34%释放到堆舱内,并主要溶解于舱底水池中。本文分析结果可为舱室剂量评估、核应急管理提供依据。%Using MELCOR code ,the accident analysis model was established for a ship reactor .The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout . The research mainly focused on the behaviors of release ,transport ,retention and the final distribution of inert gas and CsI . T he results show that 83.12% of inert gas releases from the core , and the most of inert gas exists in the containment . About 83.08% of CsI release from the core ,72.66% of w hich is detained in the debris and the primary system ,and 27.34% releases into the containment . The results can give a reference for the evaluation of cabin dose and nuclear emergency management .

  8. 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.

  9. New source terms: what do they tell us about engineered safety feature performance

    International Nuclear Information System (INIS)

    The accident behavior models which are the basis of engineered safety feature design are generally simple, non-mechanistic and concentrated on volatile radioiodine. Now data from source term studies show that models should be more mechanistic and look at other species than volatile iodine. A complete reevaluation of engineered safety features is needed

  10. Licensing design basis source term update for the Evolutionary Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The purpose of this report is to document the technical basis for a licensing source term update for the Evolutionary Advanced Light Water Reactor (ALWR) which will make the source term more physically realistic. While TID [Technical Information Document] 14844 and related regulatory guidance have served the industry well, much has been learned about source term over the last 30 years, and the ALWR Requirements Document provides an opportunity to incorporate this experience by updating the licensing source term. Further, the source term update will provide an improved basis for evolutionary ALWR accident mitigation design. Results of this work indicate that the fission product release magnitude to containment is slightly less than TID 14844 for noble gas, iodine, and semi and low volatiles, but somewhat higher for cesium and tellurium. Release timing is delayed by one hour or more after the accident initiation. The chemical form of iodine is largely aerosol with significantly less organic iodine compared to regulatory guidance which specifies mostly elemental and a relatively large fraction of organic. Containment spray aerosol removal rate was determined to be significantly higher than specified in regulatory guidance. Finally, BWR suppression pool decontamination factor was determined to be less effective than allowed by regulatory guidance early in the accident (due to the delayed release noted above) and more effective than that allowed by regulatory guidance later in the accident. It is recognized by the ALWR program that the source term update could be taken further in the direction of a physically-based source term. 47 refs., 4 figs., 11 tabs

  11. Health physics evaluation of an accident involving acute overexposure to a radiography source

    International Nuclear Information System (INIS)

    An accident, involving the loss of an iridium-192 radiographic source and the subsequent serious overexposure of a third party, is described. Health physics aspects, particularly dosimetrical aspects are addressed and compared with results obtained by means of chromosome aberration dosimetry. Details are provided on the medical observations and treatment of the patient

  12. Ablation Properties of the Carbon-Based Composites Used in Artificial Heat Source Under Fire Accident

    Institute of Scientific and Technical Information of China (English)

    TANG; Xian; HUANG; Jin-ming; ZHOU; Shao-jian; LUO; Zhi-fu

    2012-01-01

    <正>The ablation properties of the carbon-based composites used in artificial heat source under fire accident were investigated by the arc heater. In this work, we tested the carbon-based composites referring to Fig. 1. Their linear/mass ablation ratio and ablation morphologies were studied. The results showed that the carbon-based composites used in artificial heat source behaved well

  13. Transport accident frequency data, their sources and their application in risk assessment

    International Nuclear Information System (INIS)

    Base transport accident frequency data and sources of these data are presented. Both generic information and rates specific to particular routes or packages are included. Strong packages, such as those containing significant quantities of radioactive materials, will survive most of the accidents represented by these base frequencies without a containment breach. The association of severity probability distributions with a base frequency, and package and contents response, leading to the quantification of release frequency and magnitude, are often more important in risk assessment than the base frequency itself. This paper therefore also includes brief comments on techniques adopted to utilize the base frequencies. This paper reports an accident frequency data survey undertaken at the end of 1986. It has not been updated to take account of work published between January 1987 and the Report publication date. (author)

  14. TREAT light water reactor source term experiments program

    International Nuclear Information System (INIS)

    Pre-test calculations indicate that, for the STEP-1 (Source Term Experimental Program) test, cladding temperatures in excess of 42000F can be reached on a heatup transient similar to that of the AD accident sequence in a 20-min test duration. This is well above the Zircaloy melting point of approx. 33500F and should provide a degree of cladding disruption sufficient to allow a singificant release of products from the fuel into the flowing steam. The same temperature range can be reached in a 60-min-duration run to simulate the TQUW sequence for the STEP-2 test. The complete paper will present initial experimental results from these two tests and perhaps from the two TMLB' simulations run without and with control rod material in STEP-3 and STEP-4, respectively

  15. Source-term evaluations from recent core-melt experiments

    International Nuclear Information System (INIS)

    Predicted consequences of hypothetical severe reactor accidents resulting in core meltdown appear to be too conservatively projected because of the simplistic concepts often assumed for the intricate and highly variable phenomena involved. Recent demonstration work on a modest scale (1-kg) has already revealed significant variations in the mode and temperature for clad failure, in the rates of formation of zirconium alloys, in the nature of the UO2-ZrO2 eutectic mixtures, and in aerosol generation rates. The current series of core-melt demonstration experiments (at the 10-kg scale) seem to confirm that an increase in size of the meltdown mass will lead to an even further reduction in the amount of vaporized components. Source terms that are based on older release evaluations could be up to an order of magnitude too large. 6 refs., 6 figs., 2 tabs

  16. 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)

  17. Analysis of Hydrogen Source Term and Effectiveness of Hydrogen Control in Thousand Megawatt PWR Severe Accident%百万千瓦级压水堆严重事故下氢气源项及氢气空制有效性分析

    Institute of Scientific and Technical Information of China (English)

    邹杰; 佟立丽; 曹学武; 顾健; 薛峻峰; 江宇; 郝禄禄; 仇苏辰; 刘力

    2013-01-01

    针对百万千瓦级压水堆核电厂大型干式安全壳在严重事故情况下的氢气风险控制,建立了一体化事故分析模型,分别对大破口失水事故(LB-LOCA)、中破口失水事故(MB-LOCA)、小破口失水事故(SB-LOCA)、全厂断电事故(SBO)、蒸汽发生器(SG)传热管破裂事故(SGTR)以及主蒸汽管道破裂事故(MSLB)进行事故进程计算以及氢气源项分析.相对于其他事故序列,LB-LOCA下堆芯快速熔化,锆-水反应产生氢气的速率快,可以作为安全壳内氢气风险控制有效性分析的代表性事故序列.分析表明,严重事故情况下在安全壳中安装一定数量的非能动氢气复合器(PARs)能够有效去除安全壳中的氢气,消除氢气燃烧或爆炸的风险,保持安全壳的完整性.%The integrated severe accident analysis model of 100 MW PWR NPP is built to analyze the hydrogen risk under severe accidents.Large break loss of coolant accident (LB-LOCA),medium break loss of coolant accident (MB-LOCA),small break loss of coolant accident (SB-LOCA),station blackout (SBO),steam generator tube rupture (SGTR) and main steam line break (MSLB) are chosen as typical severe accident sequences to analyze the hydrogen source.Considering the hydrogen quantity of 100% zirconium water reaction,the LB-LOCA is selected as a representative sequence to evaluate the hydrogen mitigation system.The results show that a certain number of PARs could remove hydrogen and oxygen effectively,and protect the containment integrity against hydrogen deflagration or detonation.

  18. BWR Source Term Generation and Evaluation

    International Nuclear Information System (INIS)

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-000061. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality

  19. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

    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. Hazardous constituent source term. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-11-17

    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.

  1. 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)

  2. IMPACTS OF SOURCE TERM HETEROGENEITIES ON WATER PATHWAY DOSE.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.; GUSKOV, A.; POSKAS, P.; RUPERTI, N.; HANUSIK, V.; ET AL.

    2004-09-15

    specific activities and small physical sizes and for which a solution has to be found in term of long-term disposal. Together with their casing and packaging, they are one form of heterogeneous waste; many other forms of waste with heterogeneous properties exist. They may arise in very small quantities and with very specific characteristics in the case of small producers, or in larger streams with standard characteristics in others. This wide variety of waste induces three main different levels of waste heterogeneity: (1) hot spot (e.g. disused sealed sources); (2) large item inside a package (e.g. metal components); and (3) very large items to be disposed of directly in the disposal unit (e.g. irradiated pipes, vessels). Safety assessments generally assume a certain level of waste homogeneity in most of the existing or proposed disposal facilities. There is a need to evaluate the appropriateness of such an assumption and the influence on the results of safety assessment. This need is especially acute in the case of sealed sources. There are many cases where are storage conditions are poor, or there is improper management leading to a radiological accident, some with significant or detrimental impacts. Disposal in a near surface disposal facility has been used in the past for some disused sealed sources. This option is currently in use for others sealed sources, or is being studied for the rest of them. The regulatory framework differs greatly between countries. In some countries, large quantities of disused sealed sources have been disposed of without any restriction, in others their disposal is forbidden by law. In any case, evaluation of the acceptability of disposal of disused sealed sources in near surface disposal facility is of utmost importance.

  3. Long-term passive CANDU containment response after a design-basis accident

    International Nuclear Information System (INIS)

    A passive CANDU reg-sign containment system, currently being developed, is aimed at limiting the consequences of a postulated accident, by ensuring the structural integrity of the containment building and limiting fission-product release to within siting dose limits, without operator action or reliance on ac power for up to 3 d. All main functions of the containment system, i.e. energy removal, hydrogen mitigation, and fission-product retention, are to be accomplished passively. The passive CANDU containment relies on the passive emergency water system (PEWS) for energy removal after an accident and on passive autocatalytic recombiners (PAR) for hydrogen removal. The key feature of this concept, is a recirculating, buoyancy-driven flow through the recombiners and the tube banks of the PEWS. This paper presents preliminary design calculations for the PEWS tank and tube banks and a simulation of the long-term passive containment response, based on the current CANDU-6 containment, to a large loss-of-coolant/loss-of- emergency coolant injection (LOCA/LOECI) using the GOTHIC code. It is shown that a 1500-M3 PEWS tank, connected to tube banks with a total surface area of 1800 m2, can limit the second pressure peak to about 300 kPa(a) if a recirculating flow is established in the containment building. The PEWS tank water is boiling in the long term, and the peak containment temperature is 114 degrees C. 6 refs., 4 figs

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

    International Nuclear Information System (INIS)

    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

  5. Tchernobyl accident

    International Nuclear Information System (INIS)

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given

  6. Use of source term code package in the ELEBRA MX-850 system

    International Nuclear Information System (INIS)

    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.)

  7. Workshop on short-term health effects of reactor accidents: Chernobyl

    International Nuclear Information System (INIS)

    The high-dose early-effects research that has been continued has been done in the context of infrequent accidents with large radiation sources and the use of bone marrow transfusions for treating malignancies, especially leukemia. It thus seemed appropriate to bring together those who have done research on and have had experience with massive whole-body radiation. The objectives were to review what is known about the acute effects of whole-body irradiation, to review the current knowledge of therapy, and particularly of the diagnostic and immunologic problems encountered in bone marrow therapy, and to compare this knowledge with observations made to date on the Chernobyl accident radiation casualties. Dr. Robert Gale, who had helped to care for these casualties, was present at the Workshop. It was hoped that such a review would help those making continuing clinical and pathological observations on the Chernobyl casualties, and that these observations would provide a basis for recommendations for additional research that might result in improved ability to manage successfully this type of severe injury

  8. Workshop on short-term health effects of reactor accidents: Chernobyl

    Energy Technology Data Exchange (ETDEWEB)

    1986-08-08

    The high-dose early-effects research that has been continued has been done in the context of infrequent accidents with large radiation sources and the use of bone marrow transfusions for treating malignancies, especially leukemia. It thus seemed appropriate to bring together those who have done research on and have had experience with massive whole-body radiation. The objectives were to review what is known about the acute effects of whole-body irradiation, to review the current knowledge of therapy, and particularly of the diagnostic and immunologic problems encountered in bone marrow therapy, and to compare this knowledge with observations made to date on the Chernobyl accident radiation casualties. Dr. Robert Gale, who had helped to care for these casualties, was present at the Workshop. It was hoped that such a review would help those making continuing clinical and pathological observations on the Chernobyl casualties, and that these observations would provide a basis for recommendations for additional research that might result in improved ability to manage successfully this type of severe injury.

  9. Decision making framework for application of forest countermeasures in the long term after the Chernobyl accident

    International Nuclear Information System (INIS)

    After the ChNPP accident a very large part of the territories covered by natural and artificial forests are contaminated with long-lived radionuclides, especially 137Cs. To protect people against exposure associated with forest contamination in the most affected regions of the NIS countries, countermeasures have been developed and recommended for the forest management. The paper presents a decision making framework to optimise forest countermeasures in the long term after the ChNPP accident. The approach presented is based on the analysis of the main exposure pathways and application of radiological, socio-economical and ecological criteria for the selection of optimal countermeasures strategies. Because of the diversity of these criteria modern decision support technologies based on multi-attributive analysis were applied. The results of the application of this approach are presented in a selected study area (Novozybkov district, Bryansk region, Russian Federation). The results prove and emphasize the need for a flexible technique to provide the optimised forest countermeasures taking into account radioecological, social and economic features of contaminated forests

  10. Source term characterization for the radioactive contamination of a river - reservoir system

    International Nuclear Information System (INIS)

    The source term for radioactive contamination of the Doamnei-Arges river - reservoir system is more than 90% composed by 58 Co and 60 Co from TRIGA reactors cooling system. Measurable amounts of 137 Cs from Chernobyl accident fallout (1986) is coming into the system from atmosphere (resuspension) and from the catchment area (soil erosion). In this paper, a quantitative analysis of the three contamination pathways is performed. (authors)

  11. Management of severe accidents

    International Nuclear Information System (INIS)

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  12. TRIGA MARK-II source term

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my; Rawi, M. Z. M. Rawi, E-mail: mark-dennis@nuclearmalaysia.gov.my; Abu, M. P., E-mail: mark-dennis@nuclearmalaysia.gov.my [Bahagian Teknologi Reaktor, Agensi Nuklear Malaysia, 43000 Kajang (Malaysia)

    2014-02-12

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  13. TRIGA MARK-II source term

    Science.gov (United States)

    Usang, M. D.; Hamzah, N. S.; J. B., Abi M.; M. Z., M. Rawi; Abu, M. P.

    2014-02-01

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  14. Transportation accidents

    International Nuclear Information System (INIS)

    Predicting the possible consequences of transportation accidents provides a severe challenge to an analyst who must make a judgment of the likely consequences of a release event at an unpredictable time and place. Since it is impractical to try to obtain detailed knowledge of the meteorology and terrain for every potential accident location on a route or to obtain accurate descriptions of population distributions or sensitive property to be protected (data which are more likely to be more readily available when one deals with fixed-site problems), he is constrained to make conservative assumptions in response to a demanding public audience. These conservative assumptions are frequently offset by very small source terms (relative to a fixed site) created when a transport vehicle is involved in an accident. For radioactive materials, which are the principal interest of the authors, only the most elementary models have been used for assessing the consequences of release of these materials in the transportation setting. Risk analysis and environmental impact statements frequently have used the Pasquill-Gifford/gaussian techniques for releases of short duration, which are both simple and easy to apply and require a minimum amount of detailed information. However, after deciding to use such a model, the problem of selecting what specific parameters to use in specific transportation situations still presents itself. Additional complications arise because source terms are not well characterized, release rates can be variable over short and long time periods, and mechanisms by which source aerosols become entrained in air are not always obvious. Some approaches that have been used to address these problems will be reviewed with emphasis on guidelines to avoid the Worst-Case Scenario Syndrome

  15. Risk comparisons based on representative source terms with the NUREG-1150 results

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-01

    Standardized source terms, based on a specified release of fission products during potential accidents at commercial light water nuclear reactors, have been used for a long time for regulatory purposes. The siting of nuclear power plants, for example, which is governed by Part 100 of the Code of Federal Regulations Title 10, has utilized the source term recommended in TID-14844 supplemented by Regulatory Guides 1.3 and 1.4 and the Standard Review Plan. With the introduction of probabilistic risk assessment (PRA) methods, the source terms became characterized not only by the amount of fission products released, but also by the probability of the release. In the Reactor Safety Study, for example, several categories of source terms, characterized by release severity and probability, were developed for both pressurized and boiling water reactors (PWRs and BWRs). These categories were based on an understanding of the likely paths and associated phenomenology of accident progression following core damage to possible failure of the containment and release to the environment.

  16. Diagnosis and prognosis of the source term by the French Safety Institut during an emergency on a PWR

    International Nuclear Information System (INIS)

    The French approach for the diagnosis and the prognosis of the source term during an accident on a PWR is presented and the tools which have been developed to implement this approach at the Institute for Nuclear Protection and Safety (IPSN) are described. (author). 2 refs, 3 figs

  17. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  18. A Perspective on Long-Term Recovery Following the Fukushima Nuclear Accident - 12075

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S.Y. [Environmental Science Division, Argonne National Laboratory, Argonne, IL (United States)

    2012-07-01

    The tragic events at the Fukushima Daiichi Nuclear Power Station began occurring on March 11, 2011, following Japan's unprecedented earthquake and tsunami. The subsequent loss of external power and on-site cooling capacity severely compromised the plant's safety systems, and subsequently, led to core melt in the affected reactors and damage to spent nuclear fuel in the storage pools. Together with hydrogen explosions, this resulted in a substantial release of radioactive material to the environment (mostly Iodine-131 and Cesium- 137), prompting an extensive evacuation effort. The latest release estimate places the event at the highest severity level (Level 7) on the International Nuclear Event Scale, the same as the Chernobyl accident of 1986. As the utility owner endeavored to stabilize the damaged facility, environmental contamination continued to propagate and affect every aspect of daily life in the affected region of Japan. Elevated levels of radioactivity (mostly dominated by Cs-137 with the passage of time) were found in soil, drinking water, vegetation, produce, seafood, and other foodstuffs. An estimated 80,000 to 90,000 people were evacuated; more evacuations are being contemplated months after the accident, and a vast amount of land has become contaminated. Early actions were taken to ban the shipment and sale of contaminated food and drinking water, followed by later actions to ban the shipment and sale of contaminated beef, mushrooms, and seafood. As the event continues to evolve toward stabilization, the long-term recovery effort needs to commence - a process that doubtless will involve rather complex decision-making interactions between various stakeholders. Key issues that may be encountered and considered in such a process include (1) socio-political factors, (2) local economic considerations, (3) land use options, (4) remediation approaches, (5) decontamination methods, (6) radioactive waste management, (7) cleanup levels and options

  19. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F.; Gauthier, G.; Carlin, F. [and others

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  20. Peculiarities of the clinical course of radiation sickness and organizational decisions for radiation accidents with beta-gamma sources

    International Nuclear Information System (INIS)

    The analysis of a number of recent large scale accidents involving beta-gamma sources in the last 40 years, such as those of the Marshall Islands (1954); Windscale, UK (1957); Chernobyl, USSR (1986) and Goiania, Brazil (1987) demonstrates the predominance and importance of health and social impacts. (author)

  1. Literature review on metallic fuel source term for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Nam Duk; Bae, Moo Hoon; Shin, An Dong; Huh, Chang Wook [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2012-10-15

    Source term is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment, following a severe accident where a significant portion of the reactor core has melted. 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. Apart from assessing the radiological consequences for siting, it is also important for designing filtering systems and even reactor components. Overly conservative source term for light water reactor, TID 14844 demands for very fast closure of main steam isolation valves, rapid startup of emergency diesels, and safety systems designed to mitigate gaseous iodine. In spite of this importance, most of the knowledge we have for SFR source term comes from the research performed before 1980s. Moreover, majority of the work on metallic fuels was done during the late 1950's through the 1960's. This paper reviews and summarizes the main characteristics of SFR source terms based on the available literatures.

  2. Environmental radiation safety source term evaluation program

    International Nuclear Information System (INIS)

    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 238PuO2 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 238PuO2 pellets to conditions found 60 feet below the ocean surface. Ultrafilterability studies measured the solubility of 238PuO2 as a function of time, temperature, suspension concentration and molality of solvent

  3. Optimum target source term estimation for high energy electron accelerators

    Science.gov (United States)

    Nayak, M. K.; Sahu, T. K.; Nair, Haridas G.; Nandedkar, R. V.; Bandyopadhyay, Tapas; Tripathi, R. M.; Hannurkar, P. R.

    2016-05-01

    Optimum target for bremsstrahlung emission is defined as the thickness of the target material, which produces maximum bremsstrahlung yield, on interaction of electron with the target. The bremsstrahlung dose rate per unit electron beam power at a distance of 1 m from the target material gives the optimum target source term. In the present work, simulations were performed for three different electron energies, 450, 1000 and 2500 MeV using EGSnrc Monte-Carlo code to determine the optimum thickness. An empirical relation for optimum target as a function of electron energy and atomic number of the target materials is found out from results. Using the simulated optimum target thickness, experiments are conducted to determine the optimum target source term. For the experimental determination, two available electron energies, 450 MeV and 550 MeV from booster synchrotron of Indus facility is used. The optimum target source term for these two energies are also simulated. The experimental and simulated source term are found to be in very good agreement within ±3%. Based on the agreement of the simulated source term with the experimental source term at 450 MeV and 550 MeV, the same simulation methodology is used to simulate optimum target source term up to 2500 MeV. The paper describes the simulations and experiments carried out on optimum target bremsstrahlung source term and the results obtained.

  4. The long-term cooling of the RU VVER in the conditions of design beyond basis accident

    International Nuclear Information System (INIS)

    MCC rupture on outlet, inlet the reactor (with coolant double end leakage); MCC rupture in the bottom part of a loop seal (with coolant double end leakage); Rupture of the connecting pipeline of HA (JNG10-40) of the ECCS (D = 300 mm); Loss of an alternating current sources, failure of safety active systems for more than 24 hours, failure of all diesel-generators; emergency supply from accumulators; Water injection from a fuel storage pool begins after emptying of tanks GE-2 (JNG50-80) with the total mass flow rate of 3.2 kg/s. The RELAP5/ANGAR code was used. The sequence of events and the work of the systems at a guillotine rupture of the MCC is described. The computing analysis of the beyond design accident which includes MCC rupture with loss of all sources of an alternating current, including diesel engines-generators, demonstrated the following: (i) the intervention of the safety systems of the NVNPP-2 project meets all Russian and international requirements to localizing functions by NVNPP-2 containment at beyond design accidents with leaks from the reactor facility; (ii) water injection from ST after the termination of work; (iii) GE-2 provides reliable core cooling for the beyond design basis accident scenarios during 74-88 hours in dependence on the leak size and location. (P.A.)

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  10. 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.

  11. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  12. Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool

    Energy Technology Data Exchange (ETDEWEB)

    Courageot, Estelle; Sayah, Rima; Huet, Christelle [External Dosimetry Department, Institute for Radiological Protection and Nuclear Safety (IRSN), Radiological Protection and Human Health Division, Ionizing Radiation Dosimetry Laboratory, IRSN/DRPH/SDE, BP 17, 92262 Fontenay-aux-Roses Cedex (France)], E-mail: estelle.courageot@irsn.fr

    2010-05-07

    Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model. (note)

  13. Utility view of the source term and air cleaning

    International Nuclear Information System (INIS)

    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

  14. Evaluation to a long term remediation actions after Goiania radiological accident; Avaliacao a longo prazo das acoes de remediacao apos o acidente radiologico de Goiania

    Energy Technology Data Exchange (ETDEWEB)

    Rochedo, Elaine R.R.; Rio, Monica A. Pires do; Coutinho, Celia M.C.; Acar, Maria E.D.; Romeiro, Carlos H. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    Ten years after the Goiania radiological accident, the results obtained by the IRD Environmental Monitoring Program are compared to the values adopted for establishing the intervention levels at the time of the accident occurrence (1987), and to the values of the parameters obtained by European countries, after the Chernobyl accident. Significant differences were observed in parameter values, particularly, those related to a long term prediction of the contamination behaviour in an urban area. This paper shows the importance of the survey for the environmental behaviour of pollutants in tropical climate conditions. (author)

  15. 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.

  16. Long term simulation of {sup 137}Cs radioactivity in the regional ocean following the Fukushima Daiichi nuclear power plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Tsumune, D.; Tsubono, T.; Misumi, K.; Yoshida, Y.; Hayami, H. [Central Research Institute of Electric Power Industry (Japan); Aoyama, M. [Meteorological Research Institute (Japan); Uematsu, M. [University of Tokyo (Japan); Maeda, Y. [CERES, Inc. (Japan)

    2014-07-01

    A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition. A regional-scale simulation of {sup 137}Cs activity in the ocean offshore of Fukushima was carried out, the sources of radioactivity being direct release, atmospheric deposition, and the inflow of {sup 137}Cs deposited on the ocean by atmospheric deposition outside the domain of the model for more than two years. Direct releases of {sup 131}I, {sup 134}Cs, and {sup 137}Cs were estimated for 1 year after the accident by comparing simulated results and measured activities. The estimated total amounts of directly released {sup 131}I, {sup 134}Cs, and {sup 137}Cs were 11.1±2.2 PBq, 3.5±0.7 PBq, and 3.6±0.7 PBq, respectively. The contributions of each source were estimated by analysis of {sup 131}I/{sup 137}Cs and {sup 134}Cs/{sup 137}Cs activity ratios and comparisons between simulated results and measured activities of {sup 137}Cs. Simulated {sup 137}Cs activities attributable to direct release were in good agreement with measured activities close to the accident site, a result that implies that the estimated direct release rate was reasonable, while simulated {sup 137}Cs activities attributable to atmospheric deposition were low compared to measured activities. The rate of atmospheric deposition onto the ocean was underestimated because of a lack of measurements of deposition onto the ocean when atmospheric deposition rates were being estimated. Measured {sup 137}Cs activities attributable to atmospheric deposition helped to improve the accuracy of simulated atmospheric deposition rates. Simulated {sup 137}Cs activities attributable to the inflow of {sup 137}Cs deposited onto the ocean outside the domain of the model were in good agreement with measured activities in the open ocean within the

  17. MIGRATORY GAME BIRDS AS A SOURCE OF PUBLIC EXPOSURE FROM THE FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT

    Directory of Open Access Journals (Sweden)

    I. P. Stamat

    2011-01-01

    Full Text Available This article examines assessments of the impact of the Fukushima nuclear power plant accident on exposure of the Russian Federation population related to the seasonal migration of game birds. Intake of artificial radionuclides with meat of migratory game birds is shown to be one of the major pathways for the population exposure in the Far Eastern region of the country.

  18. Generalities on nuclear accidents and their short-dated and middle-dated management; Generalites sur les accidents nucleaires et leur gestion a court terme et a long terme

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    All the nuclear activities present a radiation risk. The radiation exposure of the employees or the public, may occur during normal activity or during an accident. The IRSN realized a document on this radiation risk and the actions of protection. The sanitary and medical aspects of a radiation accident are detailed. The actions of the population protection during an accident and the post accident management are also discussed. (A.L.B.)

  19. Source term assessment with ASTEC and associated uncertainty analysis using SUNSET tool

    Energy Technology Data Exchange (ETDEWEB)

    Chevalier-Jabet, K., E-mail: karine.chevalier-jabet@irsn.fr; Cousin, F.; Cantrel, L.; Séropian, C.

    2014-06-01

    Several accidental scenarios have been simulated using the ASTEC integral IRSN-GRS code for a French 1300 MWe PWR, including several break sizes or locations, highlighting the effect of safety systems and of iodine chemistry in the reactor coolant system (RCS) and in the containment on iodine source term evaluations. Iodine chemistry in the RCS and in the containment is still subject to significant uncertainties and it is thus studied in on-going R and D programs. To assess the impact of uncertainties, ASTEC has been coupled to the IRSN uncertainty propagation and sensitivity analysis tool SUNSET. Focusing on a loss of feed-water of steam generator accident, ASTEC/SUNSET calculations have been performed to assess the effect of remaining uncertainties relative to iodine behaviour on the source term. Calculations show that the postulated lack of knowledge may impact the iodine source term in the environment by at least one decade, confirming the importance of the on-going R and D programs to improve the knowledge on iodine chemistry.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  2. Activities of the central authorities in 1990-1991 on the Chernobyl accident response and social defence of it's victims. Long-term programs on the Chernobyl accident response

    International Nuclear Information System (INIS)

    Chapter presents the information on the approved All-Union Programs to eliminate the long-term consequences of the CNPP accident and on problems linked with the social protection of the victims. Three stages of the activity to overcome the consequences of the accident are studied: 1-st period - 1986; 2-nd period - 1987-1989; third period (reconstruction) -1990-1995 and subsequent years. Attention is focused on the All-Union section of the State Union and Republic Program to protect the USSR population against the effects of the Chernobyl NPP in 1991-1995 and during the period up to 2000

  3. Flowsheets and source terms for radioactive waste projections

    International Nuclear Information System (INIS)

    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, UF6 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

  4. Common Calibration Source for Monitoring Long-term Ozone Trends

    Science.gov (United States)

    Kowalewski, Matthew

    2004-01-01

    Accurate long-term satellite measurements are crucial for monitoring the recovery of the ozone layer. The slow pace of the recovery and limited lifetimes of satellite monitoring instruments demands that datasets from multiple observation systems be combined to provide the long-term accuracy needed. A fundamental component of accurately monitoring long-term trends is the calibration of these various instruments. NASA s Radiometric Calibration and Development Facility at the Goddard Space Flight Center has provided resources to minimize calibration biases between multiple instruments through the use of a common calibration source and standardized procedures traceable to national standards. The Facility s 50 cm barium sulfate integrating sphere has been used as a common calibration source for both US and international satellite instruments, including the Total Ozone Mapping Spectrometer (TOMS), Solar Backscatter Ultraviolet 2 (SBUV/2) instruments, Shuttle SBUV (SSBUV), Ozone Mapping Instrument (OMI), Global Ozone Monitoring Experiment (GOME) (ESA), Scanning Imaging SpectroMeter for Atmospheric ChartographY (SCIAMACHY) (ESA), and others. We will discuss the advantages of using a common calibration source and its effects on long-term ozone data sets. In addition, sphere calibration results from various instruments will be presented to demonstrate the accuracy of the long-term characterization of the source itself.

  5. Maximal design basis accident of fusion neutron source DEMO-TIN

    Energy Technology Data Exchange (ETDEWEB)

    Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

  6. Maximal design basis accident of fusion neutron source DEMO-TIN

    Science.gov (United States)

    Kolbasov, B. N.

    2015-12-01

    When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission-fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

  7. Long term effects of Minks of the radiation factors from the Chernobyl accident

    International Nuclear Information System (INIS)

    The study of small radiation dose influence on human and animal reproductive functions becomes more and more topical after Chernobyl Nuclear Power Plant (ChNPP) accident. In the number of cases, animals that reside in continues internal, as well as external exposure zone, have pregnancy interruption in its early stages (up to 30 days). This, without any doubts testifies for reproductive process disorder as a whole (hypophysis-ovary-uterus system) and also, as its separate links. The important thing is that a break in any one of those links leads to pregnancy interruption. Hence, in order to determine any disorders in reproductive system functional state, profound and detailed morphofunctional study of the system links (accounting for radiation exposure factors) needs to be done. Because research in this field has just started, we were unable to find any material on this topic. There are, however, some references for morphofunctional changes of endocrine glands, hypophysis in particular and sex glands, refereed to small radiation doses

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

    International Nuclear Information System (INIS)

    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, UF6 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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/sub 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.

  10. 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)

  11. Use of open source software in estimating the effects of a severe accident on the Mark II containment

    International Nuclear Information System (INIS)

    Because the spectrum of scenarios of severe accident before which must verify the integrity of the containment can be very broad, it arises here a calculation methodology to estimate the structural response of the containment without incurring in high costs for commercial software licenses, or in times and calculation excessive requirements. The capabilities of computer programs with license of open source, OpenFOAM for CFD calculations and Salome-Meca for thermal and mechanical calculations were tested. The methodology begins of the venting of mass and energy that are postulated inside the container and the values of the thermal and mechanical fields are obtained through the walls. (Author)

  12. [Method of ecological risk assessment for risk pollutants under short-term and high dose exposure in water pollution accident].

    Science.gov (United States)

    Lei, Bing-Li; Sun, Yan-Feng; Liu, Qian; Yu, Zhi-Qiang; Zeng, Xiang-Ying

    2011-11-01

    In recent years, water pollution accidents resulting in acute aquatic ecological risk and security issues become a research focus. However, in our country, the surface water quality standards and drinking water health standards were used to determine the safety of waters or not in pollution incidents due to lacking safety effect threshold or risk value for protection of aquatic life. In foreign countries, although predicted no effect concentration (PNEC) or risk value (R) of pollutants were provided for protection of aquatic organisms, the PNECs or risk values were derived based on long-term exposure toxicity data NOECs (no observed effect concentrations) and lack of short-term exposure risk or threshold values. For the short-term and high dose exposure in pollution incident, ecological risk assessment methods were discussed according to the procedures of the conventional ecological risk assessment and the water quality criteria establishment of the U.S. EPA for the protection of aquatic organisms in short-term exposure, and had a case study. At the same time, we provide some suggestions for the establishment of ecological risk assessment system in water pollution incidents. PMID:22295619

  13. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  14. Short-term associations between outdoor air pollution and visits to accident and emergency departments in London for respiratory complaints.

    Science.gov (United States)

    Atkinson, R W; Anderson, H R; Strachan, D P; Bland, J M; Bremner, S A; Ponce de Leon, A

    1999-02-01

    Many epidemiological studies have shown positive short-term associations between health and current levels of outdoor air pollution. The aim of this study was to investigate the association between air pollution and the number of visits to accident and emergency (A&E) departments in London for respiratory complaints. A&E visits include the less severe cases of acute respiratory disease and are unrestricted by bed availability. Daily counts of visits to 12 London A&E departments for asthma, other respiratory complaints, and both combined for a number of age groups were constructed from manual registers of visits for the period 1992-1994. A Poisson regression allowing for seasonal patterns, meteorological conditions and influenza epidemics was used to assess the associations between the number of visits and six pollutants: nitrogen dioxide, ozone, sulphur dioxide, carbon monoxide, and particles measured as black smoke (BS) and particles with a median aerodynamic diameter of <10 microm (PM10). After making an allowance for the multiplicity of tests, there remained strong associations between visits for all respiratory complaints and increases in SO2: a 2.8% (95% confidence interval (CI) 0.7-4.9) increase in the number of visits for a 18 microg x (-3) increase (10th-90th percentile range) and a 3.0% (95% CI 0.8-5.2) increase for a 31 microg x m(-3) increase in PM10. There were also significant associations between visits for asthma and SO2, NO2 and PM10. No significant associations between O3 and any of the respiratory complaints investigated were found. Because of the strong correlation between pollutants, it was difficult to identify a single pollutant responsible for the associations found in the analyses. This study suggests that the levels of air pollution currently experienced in London are linked to short-term increases in the number of people visiting accident and emergency departments with respiratory complaints. PMID:10065665

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

    International Nuclear Information System (INIS)

    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

  16. Radiological Source Terms for Tank Farms Safety Analysis

    International Nuclear Information System (INIS)

    This document provides Unit Liter Dose factors, atmospheric dispersion coefficients, breathing rates and instructions for using and customizing these factors for use in calculating radiological doses for accident analyses in the Hanford Tank Farms

  17. Radiological Source Terms for Tank Farms Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    COWLEY, W.L.

    2000-06-27

    This document provides Unit Liter Dose factors, atmospheric dispersion coefficients, breathing rates and instructions for using and customizing these factors for use in calculating radiological doses for accident analyses in the Hanford Tank Farms.

  18. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    For the assessment of the safety of nuclear power plants it is of great importance the analyses of severe accidents since they allow to estimate the possible failure models of the containment, and also permit knowing the magnitude and composition of the radioactive material that would be released to the environment in case of an accident upon population and the environment. This paper presents in general terms the basic principles for conducting the analysis of severe accidents, the fundamental sources in the generation of radionuclides and aerosols, the transportation and deposition processes, and also makes reference to de main codes used in the modulation of severe accidents. The final part of the paper contents information on how severe accidents are dialed with the regulatory point view in different countries

  19. National preparedness guide for exiting the emergency phase subsequent to a nuclear accident causing moderate, short-term release on French soil - working document, version 0, May 2010

    International Nuclear Information System (INIS)

    This National Guide provides basic explanations and methods to assist in drawing up a local plan for the emergency phase way-out, subsequent to a nuclear accident of moderate magnitude causing short-term (under 24 hours) radioactive release, which could possibly occur in France. The accident situations considered in this Guide have little likelihood of arising and are representative of environmental contamination accidents that might occur at French nuclear facilities covered by a special intervention plan (plan particulier d'intervention, PPI). Such accidents may cause environmental contamination warranting action for post-accident impact management within a range of ten to fifty kilometres from the accident site. To provide some perspective, the accidents considered here would be classified Levels 3, 4 or 5 (incidents or accidents causing release into the environment) on the INES scale customarily used to help the public and media to immediately understand the severity of an incident or accident in the nuclear field. This Guide was drawn up subsequently to the work carried out by the Steering Committee on Post- Accident Phase Management in the Event of a Nuclear Accident or Radiological Emergency Situation (CODIRPA), instituted by the French Nuclear Safety Authority (ASN) in June 2005, and in charge of setting out the basic principles underlying the management of nuclear post-accident situations. This version of the Guide shall be updated on the basis of the operating experience feedback received on its use. The Guide is a planning tool, intended for the Prefectures of department where a basic nuclear facility PPI has been instituted. Its purpose is to enable Prefects to plan and effectively conduct preparedness measures at the local level with the aim of winding down the emergency phase, actively involving all of the relevant actors for this purpose. The exit period from the emergency phase is defined as extending approximately one week from the end of the

  20. Evaluation of source term induced by beam loss in the superconducting linear accelerator at RAON

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Jin; Kim, Su Na; Nam, Shin Woo; Chung, Yon Sei [Rare Isotope Science Project, Institute for Basic Science, Daejeon (Korea, Republic of)

    2014-11-15

    As a new world-class heavy ion accelerator, RAON is able to accelerate heavy ions from proton to uranium with the energy up to -400 MeV/u and produce rare isotopes. These high purity, high intensity, and high energy beams generate the various secondary radiation which will impact on the shielding aspects of the main linear accelerator tunnels. In the main tunnel the secondary neutrons are produced by uniform beam-loss or accident criteria. In this paper evaluations of several source terms induced by beam-loss will be discussed along with the physics model of the Monte Carlo simulation codes. The beam-loss criteria were tested for the evaluation of source term for the main beam line tunnel of the RAON accelerator. It was found that the amount of the secondary neutrons depends on the incident angle of projectile on the beam pipe and the mass and energy of projectile. The influence of selected physics models and libraries of MCNPX and PHITS has been examined. The secondary neutrons were produced most in the CEM and LAQGSM model.

  1. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

    In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

  2. Simulation of the Syrian miniature neutron source reactor for training operators on the analysis of its anticipated operational accidents

    International Nuclear Information System (INIS)

    For the purpose of training operators and other educational aspects, a mathematical model capable of assessing potential accidents and safety implications of the research Miniature Neutron Source Reactor (MNSR) has been developed. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant temperature and xenon, and thermal hydraulics. Natural convection and point kinetics including the prompt jump and complete mixing approximations were employed. Peak power, reactivity core load, core outlet temperature, and other variables are predicted during self-limiting power excursions. Compared to related references, close results have been obtained. The simulating model proves to be a useful tool to train operators and students to assess qualitatively the transient behaviour of the MNSR as a result of sudden reactivity insertion in the core. In addition, the model was utilized to verify some of the design basis accidents already presented in both the Safety Analysis Report (SAR) and the Commissioning Report (CR) of the reactor. Furthermore, the dynamic model generates other core variables that are of interest to update the SAR on one side, and confirms others measured and reported in the CR

  3. Simulation of the Syrian Miniature Neutron Source Reactor for training operators on the analysis of its anticipated operational accidents

    International Nuclear Information System (INIS)

    For the purpose of training operators and other educational aspects, a mathematical model capable of assessing potential accidents and safety implications of the research Miniature Neutron Source Reactor (MNSR) has been developed. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feed-backs due to coolant temperature and xenon, and thermal hydraulics. Natural convection and point kinetics including the prompt jump and complete mixing approximations were employed. Peak power, reactivity core load, core outlet temperature, and other variables are predicted during self-limiting power excursions. Compared to related references, close results have been obtained. The simulating model proves to be a useful tool to train operators and students to assess qualitatively the transient behaviour of the MNSR as a result of sudden reactivity insertion in the core. In addition, the model was utilized to verify some of the design basis accidents already presented in both the Safety Analysis Report (SAR) and the Commissioning Report (CR) of the reactor, as can be seen in Table 1. Furthermore, the dynamic model generates other core variables that are of interest to update the SAR on one side, and confirms others measured and reported in the CR. (author)

  4. Short-term associations between outdoor air pollution and visits to accident and emergency departments in London for respiratory complaints

    Energy Technology Data Exchange (ETDEWEB)

    Atkinson, R.W.; Anderson, H.R.; Strachan, D.P.; Bland, J.M.; Bremner, S.A. [St. George`s Hospital Medical School, Dept. of Public Health Sciences, London (United Kingdom); Ponce de Leon, A. [IME/UERJ Rua Sao Francisco Xavier, Dept. de Estatistica, Maracana Rio de Janeiro (Brazil)

    1999-02-01

    Many epidemiological studies have shown positive short-term associations between health and current levels of outdoor air pollution. The aim of this study was to investigate the association between air pollution and the number of visits to accident and emergency (A and E) departments in London for respiratory complaints. A and E visits include the less severe cases of acute respiratory disease and are unrestricted by bed availability. Daily counts of visits to 12 London A and E departments for asthma, other respiratory complaints, and both combined for a number of age groups were constructed from manual registers of visits for the period 1992-1994. A poison regression allowing for seasonal patterns meteorological conditions and influenza epidemics was used to assess the associations between the number of visits and six pollutants: nitrogen dioxide, ozone, sulphur dioxide, carbon monoxide, and particles measured as black smoke (BS) and particles with a median aerodynamic diameter of <10 {mu}m (PM10). After making an allowance for the multiplicity of tests, there remained strong associations between visits for all respiratory complaints and increases in SO{sub 2}: a 2.8% (95% confidence interval (CI) 0.7-4.9) increase in the number of visits for a 18 {mu}g{sup .}m{sup -3} increase (10th-90th percentile range) and a 3.0% (95% Cl 0.8-5.2) increase for a 31 {mu}g{sup .}m{sup -3} increase in PM10. There were also significant associations between visits for asthma and SO{sub 2}, NO{sub 2} and PM10. No significant associations between O{sub 3} and any of the respiratory complaints investigated were found. Because of the strong correlation between pollutants, it was difficult to identify a single pollutant responsible for the associations found in the analyses. This study suggests that the levels of air pollution currently experienced in London are linked to short-term increases in the number of people visiting accident and emergency departments with respiratory complaints

  5. Short-term associations between outdoor air pollution and visits to accident and emergency departments in London for respiratory complaints

    Energy Technology Data Exchange (ETDEWEB)

    Atkinson, R.W.; Anderson, H.R.; Strachan, D.P.; Bland, J.M.; Bremner, S.A. [St. George' s Hospital Medical School, Dept. of Public Health Sciences, London (United Kingdom); Ponce de Loen, A. [IME/UERJ Rua Sao Francisco Xavier, Dept. de Estatistica, Maracana Rio de Janeiro , RJ (Brazil)

    1999-07-01

    Many epidemiological studies have shown positive short-term associations between health and current levels of outdoor air pollution. The aim of this study was to investigate the association between air pollution and the number of visits to accident and emergency (A and E) departments in London for respiratory complaints. A and E visits include the less severe cases of acute respiratory disease and are unrestricted by bed availability. Daily counts of visits to 12 London A and E departments for asthma, other respiratory complaints, and both combined for a number of age groups were constructed from manual registers of visits for the period 1992-1994. A Poisson regression allowing for seasonal patterns, meteorological conditions and influenza epidemics was used to assess the associations between the number of visits and six pollutants: nitrogen dioxide, ozone, sulphur dioxide, carbon monoxide, and particles measured as black smoke (BS) and particles with a median aerodynamic diameter of <10 {mu}m (PM10). After making an allowance for the multiplicity of tests, there remained strong associations between visits for all respiratory complaints and increases in SO{sub 2}: a 2.8% (95% confidence interval (CI) 0.7-4.9) increase in the number of visits for a 18 {mu}g{sup .}m{sup -3} increase (10th-90th percentile range) and a 3.0% (95% CI 0.8-5.2) increase for a 31 {mu}g{sup .}m{sup -3} increase in PM10. There were also significant associations between visits for asthma and SO{sub 2}, NO{sub 2} and PM10. No significant associations between O{sub 3} and any of the respiratory complaints investigated were found. Because of the strong correlation between pollutants, it was difficult to identify a single pollutant responsible for the associations found in the analyses. This study suggests that the levels of air pollution currently experineced in London are linked to short-term increases in the number of people visiting accident and emergency departments with respiratory

  6. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

    International Nuclear Information System (INIS)

    The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

  7. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  8. Inverse modelling-based reconstruction of the Chernobyl source term available for long-range transport

    Directory of Open Access Journals (Sweden)

    X. Davoine

    2007-01-01

    Full Text Available The reconstruction of the Chernobyl accident source term has been previously carried out using core inventories, but also back and forth confrontations between model simulations and activity concentration or deposited activity measurements. The approach presented in this paper is based on inverse modelling techniques. It relies both on the activity concentration measurements and on the adjoint of a chemistry-transport model. The location of the release is assumed to be known, and one is looking for a source term available for long-range transport that depends both on time and altitude. The method relies on the maximum entropy on the mean principle and exploits source positivity. The inversion results are mainly sensitive to two tuning parameters, a mass scale and the scale of the prior errors in the inversion. To overcome this hardship, we resort to the statistical L-curve method to estimate balanced values for these two parameters. Once this is done, many of the retrieved features of the source are robust within a reasonable range of parameter values. Our results favour the acknowledged three-step scenario, with a strong initial release (26 to 27 April, followed by a weak emission period of four days (28 April–1 May and again a release, longer but less intense than the initial one (2 May–6 May. The retrieved quantities of iodine-131, caesium-134 and caesium-137 that have been released are in good agreement with the latest reported estimations. Yet, a stronger apportionment of the total released activity is ascribed to the first period and less to the third one. Finer chronological details are obtained, such as a sequence of eruptive episodes in the first two days, likely related to the modulation of the boundary layer diurnal cycle. In addition, the first two-day release surges are found to have effectively reached an altitude up to the top of the domain (5000 m.

  9. 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

  10. Inverse estimation of source parameters of oceanic radioactivity dispersion models associated with the Fukushima accident

    Directory of Open Access Journals (Sweden)

    Y. Miyazawa

    2013-04-01

    Full Text Available With combined use of the ocean–atmosphere simulation models and field observation data, we evaluate the parameters associated with the total caesium-137 amounts of the direct release into the ocean and atmospheric deposition over the western North Pacific caused by the accident of Fukushima Daiichi nuclear power plant (FNPP that occurred in March 2011. The Green's function approach is adopted for the estimation of two parameters determining the total emission amounts for the period from 12 March to 6 May 2011. It is confirmed that the validity of the estimation depends on the simulation skill near FNPP. The total amount of the direct release is estimated as 5.5–5.9 × 1015 Bq, while that of the atmospheric deposition is estimated as 5.5–9.7 × 1015 Bq, which indicates broader range of the estimate than that of the direct release owing to uncertainty of the dispersion widely spread over the western North Pacific.

  11. Inverse estimation of source parameters of oceanic radioactivity dispersion models associated with the Fukushima accident

    Directory of Open Access Journals (Sweden)

    Y. Miyazawa

    2012-10-01

    Full Text Available With combined use of the ocean-atmosphere simulation models and field observation data, we evaluate the parameters associated with the total caesium-137 amounts of the direct release into the ocean and atmospheric deposition over the Western North Pacific caused by the accident of Fukushima Daiichi nuclear power plant (FNPP that occurred in March 2011. The Green's function approach is adopted for the estimation of two parameters determining the total emission amounts for the period from 12 March to 6 May 2011. It is confirmed that the validity of the estimation depends on the simulation skill near FNPP. The total amount of the direct release is estimated as 5.5–5.9 × 1015 Bq, while that of the atmospheric deposition is estimated as 5.5–9.7 × 1015 Bq, which indicates broader range of the estimate than that of the direct release owing to uncertainty of the dispersion widely spread over the Western North Pacific.

  12. Core structure heat-up and material relocation in a BWR short-term station blackout accident

    International Nuclear Information System (INIS)

    This paper presents an analytical and numerical analysis which evaluates the core-structure heat-up and subsequent relocation of molten core materials during a NWR short-term station blackout accident with ADS. A simplified one-dimensional approach coupled with bounding arguments is first presented to establish an estimate of the temperature differences within a BWR assembly at the point when structural material first begins to melt. This analysis leads to the conclusions that the control blade will be the first structure to melt and that at this point in time, overall temperature differences across the canister-blade region will not be more than 200 K. Next, a three-dimensional heat-transfer model of the canister-blade region within the core is presented that uses a diffusion approximation for the radiation heat transfer. This is compared to the one-dimensional analysis to establish its compatibility. Finally, the extension of the three-dimensional model to include melt relocation using a porous media type approximation is described. The results of this analysis suggest that under these conditions significant amounts of material will relocate to the core plate region and refreeze, potentially forming a significant blockage. The results also indicate that a large amount of lateral spreading of the melted blade and canister material into the fuel rod regions will occur during the melt progression process. 22 refs., 18 figs., 1 tab

  13. The accident of Chernobyl

    International Nuclear Information System (INIS)

    RBMK reactors (reactor control, protection systems, containment) and the nuclear power plant of Chernobyl are first presented. The scenario of the accident is given with a detailed chronology. The actions and consequences on the site are reviewed. This report then give the results of the source term estimation (fision product release, core inventory, trajectories, meteorological data...), the radioactivity measurements obtained in France. Health consequences for the French population are evoked. The medical consequences for the population who have received a high level of doses are reviewed

  14. Trace Metal Source Terms in Carbon Sequestration Environments

    Energy Technology Data Exchange (ETDEWEB)

    Karamalidis, Athanasios K; Torres, Sharon G; Hakala, J Alexandra; Shao, Hongbo; Cantrell, Kirk J; Carroll, Susan

    2012-02-05

    Carbon dioxide sequestration in deep saline and depleted oil geologic formations is feasible and promising, however, possible CO₂ or CO₂-saturated brine leakage to overlying aquifers may pose environmental and health impacts. The purpose of this study was to experimentally define trace metal source terms from the reaction of supercritical CO₂, storage reservoir brines, reservoir and cap rocks. Storage reservoir source terms for trace metals are needed to evaluate the impact of brines leaking into overlying drinking water aquifers. The trace metal release was measured from sandstones, shales, carbonates, evaporites, basalts and cements from the Frio, In Salah, Illinois Basin – Decatur, Lower Tuscaloosa, Weyburn-Midale, Bass Islands and Grand Ronde carbon sequestration geologic formations. Trace metal dissolution is tracked by measuring solution concentrations over time under conditions (e.g. pressures, temperatures, and initial brine compositions) specific to the sequestration projects. Existing metrics for Maximum Contaminant Levels (MCLs) for drinking water as defined by the U.S. Environmental Protection Agency (U.S. EPA) were used to categorize the relative significance of metal concentration changes in storage environments due to the presence of CO₂. Results indicate that Cr and Pb released from sandstone reservoir and shale cap rock exceed the MCLs by an order of magnitude while Cd and Cu were at or below drinking water thresholds. In carbonate reservoirs As exceeds the MCLs by an order of magnitude, while Cd, Cu, and Pb were at or below drinking water standards. Results from this study can be used as a reasonable estimate of the reservoir and caprock source term to further evaluate the impact of leakage on groundwater quality.

  15. TREAT light water reactor source term experiments program

    International Nuclear Information System (INIS)

    Four experiments are being conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients are used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments are integral in nature and are aimed at the physicochemical characterization, near the point of origin, of the biologically important volatile fission products released early in such accidents. Detailed program objectives are discussed, a test matrix is presented, and the test apparatus is described. Pretest analysis and preliminary results are reported for the first test

  16. Quasilinear equations with source terms on Carnot groups

    CERN Document Server

    Phuc, Nguyen Cong

    2012-01-01

    In this paper we give necessary and sufficient conditions for the existence of solutions to quasilinear equations of Lane--Emden type with measure data on a Carnot group $\\mathbb G$ of arbitrary step. The quasilinear part involves operators of the $p$-Laplacian type $\\Delta_{\\mathbb G,\\,p}\\,$, $1terms of nonlinear potentials of Th. Wolff's type. As a consequence, we characterize completely removable singularities, and prove a Liouville type theorem for supersolutions of quasilinear equations with source terms which has been known only for equations involving the sub-Laplacian ($p=2$) on the Heisenberg group.

  17. Source term evaluation for accident transients in the experimental fusion facility ITER

    Energy Technology Data Exchange (ETDEWEB)

    Virot, F.; Barrachin, M.; Cousin, F. [IRSN, BP3-13115, Saint Paul lez Durance (France)

    2015-03-15

    We have studied the transport and chemical speciation of radio-toxic and toxic species for an event of water ingress in the vacuum vessel of experimental fusion facility ITER with the ASTEC code. In particular our evaluation takes into account an assessed thermodynamic data for the beryllium gaseous species. This study shows that deposited beryllium dusts of atomic Be and Be(OH){sub 2} are formed. It also shows that Be(OT){sub 2} could exist in some conditions in the drain tank. (authors)

  18. 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

    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...

  19. Nuclear accidents and epidemiology

    International Nuclear Information System (INIS)

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  20. Long-term follow-up of the residents of the Three Mile Island accident area: 1979-1998.

    OpenAIRE

    Talbott, Evelyn O.; Youk, Ada O.; McHugh-Pemu, Kathleen P; Zborowski, Jeanne V

    2003-01-01

    The Three Mile Island (TMI) nuclear power plant accident (1979) prompted the Pennsylvania Department of Health to initiate a cohort mortality study in the TMI accident area. This study is significant because of the long follow-up (1979-1998), large cohort size (32,135), and evidence from earlier reports indicating increased cancer risks. Standardized mortality ratios (SMRs) were calculated to assess the mortality experience of the cohort compared with a local population. Relative risk (RR) re...

  1. Use of source term uncoupled in radionuclide migration equations

    Energy Technology Data Exchange (ETDEWEB)

    Silveira, Claudia Siqueira da; Lima, Zelmo Rodrigues de; Alvim, Antonio Carlos Marques [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE)]. E-mails: siqueira_claudia@yahoo.com.br; zelmolima@yahoo.com.br; aalvim@gmail.com

    2008-07-01

    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)

  2. Accident with radioactive substances in laboratory. An exercise during the education of persons in radiation protection, who are working with open radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    Stolze, B.

    2003-07-01

    In spite of carefulness it is possible,that contamination occur by handling unscaled radioactive sources or in case of an accident. It is demonstrated in an exercise managing an accident with unscaled radioactive sources. The persons, who are educated in radiation protection for handling unsealed radioactive sources, must have knowledge of theoretical regulations of the radiation protection law and of the limits in radiation protection. Also they have to know the handling to reduce possible contamination. They have to be able to calculate the dose of skin contamination. In my lecture I give some information on regulations of accidents with radioactive sources in Germany and a scenario of an accident and I explain, what is to do to manage this event. A person opened an ampoule. The activity splashed and contaminated the person's hand, arm and face. Also in the room there was a contamination. The desk and the floor were contaminated. There were 50 MBq P-32 as NaH{sub 2} P''32O{sub 3} in water solution, I give a report on practices in our courses, which the participants have to do. The radiological experts have to decontaminate the skin and they have to calculate the skin-dose and to give the information to the authorities. (Author) 4 refs.

  3. Multi-stage ranking of emergency technology alternatives for water source pollution accidents using a fuzzy group decision making tool.

    Science.gov (United States)

    Qu, Jianhua; Meng, Xianlin; You, Hong

    2016-06-01

    Due to the increasing number of unexpected water source pollution events, selection of the most appropriate disposal technology for a specific pollution scenario is of crucial importance to the security of urban water supplies. However, the formulation of the optimum option is considerably difficult owing to the substantial uncertainty of such accidents. In this research, a multi-stage technical screening and evaluation tool is proposed to determine the optimal technique scheme, considering the areas of pollutant elimination both in drinking water sources and water treatment plants. In stage 1, a CBR-based group decision tool was developed to screen available technologies for different scenarios. Then, the threat degree caused by the pollution was estimated in stage 2 using a threat evaluation system and was partitioned into four levels. For each threat level, a corresponding set of technique evaluation criteria weights was obtained using Group-G1. To identify the optimization alternatives corresponding to the different threat levels, an extension of TOPSIS, a multi-criteria interval-valued trapezoidal fuzzy decision making technique containing the four arrays of criteria weights, to a group decision environment was investigated in stage 3. The effectiveness of the developed tool was elaborated by two actual thallium-contaminated scenarios associated with different threat levels. PMID:26897576

  4. Multi-stage ranking of emergency technology alternatives for water source pollution accidents using a fuzzy group decision making tool.

    Science.gov (United States)

    Qu, Jianhua; Meng, Xianlin; You, Hong

    2016-06-01

    Due to the increasing number of unexpected water source pollution events, selection of the most appropriate disposal technology for a specific pollution scenario is of crucial importance to the security of urban water supplies. However, the formulation of the optimum option is considerably difficult owing to the substantial uncertainty of such accidents. In this research, a multi-stage technical screening and evaluation tool is proposed to determine the optimal technique scheme, considering the areas of pollutant elimination both in drinking water sources and water treatment plants. In stage 1, a CBR-based group decision tool was developed to screen available technologies for different scenarios. Then, the threat degree caused by the pollution was estimated in stage 2 using a threat evaluation system and was partitioned into four levels. For each threat level, a corresponding set of technique evaluation criteria weights was obtained using Group-G1. To identify the optimization alternatives corresponding to the different threat levels, an extension of TOPSIS, a multi-criteria interval-valued trapezoidal fuzzy decision making technique containing the four arrays of criteria weights, to a group decision environment was investigated in stage 3. The effectiveness of the developed tool was elaborated by two actual thallium-contaminated scenarios associated with different threat levels.

  5. Natural Radioactivity Source Term Based on Remote Sensing Data

    International Nuclear Information System (INIS)

    The paper describes the basic principles for applying satellite remote sensing technology to the investigation of natural radioactivity. The relationship between areas of natural background anomalies and geological characteristics is analysed systematically. The supervised classification method and spectral angle mapping are used for the extraction of remote sensing information. Geological features with elevated levels of gamma radiation can be identified on small scale maps. On-site inspections have been launched. The relationship between natural radiation level and radiation source term is becoming clearer. The study provides exact locations and targets for protection and control in areas with elevated levels of gamma radiation. The project has the potential for expanding the range of services in environmental geochemistry and remote sensing geology. It opens up a new approach for conducting research on natural radioactivity. (author)

  6. Tank waste source term inventory validation. Volume 1. Letter report

    Energy Technology Data Exchange (ETDEWEB)

    Brevick, C.H.; Gaddis, L.A.; Johnson, E.D.

    1995-04-28

    The sample data for selection of 11 radionuclides and 24 chemical analytes were extracted from six separate sample data sets, were arranged in a tabular format and were plotted on scatter plots for all of the 149 single-shell tanks, the 24 double-shell tanks and the four aging waste tanks. The solid and liquid sample data was placed in separate tables and plots. The sample data and plots were compiled from the following data sets: characterization raw sample data, recent core samples, D. Braun data base, Wastren (Van Vleet) data base, TRAC and HTCE inventories. This document is Volume I of the Letter Report entitled Tank Waste Source Term Inventory Validation.

  7. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

  8. 5.0. Depletion, activation, and spent fuel source terms

    Energy Technology Data Exchange (ETDEWEB)

    Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    SCALE’s general depletion, activation, and spent fuel source terms analysis capabilities are enabled through a family of modules related to the main ORIGEN depletion/irradiation/decay solver. The nuclide tracking in ORIGEN is based on the principle of explicitly modeling all available nuclides and transitions in the current fundamental nuclear data for decay and neutron-induced transmutation and relies on fundamental cross section and decay data in ENDF/B VII. Cross section data for materials and reaction processes not available in ENDF/B-VII are obtained from the JEFF-3.0/A special purpose European activation library containing 774 materials and 23 reaction channels with 12,617 neutron-induced reactions below 20 MeV. Resonance cross section corrections in the resolved and unresolved range are performed using a continuous-energy treatment by data modules in SCALE. All nuclear decay data, fission product yields, and gamma-ray emission data are developed from ENDF/B-VII.1 evaluations. Decay data include all ground and metastable state nuclides with half-lives greater than 1 millisecond. Using these data sources, ORIGEN currently tracks 174 actinides, 1149 fission products, and 974 activation products. The purpose of this chapter is to describe the stand-alone capabilities and underlying methodology of ORIGEN—as opposed to the integrated depletion capability it provides in all coupled neutron transport/depletion sequences in SCALE, as described in other chapters.

  9. Mathematical models for steam generator accident simulation

    International Nuclear Information System (INIS)

    In this contribution, the numerical methods used in the DeBeNe-LMFBR development for the analysis of the hydrodynamic and mechanical consequences of steam generator accidents are presented. At first the definition of the source term, i.e. the water leak rate which has to be assumed in the design basis accident as well as the thermochemistry of the sodium/water-reaction is discussed. Then the computer-codes presently used to describe the hydrodynamic and mechanical consequences of steam generator accidents on the basis of the above mentioned source term are presented. These comprise the code-system SAPHYR and the code PTANER and PISCES. Furthermore, developments which are planned or already under way for future use, such as the BEREPOT-code, are presented. (author)

  10. Accident Involving the Melting of a 137Cs Source at a Steel Works in Mexico

    International Nuclear Information System (INIS)

    On 20 June 2008, the National Nuclear Safety and Safeguards Commission (CNSNS) was notified by the firm Mexico Steel Tubes PLC (TAMSA), based in the state of Veracruz, of the presumed radioactive contamination of steelworks powders from its smelting process. This incident was detected because TAMSA produces casting powders that are sold to the firm National Zinc in the state of Nuevo Leon. National Zinc received a shipment of these steelworks powders and detected the presence of radioactive material in its radiation portals, for which reason it returned the shipment. TAMSA contracted a firm to monitor the shipment and the presence of radioactive material was detected, for which reason the CNSNS was notified. The CNSNS made various inspections to determine the origin of the contamination and found that a 137Cs source had inadvertently been melted in TAMSA's facilities. Consequently, steelworks powders and subproducts of the firm National Zinc were produced weighing around 2000 tonnes with concentrations of up to 544 130 Bq/kg. Whole body counts were performed on a total of 130 persons involved in the incident but no internal contamination was found. In addition, samples were taken from environmental strata in and around the TAMSA and National Zinc facilities but no 137Cs contamination was found. It is estimated that the source which was melted was approximately 185 GBq (5 Ci). Currently, the CNSNS is discussing, together with the firms, the strategy for managing, conditioning and storing the contaminated powders, since we do not have a final disposal site for radioactive waste in Mexico. (author)

  11. Methodology and tools for source term assessment in case of emergency - astrid project (EC 5th framework programme)

    International Nuclear Information System (INIS)

    Full text: Following the reactor accident in Three Mile Island most western European countries with nuclear power reactors started a huge safety upgrade work to improve not only reactor safety but also the introduction of mitigating systems that would minimize consequences of a severe reactor accident. Further to this also emergency preparedness arrangements were made; alarm criteria, on-site and off-site emergency structures and plans to protect population and environment. Following also the Chernobyl accident in 1986, many countries started the development of dispersion codes to calculate and better predict the consequences of a radioactive release. However, follow-ups of the Chernobyl accident in the mid 90ties revealed the need for an earlier start of assessing the actual severity of an accident to efficiently succeed in emergency response actions. By looking into the NPP, at the state of f fission product barriers and critical safety systems, the magnitude of a potential radioactive release could be predicted in a timely manner to allow emergency response to be executed even before the occurrence of a release. This is the perspective in which the development of ASTRID methodology and tool should be regarded. By focussing more an similarities than differences the ASTRID methodology aims at a solution that could be acceptable for several reactor types as well as reactor containments, different stages of technical designs, but also for use at on-site as well as off-site emergency centres. So, the scope of the ASTRID methodology outlines the common FP barriers and interrelated critical safety functions, that is, plant process parameters of importance to address, for such an assessment of the likely future state of the failed NPP and the resulting source term. The methodology maps out relevant process parameters and indicators, what and how to calculate and a structured way to summarize and conclude on potential source term and likely time projections. In a way

  12. A knowledge based severe accident handbook for PWR

    International Nuclear Information System (INIS)

    During the last decade the level of knowledge about severe accident phenomena has increased dramatically. The improved understanding has been achieved by extensive research but also from feed-back of experience from actual incidents/accidents such as Three Mile Island and Chernobyl. In Sweden, mitigating measures such as filtered venting and external water source were implemented at all nuclear power plants by 1988. In parallel the Emergency Operating Procedures (at Ringhals called Emergency Response Guidelines, ERG, and Beyond ERG, BERG) were developed to include these new features. However, the accident management system has since then been further improved and one important aspect is the long-term accident management. The new information obtained has been one of the basis for a new knowledge based handbook to support the unit leader and the Technical Support Center. The handbook contains information concerning specific issues in the BERG and advice how the organization can manage a long-term severe accident situation

  13. Source term research for ship reactor anticipated operational events

    International Nuclear Information System (INIS)

    According to the basic hypothesis of anticipated operational events, grounding on the special characters of a ship reactor, the equilibrium vapor specific activity and the cabin activity were calculated using NSRC code for the main loop and the secondary loop. The calculation results show that the computational mode of NSRC code is correct, and the NSRC code can be used to calculate radioactive effect of a ship reactor in anticipated operational events of design basis accidents. The calculation results can provide support to the safe operation of a ship nuclear power device. (authors)

  14. SOURCE TERM TARGETED THRUST FY 2005 NEW START PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    NA

    2005-10-05

    While a significant amount of work has been devoted to developing thermodynamic data. describing the sorption of radionuclides to iron oxides and other geomedia, little data exist to describe the interaction of key radionuclides found in high-level radioactive waste with the uranium surfaces expected in corroded spent nuclear fuel (SNF) waste packages. Recent work indicates that actinide adsorption to the U(VI) solids expected in the engineered barrier system may play a key role in the reduction of dissolved concentrations of radionuclides such as Np(V). However, little is known about the mechanism(s) of adsorption, nor are the thermodynamic data available to represent the phenomenon in predictive modeling codes. Unfortunately, this situation makes it difficult to consider actinide adsorption to the U(VI) silicates in either geochemical or performance assessment (PA) predictions. The primary goal in the Source Term Targeted Thrust area is to ''study processes that control radionuclide release from the waste form''. Knowledge of adsorption of actinides to U(VI) silicate solids its and parameterization in geochemical models will be an important step towards this goal.

  15. ICRP Publication 111 - Application of the Commission's recommendations to the protection of people living in long-term contaminated areas after a nuclear accident or a radiation emergency.

    Science.gov (United States)

    Lochard, J; Bogdevitch, I; Gallego, E; Hedemann-Jensen, P; McEwan, A; Nisbet, A; Oudiz, A; Oudiz, T; Strand, P; Janssens, A; Lazo, T; Carr, Z; Sugier, A; Burns, P; Carboneras, P; Cool, D; Cooper, J; Kai, M; Lecomte, J-F; Liu, H; Massera, G; McGarry, A; Mrabit, K; Mrabit, M; Sjöblom, K-L; Tsela, A; Weiss, W

    2009-06-01

    In this report, the Commission provides guidance for the protection of people living in long-term contaminated areas resulting from either a nuclear accident or a radiation emergency. The report considers the effects of such events on the affected population. This includes the pathways of human exposure, the types of exposed populations, and the characteristics of exposures. Although the focus is on radiation protection considerations, the report also recognises the complexity of post-accident situations, which cannot be managed without addressing all the affected domains of daily life, i.e. environmental, health, economic, social, psychological, cultural, ethical, political, etc. The report explains how the 2007 Recommendations apply to this type of existing exposure situation, including consideration of the justification and optimisation of protection strategies, and the introduction and application of a reference level to drive the optimisation process. The report also considers practical aspects of the implementation of protection strategies, both by authorities and the affected population. It emphasises the effectiveness of directly involving the affected population and local professionals in the management of the situation, and the responsibility of authorities at both national and local levels to create the conditions and provide the means favouring the involvement and empowerment of the population. The role of radiation monitoring, health surveillance, and the management of contaminated foodstuffs and other commodities is described in this perspective. The Annex summarises past experience of longterm contaminated areas resulting from radiation emergencies and nuclear accidents, including radiological criteria followed in carrying out remediation measures. PMID:20472181

  16. NKS-R ExCoolSe mid-term report KTH severe accidents research relevant to the NKS-ExCoolSe project

    International Nuclear Information System (INIS)

    The present mid-term progress report is prepared on the recent results from the KTH severe accident research program relevant to the objective of the ExCoolSe project sponsored by the NKS-R program. The previous PRE-MELT-DEL project at KTH sponsored by NKS provided an extensive assessment on the remaining issues of severe accidents in general and suggested the key issues to be resolved such as coolability and steam explosion energetics in ex-vessel which became a backbone of the ExCoolSe project in NKS. The EXCOOLSE project has been integrated with, and leveraged on, parallel research program at KTH on severe accident phenomena the MSWI project which is funded by the APRI program, SKI in Sweden and HSK in Switzerland and produced more understanding of the key remaining issues. During last year, the critical assessment of the existing knowledge and current SAMG and designs of Nordic BWRs identified the research focus and initiated the new series of research activities toward the resolution of the key remaining issues specifically pertaining to the Nordic BWRs.(au)

  17. Techniques for long term conditioning and storage of radium sources

    International Nuclear Information System (INIS)

    The Horia Hulubei National Institute of Research and Development for Physics and Nuclear Engineering developed its own technology for conditioning the radium spent sealed radioactive sources. The laboratory dedicated to radiological characterization, identification of radium sources as well as the encapsulation of spent sealed radioactive sources was equipped with a local ventilation system, welding devices, tightness test devices as well as radiometric portable devices. Two types of capsules have been designed for conditioning of radium spent sealed radioactive sources. For these kinds of capsules different types of storage packaging were developed. Data on the radium inventory will be presented in the paper. The paper contains the description of the process of conditioning of spent sealed radioactive sources as well as the description of the capsules and packaging. The paper describes the equipment used for the conditioning of the radium spent sealed sources. (authors)

  18. Informational uncertainties of risk assessment about accidents of chemicals

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    An analysis system of informational uncertainties for accidental risk assessment of chemicals is introduced. Statistical test methods and fuzzy sets method can do the quantitative analysis of the input parameters. The uncertainties of the model can be used by quantitative compared method for the leakage accidents of chemicals. The estimation of the leaking time is important for discussing accidental source term. The uncertain analyses of the release accident for pipeline gas (CO) liquid chlorine and liquid propane gas (LPG) have been discussed.

  19. Source term for atmospheric dispersion in pipeline rupture; Termo fonte para dispersao atmosferica em ruptura de gasoduto

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Rubem da Cunha [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Engenharia. Dept. de Engenharia Quimica]. E-mail: rreis@eq.pucrs.br; Leal, Cesar Antonio [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Dept. de Engenharia Nuclear. Programa de Pos-graduacao em Engenharia Mecanica]. E-mail: leal@vortex.ufrgs.br

    2003-07-01

    The evaluation of consequences of industrial accidents requires the determination of physical effects in the several stages that compose a sequence of events that needs to be modeled. When a dangerous product is liberated accidentally, in order to the estimate of the possible number of victims in the scenario under consideration, one needs to evaluate the amount of released material, its physical state, and the amount that becomes airborne to form the source term for dispersion. Natural gas, transported in pipelines, is already crossing the state of Rio Grande do Sul, where, as well as in another areas of the country, it will be widely used. The need to overcome large distances between distribution and consuming units of natural gas implies in the use of high pressure conditions for transportation. The modeling of the behavior of the system in case of an accidents under this condition of quite severe pressure was done in a similar fashion used in a previous work where good results were obtained. The pipeline initial conditions are supercritical. In this paper, it is presented a discussion and results of the amount of natural gas that would be liberated in the form gas for further dispersion in the atmosphere, for the case of a pipeline initially in the pressure of 7 MPa and ambient temperature. (author)

  20. Long-term therapy for polymorphic mental disorders in liquidators of the consequences of the accident at the Chernobyl nuclear power plant

    Directory of Open Access Journals (Sweden)

    V. N. Krasnov

    2012-01-01

    Full Text Available The paper gives the results of a long-term comparative therapeutic study of a large cohort of more than 500 liquidators of the consequences of the accident at the Chernobyl nuclear power plant in 1986. The patients were followed up (and periodically treated at hospital 5 years or more, usually 10—15 years. The study confirmed mainly the cerebrovascular nature of disorders following the pattern seen in moderate psychoorganic syndrome. Therapy with cerebroprotective agents having vascular vegetotropic properties could yield certain therapeutic results and, to some extent, preserve social functioning capacity in these patients.

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

    International Nuclear Information System (INIS)

    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 CaCO3-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 CaCO3-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 CaCO3-saturated solution. In general, Tc is much less leachable (<10 wt.% of the available

  2. Hanford tank residual waste - contaminant source terms and release models

    International Nuclear Information System (INIS)

    Residual waste is expected to be left in 177 underground storage tanks after closure at the U.S. Department of Energy's Hanford Site in Washington State (USA). In the long term, the residual wastes 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 to 29.1 wt%) in some tanks (C-103, C-106, and S-112) and relatively low (<1.5 wt%) in other tanks (C-202 and C-203). Gibbsite is a common mineral in tanks with high Al concentrations, while non-crystalline U-Na-C-O-P±H phases are common in the U-rich residual wastes from tanks C-202 and C-203. Iron oxides/hydroxides have been identified in all residual waste samples studied to date. Contaminant release from the residual wastes was studied by conducting batch leach tests using distilled deionized water, a Ca(OH)2-saturated solution, or a CaCO3-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 CaCO3-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 CaCO3-saturated solution. In general, Tc is much less

  3. Accident progression event tree analysis for postulated severe accidents at N Reactor

    International Nuclear Information System (INIS)

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

  4. Justification of strategies for agricultural countermeasures in the long term after the Chernobyl accident based on a cost-benefit analysis

    International Nuclear Information System (INIS)

    In the long term after the Chernobyl accident the introduction of systems of countermeasures in agriculture must be based on the optimization principle. To implement this principle, a concept was used of evaluation of the effectiveness of countermeasures based on a cost-benefit analysis. Countermeasure options were developed separately for collective and private sectors of rural settlements. For each type of farming a range of countermeasures were defined and the optimal ones were identified. The effectiveness of countermeasures was estimated on the basis of integral criteria: cost of averted collective dose (1 man-Sv), overall costs needed for countermeasures introduction and time for fulfilling legal regulations. Based on the most effective countermeasures, optimal combinations (strategies) were developed. An assessment was given of the effectiveness of countermeasures aimed at reducing the radionuclide content in animal products from collective farms and lowering doses to rural residents affected by the Chernobyl accident, based on a comparative cost benefit analysis. A study into the dynamics of 1 man-Sv cost when applying different countermeasures in the collective and private sectors allowed an identification of the most optimal measures for various time periods after the accident. The situation in the private sector is more critical than in the collective one. This is demonstrated by higher costs of countermeasures and costs of potential averted doses in the course of their application, as well as difference in times of legal regulations fulfillment. To optimize costs of the rehabilitation of agricultural lands, the most optimal in terms of meeting the standards strategy was determined, which is an address application of countermeasures. (author)

  5. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs

  6. Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA))

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.

  7. [Long-term evacuation after the nuclear accident in Fukushima ~Different daily living under low-dose radioactive suffering~].

    Science.gov (United States)

    Ishikawa, Kazunobu

    2013-01-01

    One year has passed since the Great East Japan Earthquake and the Fukushima No. 1 nuclear power plant accident. Even currently, more than 150,000 evacuees in Fukushima Prefecture are forced to leave their home and to move throughout Japan. Because of the limited space of temporary housing and the weakening of personal ties in local communities, many families need to move and have separate lives. As a consequence, Fukushima has a serious shortage of caregivers for the elderly. There have been more than 1,300 disaster-related deaths due to shock and stress after long-distance drifts from town to town. Most of the victims were the elderly, who collapsed, caught pneumonia, suffered stroke and heart attack. Concerns about the safety of low-dose radiation exposure deprived the elderly of important contact with playing outside with their grandchildren in Fukushima. Fear of invisible radioactive contamination inactivated outdoor activities such as farming, dairy, fishing, gardening, hiking and wild-vegetable/mushroom hunting, although most of these activities have been traditionally supported by the wisdom of the elderly. Several recent questionnaire investigations revealed that older evacuees wish to go home even if the environment has significant contamination. In contrast, more than half of younger generation with small children have a different attitude. Nuclear accident brought serious social pains although it did not acutely hurt our bodies.

  8. Comparison of the Chernobyl and Fukushima nuclear accidents: A review of the environmental impacts

    Energy Technology Data Exchange (ETDEWEB)

    Steinhauser, Georg, E-mail: georg.steinhauser@colostate.edu; Brandl, Alexander; Johnson, Thomas E.

    2014-02-01

    The environmental impacts of the nuclear accidents of Chernobyl and Fukushima are compared. In almost every respect, the consequences of the Chernobyl accident clearly exceeded those of the Fukushima accident. In both accidents, most of the radioactivity released was due to volatile radionuclides (noble gases, iodine, cesium, tellurium). However, the amount of refractory elements (including actinides) emitted in the course of the Chernobyl accident was approximately four orders of magnitude higher than during the Fukushima accident. For Chernobyl, a total release of 5300 PBq (excluding noble gases) has been established as the most cited source term. For Fukushima, we estimated a total source term of 520 (340–800) PBq. In the course of the Fukushima accident, the majority of the radionuclides (more than 80%) was transported offshore and deposited in the Pacific Ocean. Monitoring campaigns after both accidents reveal that the environmental impact of the Chernobyl accident was much greater than of the Fukushima accident. Both the highly contaminated areas and the evacuated areas are smaller around Fukushima and the projected health effects in Japan are significantly lower than after the Chernobyl accident. This is mainly due to the fact that food safety campaigns and evacuations worked quickly and efficiently after the Fukushima accident. In contrast to Chernobyl, no fatalities due to acute radiation effects occurred in Fukushima. - Highlights: • The environmental effects of Chernobyl and Fukushima are compared. • Releases of radionuclides from Chernobyl exceeded Fukushima by an order of magnitude. • Chernobyl caused more severe radiation-related health effects. • Overall, Chernobyl was a much more severe nuclear accident than Fukushima. • Psychological effects are neglected but important consequences of nuclear accidents.

  9. Quantum correlations in nuclear mean field theory through source terms

    CERN Document Server

    Lee, S J

    1996-01-01

    Starting from full quantum field theory, various mean field approaches are derived systematically. With a full consideration of external source dependence, the stationary phase approximation of an action gives a nuclear mean field theory which includes quantum correlation effects (such as particle-hole or ladder diagram) in a simpler way than the Brueckner-Hartree-Fock approach. Implementing further approximation, the result can be reduced to Hartree-Fock or Hartree approximation. The role of the source dependence in a mean field theory is examined.

  10. The effect of gamma-ray transport on afterheat calculations for accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S.; Latkowski, J.F.; Sanz, J.

    2000-05-01

    Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed.

  11. Accounting treatment of corporate financing by short-term sources

    OpenAIRE

    Miljak, Toni; Mijatović, Eva; Gregorić, Marina

    2014-01-01

    The function of financing is one of the primary functions of each company. It consists of making three fundamental decisions: decisions about investing, financing decision and the decision on the distribution of dividends. Financing of companies differs by maturity, by the origin of the source of funding, according to the owner of the use of financial resources and in accordance with the different accounting treatment of business events according to International Accounting Standards and the ...

  12. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  13. Nuclear accidents

    International Nuclear Information System (INIS)

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  14. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

    Information concerning 520 bicycle accidents and their victims was obtained from medical records and the victims' replies to questionnaires. The analyzed aspects included risk of injury, completeness of accident registrations by police and in hospitals, types of injuries and influence of the cyclists' age and sex, alcohol, fatigue, hunger, haste, physical disability, purpose of cycling, wearing of protective helmet and other clothing, type and quality of road surface, site of accident (road junctions, separate cycle paths, etc.) and turning manoeuvres.

  15. CHALLENGES IN SOURCE TERM MODELING OF DECONTAMINATION AND DECOMMISSIONING WASTES.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.M.

    2006-08-01

    Development of real-time predictive modeling to identify the dispersion and/or source(s) of airborne weapons of mass destruction including chemical, biological, radiological, and nuclear material in urban environments is needed to improve response to potential releases of these materials via either terrorist or accidental means. These models will also prove useful in defining airborne pollution dispersion in urban environments for pollution management/abatement programs. Predicting gas flow in an urban setting on a scale of less than a few kilometers is a complicated and challenging task due to the irregular flow paths that occur along streets and alleys and around buildings of different sizes and shapes, i.e., ''urban canyons''. In addition, air exchange between the outside and buildings and subway areas further complicate the situation. Transport models that are used to predict dispersion of WMD/CBRN materials or to back track the source of the release require high-density data and need defensible parameterizations of urban processes. Errors in the data or any of the parameter inputs or assumptions will lead to misidentification of the airborne spread or source release location(s). The need for these models to provide output in a real-time fashion if they are to be useful for emergency response provides another challenge. To improve the ability of New York City's (NYC's) emergency management teams and first response personnel to protect the public during releases of hazardous materials, the New York City Urban Dispersion Program (UDP) has been initiated. This is a four year research program being conducted from 2004 through 2007. This paper will discuss ground level and subway Perfluorocarbon tracer (PFT) release studies conducted in New York City. The studies released multiple tracers to study ground level and vertical transport of contaminants. This paper will discuss the results from these tests and how these results can be used

  16. HTGR spent fuel element decay heat and source term analysis

    International Nuclear Information System (INIS)

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm3 and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations

  17. Long Term Safe and Secure Management of Disused Sources in Turkey and Security of Radioactive Sources: Maintaining Continuous Global Control of Sources throughout their Life Cycle

    International Nuclear Information System (INIS)

    Disused radioactive sources are generated from research and nuclear applications mainly in medicine, biology, agriculture, quality control in metal processing and construction industries. In the paper, implementation of technical steps for long term safe and secure management of disused sources in Turkey are presented. After classification and pretreatment of spent sealed sources, characterization of the each disused radioactive source was done by using appropriate analysis methods. Later on, the sources were dismantled from its original shield in hot cell and immobilized in a concrete/lead shield for long term storage. (author)

  18. Persistence of airline accidents.

    Science.gov (United States)

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation.

  19. Accidents, risks and consequences

    International Nuclear Information System (INIS)

    Although the accident at Chernobyl can be considered as the worst accident in the world, it could have been worse. Other far worse situations are considered, such as a nuclear weapon hitting a nuclear reactor. Indeed the accident at Chernobyl is compared to a nuclear weapon. The consequences of Chernobyl in terms of radiation levels are discussed. Although it is believed that a similar accident could not occur in the United Kingdom, that possibility is considered. It is suggested that emergency plans should be made for just such an eventuality. Even if Chernobyl could not happen in the UK, the effects of accidents are international. The way in which nuclear reactor accidents happen is explored, taking the 1957 Windscale fire, Three Mile Island and Chernobyl as examples. Reactor designs and accident scenarios are considered. The different reactor designs are listed. As well as the Chernobyl RBMK design it is suggested that the light water reactors also have undesirable features from the point of view of safety. (U.K.)

  20. Severe accident research and management in Nordic Countries - A status report

    International Nuclear Information System (INIS)

    have been mainly concentrated on further development of accident management strategies and aids for source term predictions whereas in Finland in addition to further development of accident management strategies some important plant modifications have been carried out. (au)

  1. Severe accident research and management in Nordic Countries - A status report

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, SKI (Sweden)] (ed.)

    2002-01-01

    have been mainly concentrated on further development of accident management strategies and aids for source term predictions whereas in Finland in addition to further development of accident management strategies some important plant modifications have been carried out. (au)

  2. Soviet submarine accidents

    International Nuclear Information System (INIS)

    Although the Soviet Union has more submarines than the NATO navies combined, and the technological superiority of western submarines is diminishing, there is evidence that there are more accidents with Soviet submarines than with western submarine fleets. Whether this is due to inadequate crews or lower standards of maintenance and overhaul procedures is discussed. In particular, it is suggested that since the introduction of nuclear powered submarines, the Soviet submarine safety record has deteriorated. Information on Soviet submarine accidents is difficult to come by, but a list of some 23 accidents, mostly in nuclear submarines, between 1966 and 1986, has been compiled. The approximate date, class or type of submarine, the nature and location of the accident, the casualties and damage and the source of information are tabulated. (U.K.)

  3. Source-term and building-wake consequence modeling for the GODIVA IV reactor at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    The objectives of this work were to evaluate the consequences of a postulated accident to on-site security personnel stationed near the facility during operations of the Godiva IV critical assembly and to identify controls needed to protest these personnel in case of an extreme criticality excursion equivalent to a design-basis accident (DBA). Godiva IV, one of several critical assemblies operated by Los Alamos National Laboratory (LANL), is located within the Kiva III facility at the Technical Area 18 (TA-18) complex. The TA-18 area is located in a canyon surrounded by complex terrain features such as a steep adjacent hillside and tall stands of fir trees. This analysis was motivated by the need to evaluate the air concentrations and radiological exposure consequences to on-site personnel (guards) located within 40 to 100 m of the facility. GODIVA IV is a highly enriched 235U metal-fuel, fast-burst assembly. The DBA was defined to be a $1.40 critical pulse, which leads to an approximate burst yield of 1.3 x 1018 fissions (or ≅41.6 MJ). The DBA is postulated to lead to partial melt of the reactor assembly (approximately10% of the fuel), with subsequent release of fission products to the environment. The authors present the methodology and results of the source-term calculations, building ventilation rates, air concentrations, and consequence calculations that were performed using a multidisciplinary approach with several phenomenology models. Identification of controls needed to mitigate the consequences to near-field receptors is discussed

  4. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    International Nuclear Information System (INIS)

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value

  5. The Chernobyl accident as a source of new radiological knowledge: implications for Fukushima rehabilitation and research programmes

    International Nuclear Information System (INIS)

    The accident at the Chernobyl nuclear power plant in Ukraine in 1986 caused a huge release of radionuclides over large areas of Europe. During large scale activities focused on overcoming of its negative consequences for public health, various research programmes in radioecology, dosimetry and radiation medicine were conducted. New knowledge was applied internationally in substantial updating of radiation protection systems for emergency and existing situations of human exposure, for improvement of emergency preparedness and response. Radioecological and dosimetry models were significantly improved and validated with numerous measurement data, guidance on environmental countermeasures and monitoring elaborated and tested. New radiological knowledge can be of use in the planning and implementation of rehabilitation programmes in Japan following the Fukushima nuclear accident. In particular, the following activity areas would benefit from application of the Chernobyl experience: strategy of rehabilitation, and technology of settlement decontamination and of countermeasures applied in agriculture and forestry. The Chernobyl experience could be very helpful in planning research activities initiated by the Fukushima radionuclide fallout, i.e. environmental transfer of radionuclides, effectiveness of site-specific countermeasures, nationwide dose assessment, health effect studies, etc. (paper)

  6. Robertson-Walker fluid sources endowed with rotation characterised by quadratic terms in angular velocity parameter

    OpenAIRE

    Wiltshire, R. J.

    2003-01-01

    Einstein's equations for a Robertson-Walker fluid source endowed with rotation Einstein's equations for a Robertson-Walker fluid source endowed with rotation are presented upto and including quadratic terms in angular velocity parameter. A family of analytic solutions are obtained for the case in which the source angular velocity is purely time-dependent. A subclass of solutions is presented which merge smoothly to homogeneous rotating and non-rotating central sources. The particular solution...

  7. Revised Severe Accident Research Program plan, FY 1990--1992

    International Nuclear Information System (INIS)

    For the past 10 years, since the Three Mile Island accident, the NRC has sponsored an active research program on light-water-reactor severe accidents as part of a multi-faceted approach to reactor safety. This report describes the revised Severe Accident Research Program (SARP) and how the revisions are designed to provide confirmatory information and technical support to the NRC staff in implementing the staff's Integration Plan for Closure of Severe Accident Issues as described in SECY-88-147. The revised SARP addresses both the near-term research directed at providing a technical basis upon which decisions on important containment performance issues can be made and the long-term research needed to confirm and refine our understanding of severe accidents. In developing this plan, the staff recognized that the overall goal is to reduce the uncertainties in the source term sufficiently to enable the staff to make regulatory decisions on severe accident issues. However, the staff also recognized that for some issues it may not be practical to attempt to further reduce uncertainties, and some regulatory decisions or conclusions will have to be made with full awareness of existing uncertainties. 2 figs., 1 tab

  8. 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.

  9. 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.

  10. 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.

  11. Resuspension of fission products during severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    This report investigates the influence of resuspension phenomena on the overall radiological source term of core melt accidents in a pressurized water reactor. A review of the existing literature is given and the literature data are applied to calculations of the source term. A large scatter in the existing data was found. Depending on the scenario and on the data set chosen for the calculations the relative influence of resuspended fission products on the source term ranges from dominant to negligible. (orig.)

  12. The Contribution of Sources to the Sustained Elevated Inventory of (137)Cs in Offshore Waters East of Japan after the Fukushima Dai-ichi Nuclear Power Station Accident.

    Science.gov (United States)

    Takata, Hyoe; Kusakabe, Masashi; Inatomi, Naohiko; Ikenoue, Takahito; Hasegawa, Kazuyuki

    2016-07-01

    We have evaluated the contribution of sources of (137)Cs to the inventory of radiocesium in waters (surface area: 6160 km(2), water volume: 753 km(3)) off Fukushima Prefecture and neighboring prefectures from May 2011 to February 2015. A time-series of the inventory of (137)Cs in the offshore waters revealed a clearly decreasing trend from May 2011 (283.4 TBq) to February 2015 (1.89 TBq). The (137)Cs inventory about four years after the accident was approximately twice the background inventory of 1.1 TBq. The magnitudes of the (137)Cs influxes from sources into offshore waters for periods of 182-183 days were estimated from the first period (1 October 2011 to 31 March 2012: 15.3 TBq) to the last period (1 October 2014 to 31 March 2015: 0.41 TBq). We assumed that three sources contributed (137)Cs: continuous direct discharge from the Fukushima Dai-ichi Nuclear Power Station (FNPS) even after the massive discharge in late March 2011, desorption/dissolution from sediments, and fluvial input. Quantification of these sources indicated that the direct discharge from the FNPS is the principal source of (137)Cs to maintain the relatively high inventory in the offshore area. PMID:27282171

  13. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  14. Existence and Asymptotic Stability of Solutions for Hyperbolic Differential Inclusions with a Source Term

    Directory of Open Access Journals (Sweden)

    Park Jong Yeoul

    2007-01-01

    Full Text Available We study the existence of global weak solutions for a hyperbolic differential inclusion with a source term, and then investigate the asymptotic stability of the solutions by using Nakao lemma.

  15. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  16. Procedures for conducting probabilistic safety assessments of nuclear power plants (level 2). Accident progression, containment analysis and estimation of accident source terms

    International Nuclear Information System (INIS)

    The present publication on Level 2 PSA is based on a compilation and review of practices in various Member States. It complements Safety Series No. 50-P-4, issued in 1992, on Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1). Refs, figs and tabs

  17. Short and medium-term medical follow-up program for victims of the Goiania accident, Brazil: a proposal

    International Nuclear Information System (INIS)

    The medical surveillance of the individuals who supposedly received high radiation doses due to the breaking of Cesium-137 source from gamma radiation therapy apparatus, is presented. The study includes a serie of evaluations and procedures such as periodic clinical examinations, and diagnostic procedures. Among the special procedures unusual in routine medical practice, the radiochemical analysis of urine and feces, and the whole-body counting are mentioned, to evaluate the total and residual body burden of Cesium-137 in the organism. Following the acute phase of radiation effects on the organism, such examinations intend to detect, early, the functional or organic changes which may appear during the first years after exposure. (M.C.K.)

  18. On the numerical solution of diffusion-reaction equations with singular source terms

    NARCIS (Netherlands)

    M. Ashyraliyev; J.G. Blom; J.G. Verwer

    2008-01-01

    A numerical study is presented of reaction-diffusion problems having singular reaction source terms, singular in the sense that within the spatial domain the source is defined by a Dirac delta function expression on a lower dimensional surface.A consequence is that solutions will be continuous, but

  19. On the numerical solution of diffusion-reaction equations with singular source terms

    NARCIS (Netherlands)

    Ashyraliyev, M.; Blom, J.G.; Verwer, J.G.

    2008-01-01

    A numerical study is presented of reaction–diffusion problems having singular reaction source terms, singular in the sense that within the spatial domain the source is defined by a Dirac delta function expression on a lower dimensional surface. A consequence is that solutions will be continuous, but

  20. An Analytic Solution of Hydrodynamic Equations with Source Terms in Heavy Ion Collisions

    OpenAIRE

    Zhuang, Pengfei; Yang, Zhenwei

    2000-01-01

    The energy and baryon densities in heavy ion collisions are estimated by analytically solving a 1+1 dimensional hydrodynamical model with source terms. Particularly, a competition between the energy and baryon sources and the expansion of the system is discussed in detail.

  1. Long-term assessment of airborne radio-cesium after the Fukushima nuclear accident: re-suspension from soil and vegetation

    Science.gov (United States)

    Kajino, Mizuo; Ishizuka, Masahide; Igarashi, Yasuhito; Kita, Kazuyuki; Yoshikawa, Chisato; Inatsu, Masaru

    2016-04-01

    Long-term assessment of Cs-137 re-suspension from contaminated soil and vegetation due to the Fukushima nuclear accident in March 2011 and the on-going emission from the premises of the power plant has been conducted using a numerical simulation, a field experiment on the dust deflation at Namie in the restricted habitation area, and air concentration measurements in and out of the area, Namie and Tsukuba, respectively. The analysis period is one year from December 2012, about one and a half years from the accident, up to December 2013. The surface concentration of Cs-137 at Namie was high in the summer (~1 mBq/m3) and low in the winter (0.1-1 mBq/m3). The Cs-137 concentration was about one order smaller in Tsukuba (0.01-0.1 mBq/m3). The differences in the two sites are consistent between the observation and the simulation. Ishizuka et al. (2016) developed a numerical module of Cs-137 re-suspension associated with dust deflation based on the flux measurement in Namie. Using the module, the simulated Cs-137 from soil had a potential to account for the observed surface concentration in Namie in the winter, but underestimated by 1-2 orders of magnitude in the summer. The Tokyo Electric Power Company assessed the Cs-137 emission from the reactor buildings in 2013 as approximately 1e6 Bq/h. By using the emission rate, the simulation substantially underestimated the observation by 2-3 orders of magnitude in Namie. We simulated the re-suspension from vegetation applying a seasonal variation as a function of the green fraction map, obtained from the database of Chen and Dudhia (2001). With the constant re-suspension rate of 1e-7 [/h], the simulated vegetation re-suspension quantitatively accounted for the observed surface concentration together with its seasonal variation. Still, so far, the re-suspension mechanism has not been fully understood and thus further investigations for the understanding of the mechanisms and its long-term effects on the environment are needed.

  2. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Kenton, M.A. [Dames and Moore, Westmont, IL (United States)

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  3. From Open Source to long-term sustainability: Review of Business Models and Case studies

    OpenAIRE

    Chang, Victor; Mills, Hugo; Newhouse, Steven

    2007-01-01

    This paper presents several case studies to demonstrate how open source software can achieve long-term sustainability by adopting the relevant business models. The objectives of this paper are to study the different models, processes, and legal/licence requirements that have been successful for such transformations. We classify the business models used in the open source area into five types: (a) Support Contracts; (b) Split Licensing; (c) Community; (d) Valued-added closed source; (e) Macro ...

  4. Long-term investigations of radiocaesium activity concentrations in carps in north Croatia after the Chernobyl accident

    CERN Document Server

    Franic, Z

    2007-01-01

    Long-term investigations of radiocaesium activity concentrations in carps in the Republic of Croatia are presented. The radiocaesium levels in carps decreased exponentially and the effective ecological half-life of 137Cs in carps was estimated to be about 1 year for 1987-2002 period and 5 years for 1993-2005 period. The observed 134Cs:137Cs activity ratio in carps has been found to be similar to the ratio that has been observed in other environmental samples. Concentration factor for carps (wet weight) was roughly estimated to be 128 +/- 74 Lkg-1, which is in reasonable agreement with model prediction based on K+ concentrations in water. Estimated annual effective doses received by 134Cs and 137Cs intake due to consumption of carps for an adult member of Croatian population are small, per caput dose for the 1987 - 2005 estimated to be 0.5 +/- 0.2 microSv. Due to minor freshwater fish consumption in Croatia and low radiocaesium activity concentrations in carps, it can be concluded that carps consumption was no...

  5. Airborne source-term modeling of past and future interim storage practices of Hanford Site waste treatment facilities using airsource model

    International Nuclear Information System (INIS)

    A computer code, AIRSOURCE, was developed to characterize the airborne source terms of the waste treatment facilities at the Hanford Site. This code is based on simple engineering models of liquid to vapor material transport. The present model has data libraries for 72 inorganic, organic, and radioactive materials which are present in the waste treatment streams. The geometries in the code include many of the facilities such as vented tanks, holdup basins, and retention facilities which are used for waste treatment. Recent additions to the model have included aerosols formed by leakage sprays which will be used for accident studies. Airborne source terms were determined for the waste treatment facilities of the past and future waste treatment facilities at the Hanford Site. The waste treatment facilities used in past practices include holdup basins, soil column disposal cribs, and flow-induced air entrainment in process piping. The future practices modeling included source term characterization of large covered holdup basins ranging in capacities from 1 to 6.5 million gallons, pumping stations, filtered vent systems, and effects of maintenance activities and small breaks in the covers. The design of the retention basin cover was critical for facility acceptance. Several materials and designs were investigated. A design based on high-density polyethylene was found to be acceptable. The results of the source term modeling demonstrated that the future practices source terms were less than those of the past practices. The AIRSOURCE code is sufficiently general and has a broad range of potential applications with liquid air interfaces where the liquids can contain both radioactive and hazardous materials

  6. A novel two-stage evaluation system based on a Group-G1 approach to identify appropriate emergency treatment technology schemes in sudden water source pollution accidents.

    Science.gov (United States)

    Qu, Jianhua; Meng, Xianlin; Hu, Qi; You, Hong

    2016-02-01

    Sudden water source pollution resulting from hazardous materials has gradually become a major threat to the safety of the urban water supply. Over the past years, various treatment techniques have been proposed for the removal of the pollutants to minimize the threat of such pollutions. Given the diversity of techniques available, the current challenge is how to scientifically select the most desirable alternative for different threat degrees. Therefore, a novel two-stage evaluation system was developed based on a circulation-correction improved Group-G1 method to determine the optimal emergency treatment technology scheme, considering the areas of contaminant elimination in both drinking water sources and water treatment plants. In stage 1, the threat degree caused by the pollution was predicted using a threat evaluation index system and was subdivided into four levels. Then, a technique evaluation index system containing four sets of criteria weights was constructed in stage 2 to obtain the optimum treatment schemes corresponding to the different threat levels. The applicability of the established evaluation system was tested by a practical cadmium-contaminated accident that occurred in 2012. The results show this system capable of facilitating scientific analysis in the evaluation and selection of emergency treatment technologies for drinking water source security. PMID:26449677

  7. Status of safety technology for radiological consequence assessment of postulated accidents in liquid metal fast breeder reactors, Canoga Park, California, 29 July--31 July 1975

    International Nuclear Information System (INIS)

    State-of-the-art capabilities are examined for prediction and mitigation of radiological consequences of postulated LMFBR accidents. The following topics are treated: radioactive source terms, sodium reactions, aerosol behavior, radiological dose assessment, and engineered safeguards. (U.S.)

  8. Seminar on Comparative assessment of the environmental impact of radionuclides released during three major nuclear accidents: Kyshtym, Windscale, Chernobyl. Vol. 1

    International Nuclear Information System (INIS)

    These proceedings of seminar on comparative assessment of the environmental impact of radionuclides released during three major nuclear accidents (Kyshtym, Windscale, Chernobyl) are divided into 5 parts bearing on: part 1: accident source terms; part 2: atmospheric dispersion, resuspension, chemical and physical forms of contamination; part 3: environmental contamination and transfer; part 4: radiological implications for man and his environment; part 5: countermeasures

  9. Mitigation of Hydrogen Hazards in Severe Accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Consideration of severe accidents in nuclear power plants is an essential component of the defence in depth approach in nuclear safety. Severe accidents have very low probabilities of occurring, but may have significant consequences resulting from the degradation of nuclear fuel. The generation of hydrogen and the risk of hydrogen combustion, as well as other phenomena leading to overpressurization of the reactor containment in case of severe accidents, represent complex safety issues in relation to accident management. The combustion of hydrogen, produced primarily as a result of heated zirconium metal reacting with steam, can create short term overpressure or detonation forces that may exceed the strength of the containment structure. An understanding of these phenomena is crucial for planning and implementing effective accident management measures. Analysis of all the issues relating to hydrogen risk is an important step for any measure that is aimed at the prevention or mitigation of hydrogen combustion in reactor containments. The main objective of this publication is to contribute to the implementation of IAEA Safety Standards, in particular, two IAEA Safety Requirements: Safety of Nuclear Power Plants: Design and Safety of Nuclear Power Plants: Operation. These Requirements publications discuss computational analysis of severe accidents and accident management programmes in nuclear power plants. Specifically with regard to the risk posed by hydrogen in nuclear power reactors, computational analysis of severe accidents considers hydrogen sources, hydrogen distribution, hydrogen combustion and control and mitigation measures for hydrogen, while accident management programmes are aimed at mitigating hydrogen hazards in reactor containments.

  10. Finite Element Solutions for the Space Fractional Diffusion Equation with a Nonlinear Source Term

    Directory of Open Access Journals (Sweden)

    Y. J. Choi

    2012-01-01

    Full Text Available We consider finite element Galerkin solutions for the space fractional diffusion equation with a nonlinear source term. Existence, stability, and order of convergence of approximate solutions for the backward Euler fully discrete scheme have been discussed as well as for the semidiscrete scheme. The analytical convergent orders are obtained as O(k+hγ˜, where γ˜ is a constant depending on the order of fractional derivative. Numerical computations are presented, which confirm the theoretical results when the equation has a linear source term. When the equation has a nonlinear source term, numerical results show that the diffusivity depends on the order of fractional derivative as we expect.

  11. Radiological accidents balance in medicine; Bilan des accidents radiologiques en medecine

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, J.C.

    1995-12-31

    This work deals with the radiological accidents in medicine. In medicine, the radiation accidents on medical personnel and patients can be the result of over dosage and bad focusing of radiotherapy sealed sources. Sometimes, the accidents, if they are unknown during a time enough for the source to be spread and to expose a lot of persons (in the case of source dismantling for instance) can take considerable dimensions. Others accidents can come from bad handling of linear accelerators and from radionuclide kinetics in some therapies. Some examples of accidents are given. (O.L.). 11 refs.

  12. Post-accident gas generation from radiolysis of organic materials

    International Nuclear Information System (INIS)

    This report presents a methodology for estimating the gas generation rates resulting from radiolysis of organic materials in paints and electrical cable insulation inside a nuclear reactor containment building under design basis accident conditions. The methodology was based on absorption of the radiation energies from the post-accident fission products and the assumed gas yields of the irradiated materials. A sample calculation was made using conservative assumptions, plant-specific data of a nuclear power plant, and a radiation source term which took into account the time-dependent release and physico-chemical behavior of the fission products

  13. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  14. Short- and long-term patterns of 137Cs in fish and other aquatic organisms of small forest lakes in southern Finland since the Chernobyl accident

    International Nuclear Information System (INIS)

    We summarize the patterns of 137Cs activity concentrations and transfer into fish and other biota in four small forest lakes in southern Finland during a twenty-year period following the Chernobyl accident in April 1986. The results from summer 1986 showed fastest accumulation of 137Cs into planktivorous fishes, i.e. along the shortest food chains. Since 1987, the highest annual mean values of 137Cs have been recorded in fish occupying the highest trophic levels, for perch (Perca fluviatilis) 13,600 Bq/kg (ww) and for pike (Esox lucius) 20,700 Bq/kg (ww). At the same time, activity concentrations of 137Cs in crustacean zooplankton and Asellus aquaticus have ranged between 1000 and 19,500 Bq/kg (dw). In 2006, 5–28% of the 1987 137Cs activity concentration levels were still present in perch and pike. Since 1989 their 137Cs activity concentrations in oligohumic seepage lakes have remained significantly higher than in polyhumic drainage lakes due to the increased transfer of 137Cs into fish in the seepage lakes with lower electrolyte concentrations, longer water retention times and lower sedimentation rate. - Highlights: ► In summer 1986 highest 137Cs levels in planktivore fishes of short food chains. ► Since 1987 highest 137Cs were recorded in predatory fish of highest trophic levels. ► High variation found also in 137Cs of crustacean zooplankton and Asellus aquaticus. ► In long-term fish 137Cs higher in clear seepage lakes than in humic drainage lakes. ► Increased transfer of 137Cs into fish in seepage lakes was the reason suggested.

  15. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  16. 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

  17. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  18. Robertson-Walker fluid sources endowed with rotation characterised by quadratic terms in angular velocity parameter

    CERN Document Server

    Wiltshire, R J

    2003-01-01

    Einstein's equations for a Robertson-Walker fluid source endowed with rotation Einstein's equations for a Robertson-Walker fluid source endowed with rotation are presented upto and including quadratic terms in angular velocity parameter. A family of analytic solutions are obtained for the case in which the source angular velocity is purely time-dependent. A subclass of solutions is presented which merge smoothly to homogeneous rotating and non-rotating central sources. The particular solution for dust endowed with rotation is presented. In all cases explicit expressions, depending sinusoidally on polar angle, are given for the density and internal supporting pressure of the rotating source. In addition to the non-zero axial velocity of the fluid particles it is shown that there is also a radial component of velocity which vanishes only at the poles. The velocity four-vector has a zero component between poles.

  19. Investigation regarding the long-term security developments in the Swedish nuclear power and the response to the accident at Fukushima; Utredning avseende den laangsiktiga saekerhetsutvecklingen i den svenska kaernkraften och aatgaerder med anledning av olyckan i Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Skaanberg, Lars

    2012-07-01

    Swedish nuclear plants need to continue to work on analysis and actions in the plants, partly to meet the demands of legislation and agreed action plans, and partly due to additional security requirements on account of experiences from the Fukushima Dai-ichi accident, stress tests, security investigations and investigations relating to physical protection. It is also essential to continue with safety improvements to gradually increase margins against unforeseen events in aging plants during long-term operation.

  20. 7 CFR 1822.268 - Rates, terms, and source of funds.

    Science.gov (United States)

    2010-01-01

    ... AGRICULTURE LOANS AND GRANTS PRIMARILY FOR REAL ESTATE PURPOSES RURAL HOUSING LOANS AND GRANTS Rural Housing... 7 Agriculture 12 2010-01-01 2010-01-01 false Rates, terms, and source of funds. 1822.268 Section 1822.268 Agriculture Regulations of the Department of Agriculture (Continued) RURAL HOUSING...

  1. POINTWISE AND UPWIND DISCRETIZATIONS OF SOURCE TERMS IN OPEN-CHANNEL FLOOD ROUTING

    Institute of Scientific and Technical Information of China (English)

    MENG Jian; CAO Zhi-xian; CARLING Paul A.

    2006-01-01

    Upwind algorithms are becoming progressively popular for river flood routing due to their capability of resolving trans-critical flow regimes. For consistency, these algorithms suggest natural upwind discretization of the source term, which may be essential for natural channels with irregular geometry. Yet applications of these upwind algorithms to natural river flows are rare, and in such applications the traditional and simpler pointwise, rather than upwind discretization of the source term is used. Within the framework of a first-order upwind algorithm, this paper presents a comparison of upwind and pointwise discretizations of the source term. Numerical simulations were carried out for a selected irregular channel comprising a pool-riffle sequence in the River Lune, England with observed data. It is shown that the impact of pointwise discretization, compared to the upwind, is appreciable mainly in flow zones with the Froude number closer to or larger than unity. The discrepancy due to pointwise and upwind discretizations of the source term is negligible in flow depth and hence in water surface elevation, but well manifested in mean velocity and derived flow quantities. Also the occurrence of flow reversal and equalisation over the pool-riffle sequence in response to increasing discharges is demonstrated.

  2. Short-Term Memory Stages in Sign vs. Speech: The Source of the Serial Span Discrepancy

    Science.gov (United States)

    Hall, Matthew L.; Bavelier, Daphne

    2011-01-01

    Speakers generally outperform signers when asked to recall a list of unrelated verbal items. This phenomenon is well established, but its source has remained unclear. In this study, we evaluate the relative contribution of the three main processing stages of short-term memory--perception, encoding, and recall--in this effect. The present study…

  3. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  4. Radionuclide release calculations for selected severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. (Battelle Columbus Div., OH (USA))

    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.

  5. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

    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

  6. Effects on accidents of changes in the use of studded tyres in major cities in Norway: a long-term investigation.

    Science.gov (United States)

    Elvik, Rune; Fridstrøm, Lasse; Kaminska, Joanna; Meyer, Sunniva Frislid

    2013-05-01

    This paper reports the findings of two studies made eleven years apart in Norway (Fridstrøm, 2000; Elvik and Kaminska, 2011) to evaluate effects on accidents of changes in the use of studded tyres in major cities in Norway. The first study covered the period from 1991 to 2000, the second study covered the period from 2002 to 2009. In both these periods, large changes in the percentage of cars using studded tyres were found in the cities that were included in the study. There was, in most cities, a tendency for the use of studded tyres to go down. Effects of these changes on injury accidents were evaluated by means of negative binomial regression models, using city and day as the unit of analysis, and including more than twenty explanatory variables in order to control for confounding factors. The effects of changes in the percentage of cars using studded tyres were well described by an accident modification function (dose-response curve), relating the size of changes in the number of accident to the size of the change in the use of studded tyres. Accidents during the season when the use of studded tyres is permitted were found to increase by about 5 percent if the use of studded tyres was reduced by 25 percentage points (e.g. from 50 to 25 percent) and to decline by about 2 percent when the use of studded tyres increased by 20 percentage points. PMID:23474233

  7. On the group velocity for the shallow water equations with source terms

    International Nuclear Information System (INIS)

    The group velocity of the shallow water according to Saint-Venant's equations with source terms is analyzed. For these equations the classical group velocity relation describes the propagation velocity of a wave packet in normal dispersion e.g. in homogeneous form. The presence of source terms in momentum equation, such as the bottom slope and the friction of bed, gives rise to a singularity in the dispersion relation, causing an anomalous dispersion in which the standard group velocity becomes infinite. This non-physical result reveals that, for non-homogeneous shallow water equations, the classic relation is not appropriate for describing a wave packet. In order to overcome this difficulty we consider an asymptotic approximation, based on the Taylor series expansion, for the representation of the propagation velocity of a wave packet. The analysis includes the effects of the friction resistance term, Courant number and Froude number. Numerical results are discussed.

  8. Low-level radioactive waste source terms for the 1992 integrated data base

    Energy Technology Data Exchange (ETDEWEB)

    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{sub 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}other{close_quotes}. Fuel cycle commercial LLW source terms are normalized on the basis of net electrical output [MW(e)-year], except for UF{sub 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.

  9. Information from water ingress accident on AVR

    International Nuclear Information System (INIS)

    An ingress of water occurred in the AVR reactor in May 1978. The reactor had been shut down and cooled by forced circulation; liquid water entered the primary circuit from a leak in the superheater, evaporated in passing through the core and condensed in the lower part of the primary circuit and in the ball handling region. Various fission product activities were measured in the water by the AVR scientists and a study was started to identify the sources of these activities and to derive information which could be used in the analyses of water ingress accidents in general. The first part of this study is reported in this note. The possible source terms are considered separately and estimates of their contributions are made, supported by data from previous laboratory experiments where possible. The main conclusion is that valuable information has been derived concerning the desorption of iodine, cesium and strontium from dust and primary circuit surfaces. A minimum programme of measurements and analytical work necessary to increase this information has been identified. An example of the application of the data to a particular accident to a power reactor is given to indicate how the information can affect the calculation of consequences. For the second part of the study, better estimates of the fission product concentrations in the primary circuit prior to the accident and various measurements when the reactor is operating again are required. (orig.)

  10. Fusion of chemical, biological, and meteorological observations for agent source term estimation and hazard refinement

    Science.gov (United States)

    Bieringer, Paul E.; Rodriguez, Luna M.; Sykes, Ian; Hurst, Jonathan; Vandenberghe, Francois; Weil, Jeffrey; Bieberbach, George, Jr.; Parker, Steve; Cabell, Ryan

    2011-05-01

    Chemical and biological (CB) agent detection and effective use of these observations in hazard assessment models are key elements of our nation's CB defense program that seeks to ensure that Department of Defense (DoD) operations are minimally affected by a CB attack. Accurate hazard assessments rely heavily on the source term parameters necessary to characterize the release in the transport and dispersion (T&D) simulation. Unfortunately, these source parameters are often not known and based on rudimentary assumptions. In this presentation we describe an algorithm that utilizes variational data assimilation techniques to fuse CB and meteorological observations to characterize agent release source parameters and provide a refined hazard assessment. The underlying algorithm consists of a combination of modeling systems, including the Second order Closure Integrated PUFF model (SCIPUFF), its corresponding Source Term Estimation (STE) model, a hybrid Lagrangian-Eulerian Plume Model (LEPM), its formal adjoint, and the software infrastructure necessary to link them. SCIPUFF and its STE model are used to calculate a "first guess" source estimate. The LEPM and corresponding adjoint are then used to iteratively refine this release source estimate using variational data assimilation techniques. This algorithm has undergone preliminary testing using virtual "single realization" plume release data sets from the Virtual THreat Response Emulation and Analysis Testbed (VTHREAT) and data from the FUSION Field Trials 2007 (FFT07). The end-to-end prototype of this system that has been developed to illustrate its use within the United States (US) Joint Effects Model (JEM) will be demonstrated.

  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. Long-term security of electrical and control engineering equipment in nuclear power stations to withstand a loss of coolant accident

    International Nuclear Information System (INIS)

    Electrical and control engineering equipment, which has to function even after many years of operation in the event of a fault in a saturated steam atmosphere of 160 C maximum, is essential in nuclear power stations in order to control a loss of coolant accident. The nuclear power station operators have, for this purpose, developed verification strategies for groups of components, by means of which it is ensured that the electrical and control engineering components are capable of dealing with a loss of coolant accident even at the end of their planned operating life. (orig.)

  13. The estimation economic impacts from severe accidents of a nuclear power plant

    International Nuclear Information System (INIS)

    The severe accidents of a nuclear power plant may cause health effects in the exposed population and societal economic impacts or costs. Techniques to assess the consequences of an accident in terms of cost may be applied in studies on the design of plant safety features and in examining countermeasure options as part of emergency planning or in decision making after an accident. In this study, the costs resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of source term release parameters and meteorological data. Also, the costs were estimated for the different scenarios considering seasonal characteristics of Korea. The results can be used as essential inputs in costs/benefit analysis and in developing optimum risk reduction strategies

  14. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  15. A Methodology for a Comprehensive Probabilistic Tsunami Hazard Assessment: Multiple Sources and Short-Term Interactions

    Directory of Open Access Journals (Sweden)

    Grezio Anita

    2015-01-01

    Full Text Available We propose a methodological approach for a comprehensive and total probabilistic tsunami hazard assessment (TotPTHA, in which many different possible source types concur to the definition of the total tsunami hazard at given target sites. In a multi-hazard and multi-risk perspective, the approach allows us to consider all possible tsunamigenic sources (seismic events, slides, volcanic eruptions, asteroids, etc.. In this respect, we also formally introduce and discuss the treatment of interaction/cascade effects in the TotPTHA analysis and we demonstrate how the triggering events may induce significant temporary variations in short-term analysis of the tsunami hazard. In two target sites (the city of Naples and the island of Ischia in Italy we prove the feasibility of the TotPTHA methodology in the multi—source case considering near submarine seismic sources and submarine mass failures in the study area. The TotPTHA indicated that the tsunami hazard increases significantly by considering both the potential submarine mass failures and the submarine seismic events. Finally, the importance of the source interactions is evaluated by applying a triggering seismic event that causes relevant changes in the short-term TotPTHA.

  16. Containment performance analyses for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.

    1992-10-01

    This paper discusses salient aspects of methodology, assumptions, and modeling of various features related to estimation of source terms from two conservatively scoped severe accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code.

  17. Long-term changes in the solar photosphere associated with changes in the coronal source flux

    OpenAIRE

    Foste, S.; Lockwood, Mike

    2001-01-01

    Using sunspot observations from Greenwich and Mount Wilson, we show that the latitudinal spread of sunspot groups has increased since 1874, in a manner that closely mirrors the long-term (similar to 100 year) changes in the coronal source flux, F-s, as inferred from geomagnetic activity. This latitude spread is shown to be well correlated with the flux emergence rate required by the model of the coronal source flux variation by Solanki er al. [2000]. The time constant for the decay of this op...

  18. On stability of difference schemes. Central schemes for hyperbolic conservation laws with source terms

    OpenAIRE

    Mond, M.; Borisov, V. S.

    2009-01-01

    The stability of difference schemes for, in general, hyperbolic systems of conservation laws with source terms are studied. The basic approach is to investigate the stability of a non-linear scheme in terms of its cor- responding scheme in variations. Such an approach leads to application of the stability theory for linear equation systems to establish stability of the corresponding non-linear scheme. It is established the notion that a non-linear scheme is stable if and only if the correspon...

  19. Final report of the inter institutional project ININ-CNSNS 'Source Terms specific for the CNLV'

    International Nuclear Information System (INIS)

    The purpose of the project inter institutional ININ-CNSNS 'Source Terms Specifies for the CNLV' it is the one of implanting in the computer CYBER (CDC 180-830) of the ININ, the 'Source Term Code Package' (STCP) and to make the operation tests and corresponding operation using the data of the sample problem, for finally to liberate the package, all time that by means of the analysis of the results it is consider appropriate. In this report the results of the are presented simulation of the sequence 'Energy Losses external' (Station blackout) and 'Lost total of CA with failure of the RCIC and success of the HPCS' both with data of the Laguna Verde Central. (Author)

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

    International Nuclear Information System (INIS)

    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

  1. Final report on shipping-cask sabotage source-term investigation

    International Nuclear Information System (INIS)

    A need existed to estimate the source term resulting from a sabotage attack on a spent nuclear fuel shipping cask. An experimental program sponsored by the US NRC and conducted at Battelle's Columbus Laboratories was designed to meet that need. In the program a precision shaped charge was fired through a subscale model cask loaded with segments of spent PWR fuel rods and the radioactive material released was analyzed. This report describes these experiments and presents their results

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

    International Nuclear Information System (INIS)

    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

  3. Godunov type scheme for the linear wave equation with Coriolis source term

    OpenAIRE

    Audusse, Emmanuel; Dellacherie, Stéphane; Hieu, Do Minh; Omnes, Pascal; Penel, Yohan

    2015-01-01

    We propose a method to explain the behavior of the Godunov finite volume scheme applied to the linear wave equation with Coriolis source term at low Froude number. In particular, we use the Hodge decomposition and we study the properties of the modified equation associated to the Godunov scheme. Based on the structure of the discrete kernel of the linear operator discretized by using the Godunov scheme, we clearly explain the inaccuracy of the classical Godunov scheme at low Froude number and...

  4. Application of probability distributions for quantifying uncertainty in radionuclide source terms for Seabrook risk assessment

    International Nuclear Information System (INIS)

    In some of the recent probabilistic safety assessments, discrete probability distributions (DPDs) have been developed to express, in a quantitative form, estimates of the uncertainty and conservatism in the point estimate source term values. In the DPD approach, distributed, discrete factors, which are multipliers on the point estimate values by the selected factor are made based on available data, calculations, and engineering judgment. Initial application of the DPD approach to source terms for risk analysis was based largely on engineering judgment after review of applicable data. However, in more recent applications of the DPD approach, results from an extensive review of existing experimental data and applied calculations have been factored into the estimates. Programs currently in progress, largely sponsored by NRC and EPRI, are beginning to yield significant new information upon which to base improved estimates for the magnitude of radionuclide source terms. The most extensive of the reviews of existing data for application to the DPD approach was that performed as part of the risk assessment for the proposed Sizewell B PWR. As part of the Seabrook risk study, DPD values specifically for that plant were developed based on the Sizewell approach. They represent a significant update to the Sizewell DPD values. In addition, DPD values were developed for associated release parameters which also affect the consequence calculations

  5. Farm accidents in children.

    Science.gov (United States)

    Cogbill, T H; Busch, H M; Stiers, G R

    1985-10-01

    During a 6 1/2 year period, 105 children were admitted to the hospital as the result of trauma that occurred on farms. The mechanism of injury was animal related in 42 (40%), tractor or wagon accident in 28 (26%), farm machinery in 21 (20%), fall from farm building in six (6%), and miscellaneous in eight (8%). Injury Severity Score was calculated for each patient. An Injury Severity Score of greater than or equal to 25 was determined for 11 children (11%). Life-threatening injuries, therefore, are frequently the result of childhood activities that take place in agricultural environments. The most common injuries were orthopedic, neurologic, thoracoabdominal, and maxillofacial. There was one death in the series, and only one survivor sustained major long-term disability. Such injuries are managed with optimal outcome in a regional trauma center. Educational programs with an emphasis on prevention and safety measures may reduce the incidence of farm accidents. PMID:4047799

  6. Schematic way to find solution of the outcoupled matter wave with a source term

    Energy Technology Data Exchange (ETDEWEB)

    Prayitno, T. B. [Physics Department, Faculty of Mathematics and Natural Science, Universitas Negeri Jakarta, Jl. Pemuda Rawamangun No. 10, Jakarta, 13220 (Indonesia)

    2013-09-09

    We propose a schematic way to obtain solution of the outcoupled atom laser beam wave function in the presence of a source term where the beam is influenced by gravity. In this case, we only focus on the external potentials inside the region of Bose-Einstein condensate that are generated by electromagnetic source and gravity. Since the evolution of the atom laser beam can be portrayed through the ordinary Schrödinger equation with a source, we are allowed to express the general solution as the superposition of the homogeneous solution and particular solution. With the given external potentials and ansatz solutions, we attain that the obtained energy depends on the parameter constituting to the ratio between the longitudinal frequency and transverse frequency.

  7. 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.

  8. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    From 10 to 12 November 1986 some 80 delegates met under the auspices of the CEC working group on the safety of light-water reactors. The participants from EC Member States were joined by colleagues from Sweden, Finland and the USA and met to discuss the subject of severe accidents in LWRs. Although this seminar had been planned well before Chernobyl, the ''severe-accident-that-really-happened'' made its mark on the seminar. The four main seminar topics were: (i) high source-term accident sequences identified in PSAs, (ii) containment performance, (iii) mitigation of core melt consequences, (iv) severe accident management in LWRs. In addition to the final panel discussion there was also a separate panel discussion on lessons learned from the Chernobyl accident. These proceedings include the papers presented during the seminar and they are arranged following the seminar programme outline. The presentations and discussions of the two panels are not included in the proceedings. The general conclusions and directions following from these two panels were, however, considered in a seminar review paper which was published in the March 1987 issue of Nuclear Engineering International

  9. Validation and verification of accident consequence assessment models

    Energy Technology Data Exchange (ETDEWEB)

    Homma, T.; Togawa, O. [Japan Atomic Energy Research Inst., Tokyo (Japan); Takahashi, T. [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Arkhipov, A.N. [Chernobyl Science and Technology Centre for International Research (Ukraine)

    2001-03-01

    An accident consequence assessment code, OSCAAR, primarily designed by Japan Atomic Energy Research Institute (JAERI) for use in probabilistic safety assessment (PSA) of nuclear reactors in Japan, was applied to use for siting, emergency planning, and development of design criteria, and in the comparative risk studies of different energy systems. After verifying the code system through the international code comparison organized by CEC and OECD/NEA, the validation and improvements of the individual models and the verification of the whole OSCAAR code system were made. The cooperative research between Chernobyl Science and Technology Center for International Research (CHESCIR) and JAERI provided a valuable opportunity to test the performance of the accident consequence assessment models by comparing the model predictions with data obtained in the Chernobyl accidents. The predictive capabilities of OSCAAR were demonstrated using the accident source term and meteorological data for estimating the early exposure to the public occurred during and shortly after plume passage. The calculations indicated that ground-shine dose and inhalation dose, particularly from large nonvolatile particulates were the main contributors in the early stage of the accident. (S. Ohno)

  10. Historical analysis of US pipeline accidents triggered by natural hazards

    Science.gov (United States)

    Girgin, Serkan; Krausmann, Elisabeth

    2015-04-01

    Natural hazards, such as earthquakes, floods, landslides, or lightning, can initiate accidents in oil and gas pipelines with potentially major consequences on the population or the environment due to toxic releases, fires and explosions. Accidents of this type are also referred to as Natech events. Many major accidents highlight the risk associated with natural-hazard impact on pipelines transporting dangerous substances. For instance, in the USA in 1994, flooding of the San Jacinto River caused the rupture of 8 and the undermining of 29 pipelines by the floodwaters. About 5.5 million litres of petroleum and related products were spilled into the river and ignited. As a results, 547 people were injured and significant environmental damage occurred. Post-incident analysis is a valuable tool for better understanding the causes, dynamics and impacts of pipeline Natech accidents in support of future accident prevention and mitigation. Therefore, data on onshore hazardous-liquid pipeline accidents collected by the US Pipeline and Hazardous Materials Safety Administration (PHMSA) was analysed. For this purpose, a database-driven incident data analysis system was developed to aid the rapid review and categorization of PHMSA incident reports. Using an automated data-mining process followed by a peer review of the incident records and supported by natural hazard databases and external information sources, the pipeline Natechs were identified. As a by-product of the data-collection process, the database now includes over 800,000 incidents from all causes in industrial and transportation activities, which are automatically classified in the same way as the PHMSA record. This presentation describes the data collection and reviewing steps conducted during the study, provides information on the developed database and data analysis tools, and reports the findings of a statistical analysis of the identified hazardous liquid pipeline incidents in terms of accident dynamics and

  11. Terms and terminological combinations as a source of image in M. Sholokhov’s works

    Directory of Open Access Journals (Sweden)

    Mukhamedganova Alina Mikhailovna

    2015-12-01

    Full Text Available The article researches the sources of imagery and expressiveness in the works of M. Sholokhov, as the terms and terminological combinations. The texts of the writer shaped the function of terminological linguistic units in its implementation, as a rule, when they are used as a comparison, as well as their figurative metaphorical and metonymic use. However, M. Sholokhov as a true master of literary work uses non- traditional methods of realization of terminological linguistic unit in the art-works of pictorial means, giving it a completely new meaning, which, however, is not portable because acquired value of terminological linguistic unit does not depend on the direct meaning of the defining reality of a special sphere of human activity. So, the new meaning of the terms may not be motivated to direct meaning, and the phonetic structure of the units. In other words, these meanings are motivated by individual associations of the hero of the novel, which cause a phonetic structure of terminological units. Also it is analyzed example, when the term of the military sector in the speech of a character realized the function, however, in this example, the term itself is not important, only the essential fact that the unit is a military term.

  12. An industry perspective on strategies for the long term control and management of sources

    International Nuclear Information System (INIS)

    A strategic approach to a long term control and management programme for radioactive sources requires a life cycle source management philosophy. This is a cornerstone to an effective, comprehensive and robust programme. Regulators, manufacturers, suppliers and users of radioactive sources all have specific but complementary and even overlapping roles and responsibilities. The IAEA plays a pivotal role, providing a robust foundation for all the initiatives that are being taken to achieve the objective of safety and security. When establishing policies, regulations and practices, a risk-informed approach is fundamental to the security of sources and devices. The IAEA also brings together the key stakeholders to establish voluntary codes, standards and guidelines. These stakeholders include national regulators, non-governmental organizations (NGOs) and industry, who accompany their national competent authorities to key meetings hosted by the IAEA. Collaboration amongst these stakeholders also establishes the protocol, the enabling mechanism, for a level playing field for all suppliers and manufacturers engaged in international trade. Building on the foundation established by the IAEA, national competent authorities, manufacturers, suppliers and users establish the system of policies, rules, regulations and practices to achieve the objective of safety and security. The engagement of industry is a key success factor for achieving the objective of safety and security during the source life cycle. Manufacturers and suppliers worldwide also think it is important that industry participate in actively developing strategies for the long term control and management of sources. There is a desire of industry as a whole to collaborate with the IAEA in developing international policy, to forge a strong relationship with national legislators and regulators, and to facilitate communication, education and awareness amongst key stakeholders. Several Technical Meetings, held by the

  13. Potential health risks from postulated accidents involving the Pu-238 RTG (radioisotope thermoelectric generator) on the Ulysses solar exploration mission

    Energy Technology Data Exchange (ETDEWEB)

    Goldman, M. (California Univ., Davis, CA (USA)); Nelson, R.C. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Bollinger, L. (Air Force Inspection and Safety Center, Kirtland AFB, NM (USA)); Hoover, M.D. (Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (USA). Inhalation Toxicology Research Inst.); Templeton, W. (Pacific Northwest Lab., Richland, WA (USA)); Anspaugh, L. (Lawren

    1990-11-02

    Potential radiation impacts from launch of the Ulysses solar exploration experiment were evaluated using eight postulated accident scenarios. Lifetime individual dose estimates rarely exceeded 1 mrem. Most of the potential health effects would come from inhalation exposures immediately after an accident, rather than from ingestion of contaminated food or water, or from inhalation of resuspended plutonium from contaminated ground. For local Florida accidents (that is, during the first minute after launch), an average source term accident was estimated to cause a total added cancer risk of up to 0.2 deaths. For accidents at later times after launch, a worldwide cancer risk of up to three cases was calculated (with a four in a million probability). Upper bound estimates were calculated to be about 10 times higher. 83 refs.

  14. Accident analysis for transuranic waste management alternatives in the U.S. Department of Energy waste management program

    International Nuclear Information System (INIS)

    Preliminary accident analyses and radiological source term evaluations have been conducted for transuranic waste (TRUW) as part of the US Department of Energy (DOE) effort to manage storage, treatment, and disposal of radioactive wastes at its various sites. The approach to assessing radiological releases from facility accidents was developed in support of the Office of Environmental Management Programmatic Environmental Impact Statement (EM PEIS). The methodology developed in this work is in accordance with the latest DOE guidelines, which consider the spectrum of possible accident scenarios in the implementation of various actions evaluated in an EIS. The radiological releases from potential risk-dominant accidents in storage and treatment facilities considered in the EM PEIS TRUW alternatives are described in this paper. The results show that significant releases can be predicted for only the most severe and extremely improbable accidents sequences

  15. Regional long-term model of radioactivity dispersion and fate in the Northwestern Pacific and adjacent seas: application to the Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    The compartment model POSEIDON-R was modified and applied to the Northwestern Pacific and adjacent seas to simulate the transport and fate of radioactivity in the period 1945–2010, and to perform a radiological assessment on the releases of radioactivity due to the Fukushima Dai-ichi accident for the period 2011–2040. The model predicts the dispersion of radioactivity in the water column and in sediments, the transfer of radionuclides throughout the marine food web, and subsequent doses to humans due to the consumption of marine products. A generic predictive dynamic food-chain model is used instead of the biological concentration factor (BCF) approach. The radionuclide uptake model for fish has as a central feature the accumulation of radionuclides in the target tissue. The three layer structure of the water column makes it possible to describe the vertical structure of radioactivity in deep waters. In total 175 compartments cover the Northwestern Pacific, the East China and Yellow Seas and the East/Japan Sea. The model was validated from 137Cs data for the period 1945–2010. Calculated concentrations of 137Cs in water, bottom sediments and marine organisms in the coastal compartment, before and after the accident, are in close agreement with measurements from the Japanese agencies. The agreement for water is achieved when an additional continuous flux of 3.6 TBq y−1 is used for underground leakage of contaminated water from the Fukushima Dai-ichi NPP, during the three years following the accident. The dynamic food web model predicts that due to the delay of the transfer throughout the food web, the concentration of 137Cs for piscivorous fishes returns to background level only in 2016. For the year 2011, the calculated individual dose rate for Fukushima Prefecture due to consumption of fishery products is 3.6 μSv y−1. Following the Fukushima Dai-ichi accident the collective dose due to ingestion of marine products for Japan increased in 2011 by a factor

  16. Source and long-term behavior of transuranic aerosols in the WIPP environment.

    Science.gov (United States)

    Thakur, P; Lemons, B G

    2016-10-01

    Source and long-term behavior transuranic aerosols ((239+240)Pu, (238)Pu, and (241)Am) in the ambient air samples collected at and near the Waste Isolation Pilot Plant (WIPP) deep geologic repository site were investigated using historical data from an independent monitoring program conducted by the Carlsbad Environmental Monitoring and Research Center and an oversight monitoring program conducted by the management and operating contractor for WIPP at and near the facility. An analysis of historical data indicates frequent detections of (239+240)Pu and (241)Am, whereas (238)Pu is detected infrequently. Peaks in (239+240)Pu and (241)Am concentrations in ambient air generally occur from March to June timeframe, which is when strong and gusty winds in the area frequently give rise to blowing dust. Long-term measurements of plutonium isotopes (1985-2015) in the WIPP environment suggest that the resuspension of previously contaminated soils is likely the primary source of plutonium in the ambient air samples from WIPP and its vicinity. There is no evidence that WIPP is a source of environmental contamination that can be considered significant by any health-based standard. PMID:27394421

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

    International Nuclear Information System (INIS)

    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

  18. Accident Generated Particulate Materials and Their Characteristics -- A Review of Background Information

    Energy Technology Data Exchange (ETDEWEB)

    Sutter, S. L.

    1982-05-01

    Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of the amount of radioactive particulate material initially airborne (source term) during accidents. Pacific Northwest Laboratory (PNL) has surveyed the literature, gathering information on the amount and size of these particles that has been developed from limited experimental work, measurements made from operational accidents, and known aerosol behavior. Information useful for calculating both liquid and powder source terms is compiled in this report. Potential aerosol generating events discussed are spills, resuspension, aerodynamic entrainment, explosions and pressurized releases, comminution, and airborne chemical reactions. A discussion of liquid behavior in sprays, sparging, evaporation, and condensation as applied to accident situations is also included.

  19. User's manual of ART code for analyzing fission product transport behavior during core meltdown accident

    International Nuclear Information System (INIS)

    In a probabilistic risk assessment (PRA) it has been recognized that a core meltdown accident with a large amount of fission products released to the environment is a dominant contributor to public risk. For the evaluation of the risk, information about source terms are inevitable. In order to analyze fission product transport behavior and to evaluate source terms during a core meltdown accident, the ART code has been developed. The ART code has the following features: (1) It can treat fission product transport behavior both in a primary system and a containment system, (2) It models fission product transport caused by both gas flow and liquid flow, and (3) It includes a detailed model about transport behavior of aerosols which are released in quantity during a core meltdown accident. This report is a user's manual for the ART code and includes description of modeling, input/output data and a sample run. (author)

  20. Radioactive material release from a containment vessel during a fire accident

    Energy Technology Data Exchange (ETDEWEB)

    Hensel, S.; Norkus, J.

    2015-02-26

    A methodology is presented to determine the source term for leaks and ruptures of pressurized vessels. The generic methodology is applied to a 9975 Primary Containment Vessel (PCV) which losses containment due to a hypothesized fire accident. The release due to a vessel rupture is approximately two orders of magnitude greater than the release due to a leak.

  1. Operational source term estimation and ensemble prediction for the Grimsvoetn 2011 event

    Science.gov (United States)

    Maurer, Christian; Arnold, Delia; Klonner, Robert; Wotawa, Gerhard

    2014-05-01

    The ESA-funded international project VAST (Volcanic Ash Strategic Initiative Team) includes focusing on a realistic source term estimation in the case of volcanic eruptions as well as on an estimate of the forecast uncertainty in the resulting atmospheric dispersion calculations, which partly derive from the forecast uncertainty in the meteorological input data. SEVIRI earth observation data serve as a basis for the source term estimation, from which the total atmospheric column ash content can be estimated. In an operational environment, the already available EUMETCAST VOLE product may be used. Further an a priori source term is needed, which can be coarsely estimated according to information from previous eruptions and/or constrained with observations of the eruption column. The link between observations and the a priori source is established by runs of the atmospheric transport model FLEXPART for individual emission periods and a predefined number of vertical levels. Through minimizing the differences between observations and model results the so-called a posteriori source term can be depicted for a certain time interval as a function of height. Such a result is shown for a first test case, the eruption of the Grimsvoetn volcano on Iceland in May 2011. Once the dispersion calculations are as optimized as possible with regard to the source term, the uncertainty stemming from the forecast uncertainty of the numeric weather prediction model used is still present, adding up to the unavoidable model errors. Since it is impossible to perform FLEXPART runs for all 50 members of the Integrated Forecasting System (IFS) of ECMWF due to computational (time-storage) constraints, the number of members gets restricted to five (maximum seven) representative runs via cluster analysis. The approach used is as of Klonner (2012) where it was demonstrated that exclusive consideration of the wind components on a pressure level (e.g. 400 hPa) makes it possible to find clusters and

  2. Data assimilation in the process of source term evaluation, radioactive cloud dispersion and impacts modeling

    International Nuclear Information System (INIS)

    Improved 'data assimilation' process is developed for a real time nuclear emergency response system. Our complex approach covers phenomena as follows: - State of reactor fuel and state of radiological barriers between the fuel and the atmosphere of the environment is evaluated according to really measured technological and radiological data. - Database of hypothetical source terms is pre-calculated in advance for a specific nuclear installation and various states of fuel and radiological barriers. Predicted source term is assimilated according to actual knowledge of the state of the fuel and the barriers. - The predicted source term (release by nuclides) is used for modelling on-site radiological impacts, and these calculated modeled impacts are assimilated to radiological quantities measured on-site (monitors are placed around reactor building in the area of nuclear installation). As a result, estimated 'really observed' release is evaluated. - The really observed release (by nuclides, in Bq per time interval) is used as an input to the dispersion models (Puff Trajectory Model or Lagrangean Particle Model) and the calculated offsite impacts are assimilated to radiological quantities measured offsite in close or far vicinity of the point of release. As a result, corrected activities of puffs or particles are calculated, last step of dispersion is recalculated with new inputs and correction of impacts is performed. Innovative aspect of the presented results is in the application of data assimilation process to such extent, which covers the area from the reactor up to the impacts modeled on the terrain of the environment. The described approach of the data assimilation is implemented in real time running system 'ESTE' which is a GIS based instrument used by crisis centers of Slovak NPPs and governmental emergency response centers of various countries. (author)

  3. ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Thompson, Adam B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

  4. A Comprehensive Probabilistic Tsunami Hazard Assessment: Multiple Sources and Short-Term Interactions

    Science.gov (United States)

    Anita, G.; Selva, J.; Laura, S.

    2011-12-01

    We develop a comprehensive and total probabilistic tsunami hazard assessment (TotPTHA), in which many different possible source types concur to the definition of the total tsunami hazard at given target sites. In a multi-hazard and multi-risk perspective, such an innovative approach allows, in principle, to consider all possible tsunamigenic sources, from seismic events, to slides, asteroids, volcanic eruptions, etc. In this respect, we also formally introduce and discuss the treatment of interaction/cascade effects in the TotPTHA analysis. We demonstrate how external triggering events may induce significant temporary variations in the tsunami hazard. Because of this, such effects should always be considered, at least in short-term applications, to obtain unbiased analyses. Finally, we prove the feasibility of the TotPTHA and of the treatment of interaction/cascade effects by applying this methodology to an ideal region with realistic characteristics (Neverland).

  5. 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

    above aquifers used for water supply. The fracture network in aquitards is currently poorly described at larger depths (below 5-8 m) and the effect of sand lenses on leaching behaviour is not well understood. The microbial processes are assumed to be taking place in the fracture system......-dichloroethylene (cis-DCE), vinyl chloride (VC) and ethene. This process can be enhanced by addition of electron donors and/or bioaugmentation and is termed Enhanced Reductive Dechlorination (ERD). This work aims to improve our understanding of the physical, chemical and microbial processes governing source behaviour...... under natural and enhanced conditions. That understanding is applied to risk assessment, and to determine the relationship and time frames of source clean up and plume response. To meet that aim, field and laboratory observations are coupled to state of the art models incorporating new insights...

  6. The long-term oscillations in sunspots and related inter-sunspot sources in microwave emission

    Science.gov (United States)

    Bakunina, I. A.; Abramov-Maximov, V. E.; Smirnova, V. V.

    2016-02-01

    This work presents the microwave long-term oscillations with periods of a few tens of minutes obtained from Nobeyama radioheliograph (NoRH) at frequency 17 GHz. In two active regions the fluctuations of radio emission of different types of intersunspot sources (ISS) (compact and extended) were compared with the fluctuations in magnetic fields of sunspots. Common periods in variations of microwave emission of different type of sources and magnetic field of sunspots were discovered. The delay of 17 minutes was revealed for oscillations of the extended ISS with respect to variations of magnetic field of its tail sunspot. The model of the sunspot magnetic structure based on the concept of three magnetic fluxes for explanation of this fact is discussed.

  7. Analysis of source term modeling for low-level radioactive waste performance assessments

    International Nuclear Information System (INIS)

    Site-specific radiological performance assessments are required for the disposal of low-level radioactive waste (LLW) at both commercial and US Department of Energy facilities. This work explores source term modeling of LLW disposal facilities by using two state-of-the-art computer codes, SOURCEI and SOURCE2. An overview of the performance assessment methodology is presented, and the basic processes modeled in the SOURCE1 and SOURCE2 codes are described. Comparisons are made between the two advective models for a variety of radionuclides, transport parameters, and waste-disposal technologies. These comparisons show that, in general, the zero-order model predicts undecayed cumulative fractions leached that are slightly greater than or equal to those of the first-order model. For long-lived radionuclides, results from the two models eventually reach the same value. By contrast, for short-lived radionuclides, the zero-order model predicts a slightly higher undecayed cumulative fraction leached than does the first-order model. A new methodology, based on sensitivity and uncertainty analyses, is developed for predicting intruder scenarios. This method is demonstrated for 137Cs in a tumulus-type disposal facility. The sensitivity and uncertainty analyses incorporate input-parameter uncertainty into the evaluation of a potential time of intrusion and the remaining radionuclide inventory. Finally, conclusions from this study are presented, and recommendations for continuing work are made

  8. Basic repository source term and data sheet report: Deaf Smith County

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    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 Deaf Smith County, Texas. 2 refs., 6 tabs.

  9. EVALUATION PLAN FOR COMPARATIVE INVESTIGATION OF SOURCE TERM ESTIMATION ALGORITHMS USING FUSION FIELD TRIAL 2007 DATA

    OpenAIRE

    Platt, Nathan; Warner, Steve; Nunes, Steve M.

    2008-01-01

    Abstract: Given a warning based on detection of hazardous materials at just a few sensors, it could be useful to rapidly (minutes) provide an estimate of the source location, time of release, and amount of material released. Such an estimate can lead to refined predictions of the area impacted by the hazardous release, and can support near-term follow-on actions to investigate the cause and nature of the hazardous release. In September 2007, a short-range, highly-instrumented test wa...

  10. PBMR radionuclide source term analysis validation based on AVR operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Stoker, C.C. [PBMR, Lake Buena Vista Building, 1267 Gordon Hood Avenue, Centurion 0046 (South Africa); Olivier, L.D. [Independent Nuclear Consultants, Grahamstown (South Africa); Stassen, E.; Reitsma, F. [PBMR, Lake Buena Vista Building, 1267 Gordon Hood Avenue, Centurion 0046 (South Africa); Merwe, J.J. van der, E-mail: hanno.vdmerwe@pbmr.co.z [PBMR, Lake Buena Vista Building, 1267 Gordon Hood Avenue, Centurion 0046 (South Africa)

    2010-10-15

    The determination of radionuclide source terms is vital for any reactor design and licensing safety evaluation. This paper provides an overview of the PBMR analysis tools with specific focus on the modelling of mobile and deposited radionuclide source terms within the pressure boundary of a typical pebble-bed high temperature reactor (HTR). The main focus is on the Dust and Activity Migration and Distribution (DAMD) software code system that models the activation, migration and time-dependent distribution of dust and atomic particles in an HTR such as the AVR and PBMR. Since DAMD provides a time-dependent systems integrated model of HTR designs, most of the obvious physical phenomena relevant to source terms are at play. These include the neutron flux, activation cross-sections, radioactive decay, dust production rates, dust impurity levels, dust filter capabilities, dust particle size distributions, thermal-hydraulic parameters influencing the migration and distribution of particles throughout the main power system and subsystems, and helium coolant leakage and make-up rates. At this stage the DAMD calibration and validation is mainly based on the operational data, experiments and measurements made during 21 years of operating life of the AVR. The comparisons of the DAMD results with various AVR measurements provide confidence in the use of DAMD for the PBMR design and safety evaluations. In addition, sensitivity analyses are performed with DAMD to determine the bounding system parameters that drive the migration and distribution of radionuclides. The use of DAMD to evaluate design configurations, e.g. the effect of the introduction and placement of filters on the radionuclide distribution, is also shown. In conclusion, the importance of a systems modelling approach for radionuclide transport and distribution within the pressure boundary of a typical HTR system, is demonstrated. Since the DAMD code system is calibrated and validated against the AVR measurements it

  11. The uranium source-term mineralogy and geochemistry at the Broubster natural analogue site, Caithness

    International Nuclear Information System (INIS)

    The British Geological Survey (BGS) has been conducting a coordinated research programme at the Broubster natural analogue site in Caithness, north Scotland. This work on a natural radioactive geochemical system has been carried out with the aim of improving our confidence in using predictive models of radionuclide migration in the geosphere. This report is one of a series being produced and it concentrates on the mineralogical characterization of the uranium distribution in the limestone unit considered as the 'source-term' in the natural analogue model

  12. Detection and analysis of black spots with even small accident figures. Contribution to the Seminar on Short-term and Area-wide Evaluation of Safety Measures, Amsterdam, April 19-21, 1982, p. 75-84.

    NARCIS (Netherlands)

    Oppe, S.

    1982-01-01

    In order to detect accident black spots we have to know the probability of an accident for a traffic situation of some kind, or the mean number of accidents for some unit of time. In almost all procedures known to us, the various road locations are treated as isolated spots. With small accident figu

  13. Quantifying sediment source contributions in coastal catchments impacted by the Fukushima nuclear accident with carbon and nitrogen elemental concentrations and stable isotope ratios

    Science.gov (United States)

    Laceby, J. Patrick; Huon Huon, Sylvain; Onda, Yuichi; Evrard, Olivier

    2016-04-01

    The Fukushima Dai-ichi Nuclear Power Plant accidental release of radioactive contaminants resulted in the significant fallout of radiocesium over several coastal catchments in the Fukushima Prefecture. Radiocesium, considered to be the greatest risk to the short and long term health of the local community, is rapidly bound to fine soil particles and thus is mobilized and transported during soil erosion and runoff processes. As there has been a broad-scale decontamination of rice paddy fields and rural residential areas in the contaminated region, one important long term question is whether there is, or may be, a downstream transfer of radiocesium from forests that covered over 65% of the most contaminated region. Accordingly, carbon and nitrogen elemental concentrations and stable isotope ratios are used to determine the relative contributions of forests and rice paddies to transported sediment in three contaminated coastal catchments. Samples were taken from the three main identified sources: cultivated soils (rice paddies and fields, n=30), forest soils (n=45), and subsoils (channel bank and decontaminated soils, n = 25). Lag deposit sediment samples were obtained from five sampling campaigns that targeted the main hydrological events from October 2011 to October 2014. In total, 86 samples of deposited sediment were analyzed for particulate organic matter elemental concentrations and isotope ratios, 24 from the Mano catchment, 44 from the Niida catchment, and 18 from the Ota catchment. Mann-Whitney U-tests were used to examine the source discrimination potential of this tracing suite and select the appropriate tracers for modelling. The discriminant tracers were modelled with a concentration-dependent distribution mixing model. Preliminary results indicate that cultivated sources (predominantly rice paddies) contribute disproportionately more sediment per unit area than forested regions in these contaminated catchments. Future research will examine if there are

  14. CARNSORE: Hypothetical reactor accident study

    International Nuclear Information System (INIS)

    Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

  15. Serious accident in Peru

    International Nuclear Information System (INIS)

    A peruvian man, victim of an important accidental irradiation arrived on the Saturday twenty ninth of may 1999 to the centre of treatment of serious burns at the Percy military hospital (Clamart -France). The accident spent on the twentieth of February 1999, on the site of a hydroelectric power plant, in construction at 300 km at the East of Lima. The victim has picked up an industrial source of iridium devoted to gamma-graphy operations and put it in his back pocket; of trousers. The workman has serious radiation burns. (N.C.)

  16. Current severe accident research facilities and projects

    International Nuclear Information System (INIS)

    The Working Group on the Analysis and Management of Accidents (GAMA) is mainly composed of technical specialists in the areas of coolant system thermal-hydraulics, in-vessel protection, containment protection, and fission product retention. Its general functions include the exchange of information on national and international activities in these areas, the exchange of detailed technical information, and the discussion of progress achieved in respect of specific technical issues. Severe accident management is one of the important tasks of the group. This document is an update of the 'Current Severe Accident Research Facilities and Projects' list. Facilities and projects are sorted according to the following criteria: In-Vessel Phenomena: Core Degradation and Melt Progression, Molten Core Debris Interaction with the Reactor Pressure Vessel Lower Head and Mechanical Behaviour of Reactor Pressure Vessel Lower Head; In-Vessel and Ex-Vessel Molten Fuel/Coolant Interactions; Ex-Vessel Phenomena: Molten Core Debris/Concrete Interactions, Molten Core/Ceramic Interaction, Melt Release (including DCH), Melt Spreading and Catching Devices Studies, Melt Coolability, Corium Melt properties; Hydrogen Transport and Combustion: Mixing and Distribution, Deflagration, Deflagration-to-Detonation Transition, Passive Recombiner Performance; Mechanical Behaviour of Reactor Pressure Vessel Lower Head; Containment Structural Integrity: Containment Failure Experiment and Analysis, Material Properties and Structural Behaviour, Containment Thermal-Hydraulics, Containment Cooling, Cable Penetration Integrity; Fission Products and Aerosols: Effects of Specific Elements on Iodine Volatility, Release of Low-Volatility Fission Products/Late In-Vessel Fission Product Release, Reactor Materials Release, Aerosol and Iodine Behaviour in Reactor Coolant System and Containment, Retention, Resuspension and Revaporization in Primary Circuit, Aerosol Nucleation and Transport, Source Term, Containment

  17. Study on safety evaluation for nuclear fuel cycle facility under fire accident conditions

    International Nuclear Information System (INIS)

    Hot test at Rokkasho Reprocessing plant has been started since last year. In addition, construction of the MOX fuel fabrication facility at Rokkasho site is planning. So, the importance of safety evaluation of the nuclear fuel cycle facility is increasing. Under the fire accident, one of the serious postulated accidents in the nuclear fuel cycle facility, the equipments (glove-box, ventilation system, ventilation filters etc.) for the confinement of the radioactive materials within the facility could be damaged by a large amount of heat and smoke released from the combustion source. Therefore, the fundamental data and models calculating for the amount of heat and smoke released from the combustion source under such accident are important for the safety evaluation of the facility. In JAERI, the study focused on the evaluation of amount of heat and smoke released from the combustion source is planning. In this paper, the outline of experimental apparatus, measurement items and evaluation terms are described. (author)

  18. The Role of Countermeasures in Mitigating the Radiological Consequences of Nuclear Power Plant Accidents

    International Nuclear Information System (INIS)

    During the Fukushima accident the mitigation actions played an important role to decrease the consequences of the accident. The countermeasures are the actions that should be taken after the occurrence of a nuclear accident to protect the public against the associated risk. The actions may be represented by sheltering, evacuation, distribution of stable iodine tablets and/or relocation. This study represents a comprehensive probabilistic study to investigate the role of the adoption of the countermeasures in case of a hypothetical accident of type LOCA for a nuclear power plant of PWR (1000 Mw) type. This work was achieved through running of the PC COSYMA(1) code. The effective doses in different organs, short and long term health effects, and the associated risks were calculated with and without countermeasures. In addition, the overall costs of the accident and the costs of countermeasures are estimated which represent our first trials to know how much the postulated accident costs. The source term of a hypothetical accident is determined by knowing the activity of the core inventory. The meteorological conditions around the site in addition to the population distribution were utilized as input parameters. The stability conditions and the height of atmospheric boundary layers ABL of the concerned site were determined by developing a computer program utilizing Pasquill-Gifford atmospheric stability conditions. The results showed that, the area around the site requires early and late countermeasures actions after the accident especially in the downwind sectors. For late countermeasures, the duration of relocation ranged from about two to 10 years. The adoption of the countermeasures increases the costs of emergency planning by 40% but reduces the risk associated with the accident. (author)

  19. Regional long-term model of radioactivity dispersion and fate in the Northwestern Pacific and adjacent seas: application to the Fukushima Dai-ichi accident.

    Science.gov (United States)

    Maderich, V; Bezhenar, R; Heling, R; de With, G; Jung, K T; Myoung, J G; Cho, Y-K; Qiao, F; Robertson, L

    2014-05-01

    The compartment model POSEIDON-R was modified and applied to the Northwestern Pacific and adjacent seas to simulate the transport and fate of radioactivity in the period 1945-2010, and to perform a radiological assessment on the releases of radioactivity due to the Fukushima Dai-ichi accident for the period 2011-2040. The model predicts the dispersion of radioactivity in the water column and in sediments, the transfer of radionuclides throughout the marine food web, and subsequent doses to humans due to the consumption of marine products. A generic predictive dynamic food-chain model is used instead of the biological concentration factor (BCF) approach. The radionuclide uptake model for fish has as a central feature the accumulation of radionuclides in the target tissue. The three layer structure of the water column makes it possible to describe the vertical structure of radioactivity in deep waters. In total 175 compartments cover the Northwestern Pacific, the East China and Yellow Seas and the East/Japan Sea. The model was validated from (137)Cs data for the period 1945-2010. Calculated concentrations of (137)Cs in water, bottom sediments and marine organisms in the coastal compartment, before and after the accident, are in close agreement with measurements from the Japanese agencies. The agreement for water is achieved when an additional continuous flux of 3.6 TBq y(-1) is used for underground leakage of contaminated water from the Fukushima Dai-ichi NPP, during the three years following the accident. The dynamic food web model predicts that due to the delay of the transfer throughout the food web, the concentration of (137)Cs for piscivorous fishes returns to background level only in 2016. For the year 2011, the calculated individual dose rate for Fukushima Prefecture due to consumption of fishery products is 3.6 μSv y(-1). Following the Fukushima Dai-ichi accident the collective dose due to ingestion of marine products for Japan increased in 2011 by a

  20. IAEA-MEL case studies. Pt.1. Source terms and transport processes

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, M.S. [International Atomic Energy Agency, Monaco (Monaco). Marine Environmental Laboratory

    1997-12-31

    This IAEA review summarises the nuclide inventory and process-related aspects of investigations at sites which include disposals and dispersions of radioactive waste in the Arctic Seas, N.E. Atlantic Ocean, Far Eastern Seas and the Irish Sea, accidental inputs to the Norwegian Sea and Sea of Okhotsk, releases from weapons test environments in the Pacific Ocean and fallout from the Chernobyl accident, particularly in the Baltic and Black Seas. In many of these cases, monitoring, behavioural and modelling studies have been carried out. The Monaco laboratory is using natural and man-made radionuclides to trace particle flux and sedimentation, particularly in S.E. Asia and the Mediterranean Sea. A perspective on the results of these studies has been maintained in a new and growing marine radioactivity database which will soon be available to Member States. Radiological conclusions can be compared to the recently published results of a coordinated research programme organised by IAEA-MEL on sources of radioactivity in the marine environment and their relative contributions to overall dose assessment from marine radioactivity (the MARDOS project). Finally, some future plans within the IAEA-MEL programme are outlined. For example, the Government of Japan is funding a strategically planned follow-up to the Geosecs programme to update information on open ocean radionuclide distributions. Additional IAEA funds have also been allocated for study of marine inputs of technologically enhanced natural radionuclides from the oil and phosphogypsum industries. The laboratory is also developing and applying in-situ monitoring techniques for continuous and ROV-mounted surveys of radionuclide inventories in water and sediment. The presentation will end with a call for participation in an IAEA Symposium on Marine Pollution, containing various sessions related to tracer and radiological aspects of marine radioactivity, to be held in Monaco in 1998 as part of the United Nations programme

  1. Level 3 PSA and it's implementation for PWR accident

    International Nuclear Information System (INIS)

    Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA) is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidents of PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The series of calculations performed are: calculate the source terms of the core damaged, modeling of meteorological conditions and environmental site, exposure pathway modeling, analysis of radionuclide dispersion and transport phenomena in the environment, radionuclide deposition analysis, analysis of radiation dose, protection & mitigation analysis, and risk analysis. The assessment uses a series of subsystems on PC Cosyma software. The results prove that the safety assessment using Level 3 PSA methodology is very effective and comprehensive estimate the impact, consequences, risks, nuclear emergency preparedness, and the reactor accident management especially for severe accidents or beyond design basis accidents of nuclear power plants. The results of the assessment can be used as a feedback to safety assessment of Level 1 PSA and Level 2 PSA. (author)

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

    International Nuclear Information System (INIS)

    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

  3. Reference Spent Fuel and Its Source Terms for a Design of Deep Geological Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

    In this study, current status and future trend of domestic spent fuels were analyzed to propose reference spent nuclear fuel. And then, source terms needed for design of a deep geological disposal system were calculated using ORIGEN-ARP. The reference spent fuels selected based on assembly physical dimension, inventory projection, trend of initial enrichment of 235U, discharge burnup are as follows; The 17x17 Korean Optimized Fuel Assembly with initial enrichment of 4.0 wt.% 235U and discharge burnup of 45 GWD/MTU was adopted as a low-burnup representative fuel. For the high-burnup representative fuel, 16x16 Korean Standard Fuel Assembly with initial enrichment of 4.5 wt.% 235U and discharge burnup of 55 GWD/MTU was chosen. CANDU fuel with initial enrichment of 0.711 wt.% 235U and discharge burnup of 7.5 GWD/MTU was also considered. For these reference fuels, decay heat, radiation intensity and spectrum, nuclide concentration, and individual nuclide radioactivity were calculated using ORIGEN-ARP for a disposal system design. It is expected that the source terms estimated in this study will be applied to the disposal system development in the future.

  4. 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)

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

    International Nuclear Information System (INIS)

    The source term, isolation capacity, and long-term radiological exposure of 14C from the Swedish underground repository for low and intermediate level waste (SFR) is assessed. The prospective amount of 14C in the repository is assumed to be 5 TBq. Spent ion exchange resins will be the dominant source of 14C. The pore water in the concrete repository is expected to maintain a pH of >10.5 for a period of at least 106 y. The cement matrix of the repository will retain most of the 14CO32- initially present. Bacterial production of CO2 and CH4 from degradation of ion-exchange resins and bitumen may contribute to 14C release to the biosphere. However, CH4 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 2.4x104 years. A total leakage of 8.4 GBq of 14C from the repository will cause a total collective dose commitment of 1.1 manSv or 130 manSv TBq-1. (authors)

  6. 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)

    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

  7. Development of severe accident management advisory and training simulator (SAMAT)

    International Nuclear Information System (INIS)

    The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management

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

    Energy Technology Data Exchange (ETDEWEB)

    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 /sup 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 /sup 238/PuO/sub 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 /sup 238/PuO/sub 2/ aerosol due to coagulation with an aerosol of clay or sandy loam soil.

  9. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Following the accident at Chernobyl nuclear reactor, WHO organized on 6 May 1986 in Copenhagen a one day consultation of experts with knowledge in the fields of meteorology, radiation protection, biological effects, reactor technology, emergency procedures, public health and psychology in order to analyse the development of events and their consequences and to provide guidance as to the needs for immediate public health action. The present report provides detailed information on the transportation and dispersion of the radioactive material in the atmosphere, especially volatile elements, during the release period 26 April - 5 May. Presented are the calculated directions and locations of the radioactive plume over Europe in the first 5 days after the accident, submitted by the Swedish Meteorological and Hydrological Institute. The calculations have been made for two heights, 1500m and 750m and the plume directions are grouped into five periods, covering five European areas. The consequences of the accident inside the USSR and the radiological consequences outside the USSR are presented including the exposure routes and the biological effects, paying particular attention to iodine-131 effects. Summarized are the first reported measured exposure rates above background, iodine-131 deposition and concentrations in milk and the remedial actions taken in various European countries. Concerning the cesium-137 problem, based on the UNSCEAR assessment of the consequences of the nuclear fallout, one concludes that the cesium contamination outside the USSR is not likely to cause any serious problems. Finally, the conclusions and the recommendations of the meeting, taking into account both the short-term and longer term considerations are presented

  10. 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

  11. 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.

  12. Some issues on the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors Amendment after JCO criticality accident

    International Nuclear Information System (INIS)

    As the Amendment of the Law for the Regulation of Nuclear Material, Nuclear Fuel Material and Reactors on an opportunity of the JCO criticality accident can be almost evaluated at a viewpoint of upgrading on effectiveness of safety regulation, it is thought to remain a large problem to rely on only enforcement of regulation due to amendment of the Law at future accident. In future, it can be also said to be important subjects to further expand a philosophy on the regulation (material regulation) focussed to hazards of nuclear material itself, not only to secure effectiveness on the multi-complementary safety regulation due to the administrative agency and the Nuclear Safety Commission but also to prepare a mechanism reflexible of a new information to the safety regulation, and to prepare a mechanism to assist adequate business execution and so forth of enterprises. (G.K.)

  13. Microbial characterization for the Source-Term Waste Test Program (STTP) at Los Alamos

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, P.A.; Strietelmeier, B.A.; Pansoy-Hjelvik, M.E.; Villarreal, R.

    1999-04-01

    The effects of microbial activity on the performance of the proposed underground nuclear waste repository, the Waste Isolation Pilot Plant (WIPP) at Carlsbad, New Mexico are being studied at Los Alamos National Laboratory (LANL) as part of an ex situ large-scale experiment. Actual actinide-containing waste is being used to predict the effect of potential brine inundation in the repository in the distant future. The study conditions are meant to simulate what might exist should the underground repository be flooded hundreds of years after closure as a result of inadvertent drilling into brine pockets below the repository. The Department of Energy (DOE) selected LANL to conduct the Actinide Source-Term Waste Test Program (STTP) to confirm the predictive capability of computer models being developed at Sandia National Laboratory.

  14. Microbial characterization for the Source-Term Waste Test Program (STTP) at Los Alamos

    International Nuclear Information System (INIS)

    The effects of microbial activity on the performance of the proposed underground nuclear waste repository, the Waste Isolation Pilot Plant (WIPP) at Carlsbad, New Mexico are being studied at Los Alamos National Laboratory (LANL) as part of an ex situ large-scale experiment. Actual actinide-containing waste is being used to predict the effect of potential brine inundation in the repository in the distant future. The study conditions are meant to simulate what might exist should the underground repository be flooded hundreds of years after closure as a result of inadvertent drilling into brine pockets below the repository. The Department of Energy (DOE) selected LANL to conduct the Actinide Source-Term Waste Test Program (STTP) to confirm the predictive capability of computer models being developed at Sandia National Laboratory

  15. Use of a parametric source term estimating code in level 2/3 PSAs

    International Nuclear Information System (INIS)

    Use of the XSOR methodology for establishing fission releases, provides a convenient and simple means of evaluating source terms, allowing for rapid evaluations of a large number of release classes. It offers a simple means of addressing core melt phenomenology and plant defense strategies that are not modeled in codes like MAAP and MELCOR and for accounting for and quantifying phenomenological uncertainties without relying on large number of complex deterministic analyses. This paper describes the application of the XSOR methodology to both BWR and PWR plants, the improvements to the US Reactor Risk Study XSOR models made by ABB, and compares the results of representative XSOR analyses with a similar uncertainty analysis of a MAAP 3.0B simulation

  16. Improved treatment of source terms in TVD scheme for shallow water equations

    Science.gov (United States)

    Tseng, Ming-Hseng

    2004-06-01

    A number of high-resolution schemes have been recently developed to solve the homogeneous form of the shallow water equations. However, most approximate Riemann solvers experience difficulties with natural river applications if the irregular bed topography is not handled correctly. Based on the finite-difference flux-limited total variation diminishing (TVD) scheme, this paper develops a simple approach to handle the source terms for the one-dimensional open channel flow simulation with rapidly varying bed topography. Conclusions on the validity of the operator-splitting approach, the eigenvector-projection approach, and the proposed approach are presented. Analytical solution, experimental data, and available numerical result comparisons are shown to demonstrate the accuracy, robustness, stability, simplicity, and applicability of the proposed model.

  17. Long-term Periodicity Analysis of Polarization Variation for Radio Sources

    Indian Academy of Sciences (India)

    Yuhai Yuan

    2011-03-01

    We use the database of University of Michigan Radio Astronomy Observatory (UMRAO) at three radio bands (4.8, 8 and 14.5 GHz) to analyse the long-term polarization variation in search of the possible periodicity. Using the power spectral analysis method (PSA), the Jurkevich method and the discrete correlation function (DCF) method, we find that there are 16 sources lying in periodicity. The results show the astrophysically meaningful periodicity covering 2.1 years to 16.2 years at 4.8 GHz, 2.8 years to 16.3 years at 8 GHz, and 1.8 years to 16.6 years at 14.5 GHz.

  18. Lessons Learned Through the Follow-up of the Long-Term Effects of Over-Exposure to an Ir192 Industrial Radiography Source in Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, A.; Rabbani, G.; Hossain, M. K.; Alam, M. K.; Koddus, A.

    2003-02-24

    An industrial radiographer was accidentally over-exposed while taking the radiograph of weld-joints of gas pipe-lines in 1985 in Bangladesh. Symptoms of high radiation exposure occurred immediately after the accident and skin erythema developed leading to progressive tissue deterioration. The consequences of this over-exposure is being followed up to assess the long-term effects of ionizing radiation on the victim. Progressive tissue deteriorations have already led to multiple surgeries and successive amputations of the finger-tips so far. Lessons learned from this accident are also reported in this paper.

  19. Nuclear law and radiological accidents

    International Nuclear Information System (INIS)

    Nuclear activities in Brazil, and particularly the radiological accident of Goiania, are examined in the light of the environmental and nuclear laws of Brazil and the issue of responsibility. The absence of legislation covering radioactive wastes as well as the restrictions on Brazilian States to issue regulations covering nuclear activities are reviewed. The radiological accident and its consequences, including the protection and compensation of the victims, the responsibility of the shareholders of the Instituto Goiano de Radioterapia, operator of the radioactive source, the provisional storage and the final disposal at Abadia de Goias of the radioactive waste generated by the accident are reviewed. Finally, nuclear responsibility, the inapplicability of the Law 6453/77 which deals with nuclear damages, and the state liability regime are analysed in accordance with the principles of the Brazilian Federal Constitution. (author)

  20. Use of open source software in estimating the effects of a severe accident on the Mark II containment; Uso de software de fuente abierta en la estimacion de los efectos de un accidente severo sobre la contencion Mark II

    Energy Technology Data Exchange (ETDEWEB)

    Sainz, E.; Arguelles, R., E-mail: eduardo.sainz@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    Because the spectrum of scenarios of severe accident before which must verify the integrity of the containment can be very broad, it arises here a calculation methodology to estimate the structural response of the containment without incurring in high costs for commercial software licenses, or in times and calculation excessive requirements. The capabilities of computer programs with license of open source, OpenFOAM for CFD calculations and Salome-Meca for thermal and mechanical calculations were tested. The methodology begins of the venting of mass and energy that are postulated inside the container and the values of the thermal and mechanical fields are obtained through the walls. (Author)

  1. Long-term aerosol measurements in Gran Canaria, Canary Islands: Particle concentration, sources and elemental composition

    Science.gov (United States)

    Gelado-Caballero, MaríA. D.; López-GarcíA, Patricia; Prieto, Sandra; Patey, Matthew D.; Collado, Cayetano; HéRnáNdez-Brito, José J.

    2012-02-01

    There are very few sets of long-term measurements of aerosol concentrations over the North Atlantic Ocean, yet such data is invaluable in quantifying atmospheric dust inputs to this ocean region. We present an 8-year record of total suspended particles (TSP) collected at three stations on Gran Canaria Island, Spain (Taliarte at sea level, Tafira 269 m above sea level (a.s.l.) and Pico de la Gorra 1930 m a.s.l.). Using wet and dry deposition measurements, the mean dust flux was calculated at 42.3 mg m-2 d-1. Air mass back trajectories (HYSPLIT, NOAA) suggested that the Sahara desert is the major source of African dust (dominant during 32-50% of days), while the Sahel desert was the major source only 2-10% of the time (maximum in summer). Elemental composition ratios of African samples indicate that, despite the homogeneity of the dust in collected samples, some signatures of the bedrocks can still be detected. Differences were found for the Sahel, Central Sahara and North of Sahara regions in Ti/Al, Mg/Al and Ca/Al ratios, respectively. Elements often associated with pollution (Pb, Cd, Ni, Zn) appeared to share a common origin, while Cu may have a predominantly local source, as suggested by a decrease in the enrichment factor (EF) of Cu during dust events. The inter-annual variability of dust concentrations is investigated in this work. During winter, African dust concentration measurements at the Pico de la Gorra station were found to correlate with the North Atlantic Oscillation (NAO) index.

  2. Atmospheric moisture transports to the Arctic from different reanalyses: comparative assessment and analysis of source terms

    Science.gov (United States)

    Dufour, Ambroise; Zolina, Olga; Gulev, Sergey

    2014-05-01

    Accurate knowledge of the Arctic heat and moisture balances is critically important for understanding mechanisms of polar climate change and the observed amplification of the Arctic warming. Basic characteristics of the atmosphere in the Arctic region have quite a large spread in the modern era and first generation reanalyses, thus preventing effective use of reanalyses for the assessment of atmospheric moisture and heat transports and analysis of variability in the source terms. We used Eulerian approach to derive and intercompare to each other estimates of the moisture transports in the atmosphere from 5 reanalyses (ERA-Interim, MERRA, NCEP-CFSR, JRA-25, NCEP-1). Computational procedure involved decomposition of the velocity and moisture fields into mean conditions and variations around the mean. This concept allowed for the further association of the mean and eddy transports with large scale circulation modes (mean component) and synoptic transients (eddy component). The latter was associated with the characteristics of cyclone activity derived from the same reanalyses using state of the art numerical algorithm for cyclone identification and tracking. Atmospheric moisture transport is most intense over the GIN Sea and the North European basin, however over this area of the most intense transports, the contributions from the eddy and mean transport components are not correlated hinting on different pattern of variability in moisture fluxes due to cyclone activity and mean circulation. Decadal scale variability in the atmospheric moisture transports has been further associated with the Arctic-scale and regional differences between local precipitation and evaporation as well as with the magnitude of the storage terms. Potential mechanisms of variability in these terms are discussed.

  3. To improve nuclear plant safety by learning from accident's experience

    International Nuclear Information System (INIS)

    The ultimate goal of this study is to produce an expert system that enables the experience (records and information) gained from accidents to be put to use towards improving nuclear plant safety. A number of examples have been investigated, both domestic and overseas, in which experience gained from accidents was utilized by utilities in managing and operating their nuclear power stations to improve safety. The result of investigation has been used to create a general 'basic flow' to make the best use of experience. The ultimate goal is achieved by carrying out this 'basic flow' with artificial intelligence (AI). To do this, it is necessary (1) to apply language analysis to process the source information (primary data base; domestic and overseas accident's reports) into the secondary data base, and (2) to establish an expert system for selecting (screening) significant events from the secondary data base. In the processing described in item (1), a multi-lingual thesaurus for nuclear-related terms become necessary because the source information (primary data bases) itself is multi-lingual. In the work described in item (2), the utilization of probabilistic safety assessment (PSA), for example, is a candidate method for judging the significance of events. Achieving the goal thus requires developing various new techniques. As the first step of the above long-term study project, this report proposes the 'basic flow' and presents the concept of how the nuclear-related AI can be used to carry out this 'basic flow'. (author)

  4. WASTE-ACC: A computer model for analysis of waste management accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nabelssi, B.K.; Folga, S.; Kohout, E.J.; Mueller, C.J.; Roglans-Ribas, J.

    1996-12-01

    In support of the U.S. Department of Energy`s (DOE`s) Waste Management Programmatic Environmental Impact Statement, Argonne National Laboratory has developed WASTE-ACC, a computational framework and integrated PC-based database system, to assess atmospheric releases from facility accidents. WASTE-ACC facilitates the many calculations for the accident analyses necessitated by the numerous combinations of waste types, waste management process technologies, facility locations, and site consolidation strategies in the waste management alternatives across the DOE complex. WASTE-ACC is a comprehensive tool that can effectively test future DOE waste management alternatives and assumptions. The computational framework can access several relational databases to calculate atmospheric releases. The databases contain throughput volumes, waste profiles, treatment process parameters, and accident data such as frequencies of initiators, conditional probabilities of subsequent events, and source term release parameters of the various waste forms under accident stresses. This report describes the computational framework and supporting databases used to conduct accident analyses and to develop source terms to assess potential health impacts that may affect on-site workers and off-site members of the public under various DOE waste management alternatives.

  5. CFD Simulation on the Main Steam Line Break Accident with the ATLAS (SLB-GB-02T) in Terms of Thermal Mixing and Asymmetry Effects

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sanggyu; Kim, Hangon; Park, Youngsheop [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MSLB accident was initiated by a double-ended guillotine break at one of Steam Generator (SG). The MSLB accident is characterized as an increase in heat removal by the secondary system. This causes excessive heat removal from reactor coolant system and decrease in reactor coolant temperatures. As a result, the core reactivity is increased by the negative moderator and Doppler reactivity coefficients. The viewpoint of MSLB analysis is to know whether the thermal asymmetry effect is sustained on the core or not when the cold water flows from the affected SG to the affected RCS loop and the hot water flows from the intact SG to the intact RCS loop. KHNP CRI, as a participant of the group B, performs the CFD analysis to analyze the thermal mixing and asymmetry effect on the downcomer and the core using the ANSYS CFX version 14.57 code. In general, CFD code still has a limitation on the application of two-phase phenomena, but the applicability and accuracy in the single phase flow condition is validated by many researchers in recent years. Based on the experiment, the primary fluid is maintained under subcooled liquid phase without the vapor phase, and thereupon CFD code can be applicable in the simulation of ATLAS vessel flow. In this paper, the CFD analysis model is bounded by the ATLAS vessel, and the RCS loop flow rate and temperature is treated as a boundary condition using the experiment data. Firstly, a steady-state analysis is conducted and this result is analyzed as below. - Major fluid behavior such as pressure, temperature, flow rate is compared with experiment data - a axial pressure distribution in the downcomer and the core Based on the steady-state results, a transient calculation is performed and the boundary conditions are determined by the experimental data. 2 cases are calculated as below - Case1: Period from the break initiation time to the reactor trip time by low SG pressure (LSGP) signal. This period represents that the temperature difference

  6. CFD Simulation on the Main Steam Line Break Accident with the ATLAS (SLB-GB-02T) in Terms of Thermal Mixing and Asymmetry Effects

    International Nuclear Information System (INIS)

    The MSLB accident was initiated by a double-ended guillotine break at one of Steam Generator (SG). The MSLB accident is characterized as an increase in heat removal by the secondary system. This causes excessive heat removal from reactor coolant system and decrease in reactor coolant temperatures. As a result, the core reactivity is increased by the negative moderator and Doppler reactivity coefficients. The viewpoint of MSLB analysis is to know whether the thermal asymmetry effect is sustained on the core or not when the cold water flows from the affected SG to the affected RCS loop and the hot water flows from the intact SG to the intact RCS loop. KHNP CRI, as a participant of the group B, performs the CFD analysis to analyze the thermal mixing and asymmetry effect on the downcomer and the core using the ANSYS CFX version 14.57 code. In general, CFD code still has a limitation on the application of two-phase phenomena, but the applicability and accuracy in the single phase flow condition is validated by many researchers in recent years. Based on the experiment, the primary fluid is maintained under subcooled liquid phase without the vapor phase, and thereupon CFD code can be applicable in the simulation of ATLAS vessel flow. In this paper, the CFD analysis model is bounded by the ATLAS vessel, and the RCS loop flow rate and temperature is treated as a boundary condition using the experiment data. Firstly, a steady-state analysis is conducted and this result is analyzed as below. - Major fluid behavior such as pressure, temperature, flow rate is compared with experiment data - a axial pressure distribution in the downcomer and the core Based on the steady-state results, a transient calculation is performed and the boundary conditions are determined by the experimental data. 2 cases are calculated as below - Case1: Period from the break initiation time to the reactor trip time by low SG pressure (LSGP) signal. This period represents that the temperature difference

  7. Classical and quantum parts in Madelung variables: Splitting the source term of the Einstein equation into classical and quantum parts

    Directory of Open Access Journals (Sweden)

    Biró T.S.

    2014-01-01

    Full Text Available Postulating a particular quantum correction to the source term in the classical Einstein equation we identify the conformal content of the above action and obtain classical gravitation for massive particles, but with a cosmological term representing off-mass-shell contribution to the energy-momentum tensor.

  8. Ranking of severe accident research priorities

    Energy Technology Data Exchange (ETDEWEB)

    Schwinges, B. [Gesell Anlagen and Reaktorsicherheit GRS mbH, D-50667 Cologne (Germany); Journeau, C. [CEA Cadarache, DEN STRI LMA, F-13115 St Paul Les Durance (France); Haste, T. [Paul Scherrer Inst, NES LTH, OVGA 312, CH-5232 Villigen (Switzerland); Meyer, L.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Trambauer, K. [GRS mbH, Forschungsgelande, D-85748 Garching (Germany)

    2010-07-01

    The objectives of the SARNET network are to define common research programmes in the field of severe accidents and to develop common computer tools and methodologies for safety assessment in this field. To reach these objectives, one of the work packages, named 'Severe Accident Research Priorities' (SARP), aimed at reviewing and reassessing the priorities of research issues as a basis to harmonize and to re-orient research programmes, to define new ones, and to close - if possible - resolved issues on a common basis. The work was performed in close collaboration with 8 participating institutions, led by GRS, representing technical safety organisations, industry and utilities (IRSN, CEA, EDF, FZK, GRS, KTH, TUS, VTT). This action made use notably of (1) the outcomes of the EURSAFE project in the 5. Framework Programme, i. e. the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the validation and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other SARNET tasks, and (4) the outcome of the research performed in the three thematic sub-domains of SARNET (corium, containment and source term). The main outcome of EURSAFE was a list of 21 topics which included recommendations for experimental programmes and code developments. This list formed the basis of the work in SARP. Also the methodology applied in EURSAFE to consider both the risk potential and the severe accident issues where large uncertainties still subsist was adopted. The analyses of the progress of research and development activities considered whether (1) any research issue was resolved due to reduction of uncertainties or gain of scientific insights, (2) any new issue had to be added to the list of needed research, (3) any new process or phenomenon had to be included in the general PIRT list taking into account the safety relevance and the lack of knowledge, and (4) any new accident management program has to be

  9. Consequences of severe nuclear accidents in Europe

    Science.gov (United States)

    Seibert, Petra; Arnold, Delia; Mraz, Gabriele; Arnold, Nikolaus; Gufler, Klaus; Kromp-Kolb, Helga; Kromp, Wolfgang; Sutter, Philipp

    2013-04-01

    A first part of the presentation is devoted to the consequences of the severe accident in the 1986 Chernobyl NPP. It lead to a substantial radioactive contaminated of large parts of Europe and thus raised the awareness for off-site nuclear accident consequences. Spatial patterns of the (transient) contamination of the air and (persistent) contamination of the ground were studied by both measurements and model simulations. For a variety of reasons, ground contamination measurements have variability at a range of spatial scales. Results will be reviewed and discussed. Model simulations, including inverse modelling, have shown that the standard source term as defined in the ATMES study (1990) needs to be updated. Sensitive measurements of airborne activities still reveal the presence of low levels of airborne radiocaesium over the northern hemisphere which stems from resuspension. Over time scales of months and years, the distribution of radionuclides in the Earth system is constantly changing, for example relocated within plants, between plants and soil, in the soil, and into water bodies. Motivated by the permanent risk of transboundary impacts from potential major nuclear accidents, the multidisciplinary project flexRISK (see http://flexRISK.boku.ac.at) has been carried out from 2009 to 2012 in Austria to quantify such risks and hazards. An overview of methods and results of flexRISK is given as a second part of the presentation. For each of the 228 NPPs, severe accidents were identified together with relevant inventories, release fractions, and release frequencies. Then, Europe-wide dispersion and dose calculations were performed for 2788 cases, using the Lagrangian particle model FLEXPART. Maps of single-case results as well as various aggregated risk parameters were produced. It was found that substantial consequences (intervention measures) are possible for distances up to 500-1000 km, and occur more frequently for a distance range up to 100-300 km, which is in

  10. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996; TOPICAL

    International Nuclear Information System (INIS)

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration

  11. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration.

  12. An overview of the last 20 years of radiological accidents with high over-exposure and cases of fatalities - P1.040

    International Nuclear Information System (INIS)

    This document discusses the main accidents that have happened in the last two decades, in terms of causes, consequences, similarities and lessons learned when sealed sources have been damaged, lost, stolen and abandoned. In considerable majority death and serious injuries were resulted from failures in the safety system for radiation sources and for the security of radioactive materials. (author)

  13. Summary of intensive monitoring for radionuclides in fishery products after Fukushima accident and comparison to the results of long term monitoring program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Ken; Morita, Takami; Shigenobu, Yuya; Takagi, Kaori; Miki, Shizuho; Kaeriyama, Hideki; Ambe, Daisuke; Ono, Tsuneo [National Research Institute of Fisheries Science, Fisheries Research Agency, 2-12-4, Fukuura, Kanazawa, Yokohama, Kanagawa 236-8648 (Japan)

    2014-07-01

    Monitoring of artificial radionuclides in fishery products started in late 1950's and has over half a century history in Japan. Fisheries Agency (FA), Fisheries Research Agency (FRA) and prefectural organizations have been conducting the monitoring. The intensive monitoring of radioactive material in fishery products started in the late of March 2011 immediately after the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident for the coastal area faced to the Pacific Ocean in the eastern area of Japan and for the offshore area. The purpose of this monitoring is to avoid the distribution of fishery products containing radioactive cesium in concentration over the Japanese standard limit (Cs-134+Cs-137: 100 Bq/kg). Japanese monitoring data are opened to public on the FA's and FRA's web page. In this study, we resume the results of intensive monitoring started in 2011 and show the temporal change of concentration of radioactive cesium in fishery product. In Fukushima Prefecture, the ratio of samples exceeding the Japanese standard limit shows a steadily decreasing trend from 53% at Mar.- Jun. 2011 to below 3% at Jul. - Sep. 2013. In other prefectures, the ratio was 6.5% at Mar.- Jun. 2011 and fell gradually, and it has been below to 1% at Jul. - Sep. 2013. In presentation, we show the results of monitoring conducted in the coastal area and deeper water around Japan from long gamma ray measurements with ashed samples by high purified germanium gamma spectrometry. (authors)

  14. 1976 Hanford americium accident

    Energy Technology Data Exchange (ETDEWEB)

    Heid, K R; Breitenstein, B D; Palmer, H E; McMurray, B J; Wald, N

    1979-01-01

    This report presents the 2.5-year medical course of a 64-year-old Hanford nuclear chemical operator who was involved in an accident in an americium recovery facility in August 1976. He was heavily externally contaminated with americium, sustained a substantial internal deposition of this isotope, and was burned with concentrated nitric acid and injured by flying debris about the face and neck. The medical care given the patient, including the decontamination efforts and clinical laboratory studies, are discussed. In-vivo measurements were used to estimate the dose rates and the accumulated doses to body organs. Urinary and fecal excreta were collected and analyzed for americium content. Interpretation of these data was complicated by the fact that the intake resulted both from inhalation and from solubilization of the americium embedded in facial tissues. A total of 1100 ..mu..Ci was excreted in urine and feces during the first 2 years following the accident. The long-term use of diethylenetriaminepentate (DTPA), used principally as the zinc salt, is discussed including the method, route of administration, and effectiveness. To date, the patient has apparently experienced no complications attributable to this extensive course of therapy, even though he has been given approximately 560 grams of DTPA. 4 figures, 1 table.

  15. Planning for the Handling of Radiation Accidents

    International Nuclear Information System (INIS)

    The developing atomic energy programmes and the widespread use of radiation sources in medicine, agriculture, industry and research have had admirable safety records. Throughout the world the number of known accidents in which persons have been exposed to harmful am ounts of ionizing radiation is relatively small, and only a few deaths have occurred. Meticulous precautions are being taken to maintain this good record in all work with radiation sources and to keep the exposure of persons as low as practicable. In spite of all the precautions that are taken, accidents may occur and they may be accompanied by the injury or death of persons and damage to property. It is only prudent to take those steps that are practicable to prevent accidents and to plan in advance the emergency action that would limit the injuries and damage caused by those accidents that do occur. Emergency plans should be sufficiently broad to cover unforeseen or very improbable accidents as well as those that are considered credible. Some accidents may involve only the workers in an establishment, those working directly with the source and possibly their colleagues. Other accidents may have consequences, notably in the form of radioactive contamination of the environment, that affect the general public, possibly far from the site of the accident. The preparation of plans for dealing with radiation accidents is therefore obligatory both for the various authorities that are responsible for protecting the health and the food and water supplies of the public, and for the operator of an installation containing radiation sources.

  16. Long-term fluctuations of hailstorms in South Moravia, Czech Republic: synthesis of different data sources

    Science.gov (United States)

    Chromá, Kateřina; Brázdil, Rudolf; Dolák, Lukáš; Řezníčková, Ladislava; Valášek, Hubert; Zahradníček, Pavel

    2016-04-01

    Hailstorms belong to natural phenomena causing great material damage in present time, similarly as it was in the past. In Moravia (eastern part of the Czech Republic), systematic meteorological observations started generally in the latter half of the 19th century. Therefore, in order to create long-term series of hailstorms, it is necessary to search for other sources of information. Different types of documentary evidence are used in historical climatology, such as annals, chronicles, diaries, private letters, newspapers etc. Besides them, institutional documentary evidence of economic and administrative character (e.g. taxation records) has particular importance. This study aims to create a long-term series of hailstorms in South Moravia using various types of documentary evidence (such as taxation records, family archives, chronicles and newspapers which are the most important) and systematic meteorological observations in the station network. Although available hailstorm data cover the 1541-2014 period, incomplete documentary evidence allows reasonable analysis of fluctuations in hailstorm frequency only since the 1770s. The series compiled from documentary data and systematic meteorological observations is used to identify periods of lower and higher hailstorm frequency. Existing data may be used also for the study of spatial hailstorm variability. Basic uncertainties of compiled hailstorm series are discussed. Despite some bias in hailstorm data, South-Moravian hailstorm series significantly extends our knowledge about this phenomenon in the south-eastern part of the Czech Republic. The study is a part of the research project "Hydrometeorological extremes in Southern Moravia derived from documentary evidence" supported by the Grant Agency of the Czech Republic, reg. no. 13-19831S.

  17. The effect of work accidents on the efficiency of production in the coal sector

    Directory of Open Access Journals (Sweden)

    Yaşar Kasap

    2011-05-01

    Full Text Available In comparison with other sectors, mining is one of the sectors with the highest rates of work accidents. Such accidents negatively affect a country’s economy by wasting domestic resources and causing losses of both labour force and working days. What distinguishes mining from other branches of industry is that its working environments change continually and the working conditions are particularly harsh. Because of the practice of labour-intensive underground production methods, which leads to an increase in risk factors in terms of work accidents, and the fact that coal is a leading resource in meeting the ever-increasing demand for energy, this study investigated how work accidents affected the efficiency of production in the Turkish Hard Coal Enterprise (TTK between 1987 and 2006. Using data envelopment analysis, the overall sources of technical inefficiency in the years examined were determined. The results from this analysis revealed that the overall technical efficiency was as low as 69.7%, particularly as a result of the disaster in 1992; work accidents therefore had a negative effect on production efficiency. The greatest degree of pure technical inefficiency was found to have occurred in the period between 1992 and 2000, when the highest number of work accidents were noted, whilst the greatest degree of scale inefficiency was found to have occurred between 1987 and 1993. Because TTK has a prominent position among institutions and attaches great importance to workers’ health and safety, an increase was noted in efficiency scores after 1993.

  18. Improvement of the following accident dose assessment system (II)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Enn Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae; Choi, Young Gil [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-12-01

    The FADAS and its database have been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and does not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively. 7 refs., 8 figs., 6 tabs. (Author)

  19. Improvement of the following accident dose assessment system

    International Nuclear Information System (INIS)

    The FADAS has been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and dose not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively

  20. Improvement of the following accident dose assessment system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Enn Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae; Choi, Young Gil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-12-15

    The FADAS has been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and dose not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively.

  1. JCO criticality accident termination operation

    International Nuclear Information System (INIS)

    In 2001, we summarized the circumstances surrounding termination of the JCO criticality accident based on testimony in the Mito District Court on December 17, 2001. JCO was the company for uranium fuels production in Japan. That document was assembled based on actual testimony in the belief that a description of the work involved in termination of the accident would be useful in some way for preventing nuclear disasters in the future. The description focuses on the witness' own behavior, and what he saw and heard, and thus is written from the perspective of action by one individual. This was done simply because it was easier for the witness to write down his memories as he remembers them. Description of the activities of other organizations and people is provided only as necessary, to ensure that consistency in the descriptive approach is not lost. The essentials of this report were rewritten as a third-person objective description in the summary of the report by the Atomic Energy Society of Japan (AESJ). Since then, comments have been received from sources such as former members of the Nuclear Safety Commission (Dr. Kenji Sumita and Dr. Akira Kanagawa), concerned parties from the former Science and Technology Agency, and reports from the JCO Criticality Accident Investigation Committee of the AESJ, and thus this report was rewritten to correct incorrect information, and add material where that was felt to be necessary. This year is the tenth year of the JCO criticality accident. To mark this occasion we have decided to translate the record of what occurred at the accident site into English so that more people can draw lessons from this accident. This report is an English version of JAEA-Technology 2009-073. (author)

  2. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production missions at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines

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

    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); Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    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.

  4. High order finite difference methods with subcell resolution for advection equations with stiff source terms

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Wei [Deprartment of Mathematics. Florida Intl Univ., Miami, FL (United States); Shu, Chi-Wang [Division of Applied Mathematics. Brown Univ., Providence, RI (United States); Yee, H.C. [NASA Ames Research Center (ARC), Moffett Field, Mountain View, CA (United States); Sjögreen, Björn [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2012-01-01

    A new high order finite-difference method utilizing the idea of Harten ENO subcell resolution method is proposed for chemical reactive flows and combustion. In reaction problems, when the reaction time scale is very small, e.g., orders of magnitude smaller than the fluid dynamics time scales, the governing equations will become very stiff. Wrong propagation speed of discontinuity may occur due to the underresolved numerical solution in both space and time. The present proposed method is a modified fractional step method which solves the convection step and reaction step separately. In the convection step, any high order shock-capturing method can be used. In the reaction step, an ODE solver is applied but with the computed flow variables in the shock region modified by the Harten subcell resolution idea. For numerical experiments, a fifth-order finite-difference WENO scheme and its anti-diffusion WENO variant are considered. A wide range of 1D and 2D scalar and Euler system test cases are investigated. Studies indicate that for the considered test cases, the new method maintains high order accuracy in space for smooth flows, and for stiff source terms with discontinuities, it can capture the correct propagation speed of discontinuities in very coarse meshes with reasonable CFL numbers.

  5. Implementation of a source term control program in a mature boiling water reactor.

    Science.gov (United States)

    Vargo, G J; Jarvis, A J; Remark, J F

    1991-06-01

    The implementation and results of a source term control program implemented at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients. PMID:2032839

  6. Long-term Performance of Moderate Heat Portland Cement with Double-expansive Sources

    Institute of Scientific and Technical Information of China (English)

    YE Qing; CHEN Huxing; KONG Deyu; WANG Shangxian; LOU Zonghan

    2007-01-01

    The long-term performance of moderate heat Portland cement with double-expansive sources (DE cement) in the system of high MgO clinker and gypsum was studied by XRD, SEM/EDAX and test methods for strength and expansion of cement. Results indicate that the periclase particle, whose size was 5-7.5 μm in DE cement clinker containing 4.8 % MgO, existed individually. The periclase hydration in hardened DE cement paste started at about 60 days and completed up to 2 000 days, and ettringite in the paste was stable from 3 days to 2 000 days. Under the conditions of 4.5%-5.0 % MgO in clinker and 2.8%-3.4 %SO3 in cement,ettringite expansion and brucite expansion in DE cement paste had a continuity, entirety and stability. At the ages of 90, 365, 730 and 2 000 days the expansion of the paste reached 0.07%-0.11%, 0.16%-0.21%, 0.21%-0.27% and 0.29%-0.38 %, respectively. The results suggest that by using this cement in mass concrete it may compensate its temperature shrinkage and autogenous shrinkage to some extent.

  7. Implementation of a source term control program in a mature boiling water reactor

    International Nuclear Information System (INIS)

    The implementation and results of a source term control program at the James A. FitzPatrick Nuclear Power Plant (JAF), a mature boiling water reactor (BWR) facility that has been in commercial operation since 1975, are discussed. Following a chemical decontamination of the reactor water recirculation piping in the Reload 8/Cycle 9 refueling outage in 1988, hydrogen water chemistry (HWC) and feedwater Zn addition were implemented. This is the first application of both HWC and feedwater Zn addition in a BWR facility. The radiological benefits and impacts of combined operation of HWC and feedwater Zn addition at JAF during Cycle 9 are detailed and summarized. The implementation of hydrogen water chemistry resulted in a significant transport of corrosion products within the reactor coolant system that was greater than anticipated. Feedwater Zn addition appears to be effective in controlling buildup of other activated corrosion products such as 60Co on reactor water recirculation piping; however, adverse impacts were encountered. The major adverse impact of feedwater Zn addition is the production of 65Zn that is released during plant outages and operational transients

  8. On the convergence rate of operator splitting for Hamilton-Jacobi equations with source terms

    Energy Technology Data Exchange (ETDEWEB)

    Jakobsen, Espen R.; Karlsen, Kenneth H.; Risebro, Nils Henrik

    2000-02-01

    We establish a rate of convergence for a semi-discrete operator splitting method applied to Hamilton-Jacobi equations with source terms. The method is based on sequentially solving a Hamilton-Jacobi equation and an ordinary differential equation. The Hamilton-Jacobi equation is solved exactly while the ordinary differential equation is solved exactly or by an explicit Euler method. We prove that the L{sup {infinity}} error associated with the operator splitting method is bounded by O({delta}t), where {delta}t is the splitting (or time) step. This error bound is an improvement over the existing O((sqroot)({delta}t)) bound due to Souganidis [40]. In the one dimensional case, we present a fully discrete splitting method based on an unconditionally stable front tracking method for homogenuous Hamilton-Jacobi equations. It is proved that this fully discrete splitting method possesses a linear convergence rate. Moreover, numerical results are presented to illustrate the theoreticle convergence results. (author)

  9. The CALIBRE source-term code: Technical documentation for version 2

    Energy Technology Data Exchange (ETDEWEB)

    Worgan, K.J.; Robinson, P.C. [Intera Information Technologies, Henley-on-Thames, Oxfordshire (United Kingdom)

    1995-03-01

    The CALIBRE source term model has been updated, to make it a computationally more robust and faster code, and to incorporate some new features. These include the addition of a canister pinhole release model and a `tunnel` option. The tunnel is represented by a zero concentration boundary condition at the interface of the backfilled tunnel with the bentonite and rock. The redox front model has been incorporated into the radionuclide transport code, so that a redox calculations may be run alone or prior to a radionuclide migration calculation. Redox front results can also be saved and re-used for radionuclide transport. More generally, the code simulates the diffusion of radionuclides from a high-level waste canister, through a backfill region and into a fractured rock matrix. The model includes chain decay and ingrowth, linear equilibrium sorption, solubility limiting and response to a redox front as it emerges from the canister and migrates through the near-field. Radial advection (which approximates the advection downstream from the canister and buffer) is applied in the fracture, in addi- tion to diffusion. There is also an option to allow for fixed total concentrations (i e concentrations sorbed on the solid and dissolved in the pore water) of naturally occurring isotopes in the fractured rock. This document describes the mathematical model and numerical methods used in developing CALIBRE, together with a number of verification tests which compare the results with those computed using analytic solutions. 11 refs, tabs.

  10. 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)

  11. Supervisor's accident investigation handbook

    International Nuclear Information System (INIS)

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

  12. Radiological accidents potentially important to human health risk in the U.S. Department of Energy waste management program

    International Nuclear Information System (INIS)

    Human health risks as a consequence of potential radiological releases resulting from plausible accident scenarios constitute an important consideration in the US Department of Energy (DOE) national program to manage the treatment, storage, and disposal of wastes. As part of this program, the Office of Environmental Management (EM) is currently preparing a Programmatic Environmental Impact Statement (PEIS) that evaluates the risks that could result from managing five different waste types. This paper (1) briefly reviews the overall approach used to assess process and facility accidents for the EM PEIS; (2) summarizes the key inventory, storage, and treatment characteristics of the various DOE waste types important to the selection of accidents; (3) discusses in detail the key assumptions in modeling risk-dominant accidents; and (4) relates comparative source term results and sensitivities

  13. Nuclear industry after the Fukushima accident

    International Nuclear Information System (INIS)

    This special dossier about the situation of nuclear industry two years after the Fukushima accident comprises 15 contributions dealing with: the nuclear industry two years after the Fukushima accident (Bernard Salha); a low-carbon electricity at a reasonable cost (Christophe Behar); nuclear engineering has to gain even more efficiency (Thomas Branche); how to dispose off the most radioactive wastes (Marie-Claude Dupuis, Thibaud Labalette); ensuring the continuation for more than 40 years onward (Denis Gasquet); developing and investing in the future (Philippe Knoche); more than just signing contracts (Dominique Lagarde); immersed power plants, an innovative concept (Bernard Planchais); R and D as a source of innovation for safety and performances (Jean-Pierre West); dismantlement, a very long term market (Jerome Stubler, Bruno Lancia); a reference industrial model (Herve Machenaud); recruiting and training (Andre Einaudi); a diversity of modern reactors and a world market in rebirth (Philippe Anglaret); an industrial revolution is necessary (Yves Brachet); contracts adapted to sensible works (Philippe Bonnave)

  14. Structural and containment response to LMFBR accidents

    International Nuclear Information System (INIS)

    The adequacy of the containment of fast reactors has been traditionally evaluated by analyzing the response of the containment to a spectrum of core disruptive accidents. The current approach in the U.S. is to consider fast reactor response to accidents in terms of four lines of assurance (LOAs). Thus, LOA-1 is to prevent accidents, LOA-2 is to limit core damage, LOA-3 is to control accident progression and LOA-4 is to attenuate radiological consequences. Thus, the programs on the adequacy of containment response fall into LOA-3. Significant programs to evaluate the response of the containment to core disruptive accidents and, thereby, to assure control of accident progression are in progress. These include evaluating the mechanical response of the primary system to core disruptive accidents and evaluating the thermal response of the reactor structures to core melting, including the effects this causes on the secondary containment. The analysis of structural response employs calculated pressure-volume-time loading functions. The results of the analyses establish the response of the containment to the prescribed loadings. The analysis of thermal response requires an assessment of the distribution and state of the fuel, fission products and activated materials from accident initiation to final disposition in a stable configuration

  15. Framework for accident management

    International Nuclear Information System (INIS)

    Accident management is an essential element of the Nuclear Regulatory Commission (NRC) Integration Plan for the closure of severe accident issues. This element will consolidate the results from other key elements; such as the Individual Plant Examination (IPE), the Containment Performance Improvement, and the Severe Accident Research Programs, in a form that can be used to enhance the safety programs for nuclear power plants. The NRC is currently conducting an Accident Management Program that is intended to aid in defining the scope and attributes of an accident management program for nuclear power plants. The accident management plan will ensure that a plant specific program is developed and implemented to promote the most effective use of available utility resources (people and hardware) to prevent and mitigate severe accidents. Hardware changes or other plant modifications to reduce the frequency of severe accidents are not a central aim of this program. To accomplish the outlined objectives, the NRC has developed an accident management framework that is comprised of five elements: (1) accident management strategies, (2) training, (3) guidance and computational aids, (4) instrumentation, and (5) delineation of decision making responsibilities. A process for the development of an accident management program has been identified using these NRC framework elements

  16. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  17. Efficient construction of high-resolution TVD conservative schemes for equations with source terms: application to shallow water flows

    Science.gov (United States)

    Burguete, J.; García-Navarro, P.

    2001-09-01

    High-resolution total variation diminishing (TVD) schemes are widely used for the numerical approximation of hyperbolic conservation laws. Their extension to equations with source terms involving spatial derivatives is not obvious. In this work, efficient ways of constructing conservative schemes from the conservative, non-conservative or characteristic form of the equations are described in detail. An upwind, as opposed to a pointwise, treatment of the source terms is adopted here, and a new technique is proposed in which source terms are included in the flux limiter functions to get a complete second-order compact scheme. A new correction to fix the entropy problem is also presented and a robust treatment of the boundary conditions according to the discretization used is stated. Copyright

  18. Balancing source terms and flux gradients in high-resolution Godunov methods: The quasi-steady wave-propagation algorithm

    Energy Technology Data Exchange (ETDEWEB)

    LeVeque, R.J. [Univ. of Washington, Seattle, WA (United States)

    1998-10-10

    Conservation laws with source terms often have steady states in which the flux gradients are nonzero but exactly balanced by source terms. Many numerical methods (e.g., fractional step methods) have difficulty preserving such steady states and cannot accurately calculate small perturbations of such states. Here a variant of the wave-propagation algorithm is developed which addresses this problem by introducing a Riemann problem in the center of each grid cell whose flux difference exactly cancels the source term. This leads to modified Riemann problems at the cell edges in which the jump now corresponds to perturbations from the steady state. Computing waves and limiters based on the solution to these Riemann problems gives high-resolution results. The 1D and 2D shallow water equations for flow over arbitrary bottom topography are used as an example, though the ideas apply to many other systems. The method is easily implemented in the software package CLAWPACK.

  19. Radiation Protection Aspects of Primary Water Chemistry and Source-term Management Report

    International Nuclear Information System (INIS)

    Since the beginning of the 1990's, occupational exposures in nuclear power plant has strongly decreased, outlining efforts achieved by worldwide nuclear operators in order to reach and maintain occupational exposure as low as reasonably achievable (ALARA) in accordance with international recommendations and national regulations. These efforts have focused on both technical and organisational aspects. According to many radiation protection experts, one of the key features to reach this goal is the management of the primary system water chemistry and the ability to avoid dissemination of radioactivity within the system. It outlines the importance for radiation protection staff to work closely with chemistry staff (as well as operation staff) and thus to have sufficient knowledge to understand the links between chemistry and the generation of radiation field. This report was prepared with the primary objective to provide such knowledge to 'non-chemist'. The publication primarily focuses on three topics dealing with water chemistry, source term management and remediation techniques. One key objective of the report is to provide current knowledge regarding these topics and to address clearly related radiation protection issues. In that mind, the report prepared by the EGWC was also reviewed by radiation protection experts. In order to address various designs, PWRs, VVERs, PHWRs and BWRs are addressed within the document. Additionally, available information addressing current operating units and lessons learnt is outlined with choices that have been made for the design of new plants. Chapter 3 of this report addresses current practices regarding primary chemistry management for different designs, 'how to limit activity in the primary circuit and to minimise contamination'. General information is provided regarding activation, corrosion and transport of activated materials in the primary circuit (background on radiation field generation). Primary chemistry aspects that

  20. Particle generation methods applied in large-scale experiments on aerosol behaviour and source term studies

    International Nuclear Information System (INIS)

    In aerosol research aerosols of known size, shape, and density are highly desirable because most aerosols properties depend strongly on particle size. However, such constant and reproducible generation of those aerosol particles whose size and concentration can be easily controlled, can be achieved only in laboratory-scale tests. In large scale experiments, different generation methods for various elements and compounds have been applied. This work presents, in a brief from, a review of applications of these methods used in large scale experiments on aerosol behaviour and source term. Description of generation method and generated aerosol transport conditions is followed by properties of obtained aerosol, aerosol instrumentation used, and the scheme of aerosol generation system-wherever it was available. An information concerning aerosol generation particular purposes and reference number(s) is given at the end of a particular case. These methods reviewed are: evaporation-condensation, using a furnace heating and using a plasma torch; atomization of liquid, using compressed air nebulizers, ultrasonic nebulizers and atomization of liquid suspension; and dispersion of powders. Among the projects included in this worked are: ACE, LACE, GE Experiments, EPRI Experiments, LACE-Spain. UKAEA Experiments, BNWL Experiments, ORNL Experiments, MARVIKEN, SPARTA and DEMONA. The aim chemical compounds studied are: Ba, Cs, CsOH, CsI, Ni, Cr, NaI, TeO2, UO2Al2O3, Al2SiO5, B2O3, Cd, CdO, Fe2O3, MnO, SiO2, AgO, SnO2, Te, U3O8, BaO, CsCl, CsNO3, Urania, RuO2, TiO2, Al(OH)3, BaSO4, Eu2O3 and Sn. (Author)

  1. Learning lessons from Natech accidents - the eNATECH accident database

    Science.gov (United States)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

    When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

  2. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA) organized in June 1994, in collaboration with the Swedish Nuclear Power Inspectorate (SKI), a Specialist Meeting on Selected Containment Severe Accident Management Strategies, to discuss their feasibility, effectiveness, benefits and drawbacks, and long-term impact. The meeting focused on water reactors, mainly on existing systems. The technical content covered topics such as general aspects of accident management strategies in OECD Member countries, hydrogen management techniques and other containment accident management strategies, surveillance and protection of the containment function. The main conclusions of the meeting are summarized in the paper. (author)

  3. On monotonicity, stability, and construction of central schemes for hyperbolic conservation laws with source terms (Revised Version)

    OpenAIRE

    Borisov, V. S.; Mond, M.

    2007-01-01

    The monotonicity and stability of difference schemes for, in general, hyperbolic systems of conservation laws with source terms are studied. The basic approach is to investigate the stability and monotonicity of a non-linear scheme in terms of its corresponding scheme in variations. Such an approach leads to application of the stability theory for linear equation systems to establish stability of the corresponding non-linear scheme. The main methodological innovation is the theorems establish...

  4. Safety against releases in severe accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I.; Berg, Oe.; Nonboel, E. [eds.

    1997-12-01

    The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au) 39 refs.

  5. Safety against releases in severe accidents. Final report

    International Nuclear Information System (INIS)

    The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au)

  6. The Influence of Seasonal Characteristics on the Accident Consequences Analysis

    International Nuclear Information System (INIS)

    In order to examine the influence of seasonal characteristics on accident consequences, we defined a limited number of basic spectra based on the relative importance of source term release parameters and meteorological conditions on offsite health effects and economic impacts. We then investigated the variation in numbers and frequency of early health effects and economic impacts resulting from the severe accidents of the YGN 3 and 4 nuclear power plants from spectrum to spectrum by using MACCS code. These investigations were for meteorological conditions defined as typical on an annual basis. Also, we investigated the variation in numbers and frequency of early health effects and economic impacts for the same standard spectra for meteorological conditions defined as typical on a seasonal basis recognizing that there are four seasons with distinct meteorological characteristics. Results show that there are large differences in consequences from spectrum to spectrum, although an equal amount and mix of radioactive material is released to the atmosphere in each case. Therefore, release parameters and meteorological data have to be well characterized in order to estimate accident consequences resulting from an accident accurately. Also, there are large differences in the estimated number of health effects and economic impacts from season to season due to distinct seasonal variations in meteorological conditions in Korea. In fall, the early fatalities and early fatality risk show minimum values due to enhanced dispersion arising from increased atmospheric instability, and the early fatalities show maximum value in summer due to a large rainfall rate. On the contrast, the economic cost shows maximum value in fall and minimum in summer due to different atmospheric dispersion and rainfall rate. Therefore, it is necessary to consider seasonal characteristics in developing emergency response strategies for reducing offsite early health risks in the event of a severe

  7. Framework for accident management

    International Nuclear Information System (INIS)

    A program is being conducted to establish those attributes of a severe accident management plan which are necessary to assure effective response to all credible severe accidents and to develop guidance for their incorporation in a plant's Accident Management Plan. This program is one part of the Accident Management Research Program being conducted by the U. S. Nuclear Regulatory Commission (NRC). The approach used in establishing attributes and developing guidance includes three steps. In the first step the general attributes of an accident management plan were identified based on: (1) the objectives established for the NRC accident management program, (2) the elements of an accident management framework identified by the NRC, and (3) a review of the processes used in developing the currently used approach for classifying and analyzing accidents. For the second step, a process was defined that uses the general attributes identified from the first step to develop an accident management plan. The third step applied the process defined in the second step at a nuclear power plant to refine and develop it into a benchmark accident management plan. Step one is completed, step two is underway and step three has not yet begun

  8. Medical consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Galstyan I.A.

    2015-12-01

    Full Text Available Aim: to study the long-term effects of acute radiation syndrome (ARS, developed at the victims of the Chernobyl accident. Material and Methods. 237 people were exposed during the accident, 134 of them were diagnosed with ARS. Dynamic observation implies a thorough annual examination in a hospital. Results. In the first 1.5-2 years after the ARS mean group indices of peripheral blood have returned to normal. However, many patients had transient expressed moderate cytopenias. Granulocytopenia, thrombocytopenia, lymphopenia and erythropenia were the most frequently observed things during the first 5 years after the accident. After 5 years their occurences lowered. In 11 patients the radiation cataract was detected. A threshold dose for its development is a dose of 3.2 Gy Long-term effects of local radiation lesions (LRL range from mild skin figure smoothing to a distinct fibrous scarring, contractures, persistently recurrent late radiation ulcers. During all years of observation we found 8 solid tumors, including 2 thyroid cancers. 5 hematologic diseases were found. During 29 years 26 ARS survivors died of various causes. Conclusion. The health of ones with long-term ARS effects is determined by the evolution of the LRL effects on skin, radiation cataracts, hema-tological diseases and the accession of of various somatic diseases, not caused by radiation.

  9. National registration of accidents in Iceland.

    Science.gov (United States)

    Olafsson, O; Axelsson, J

    1992-01-01

    Community based registration of accidents has been employed in Iceland from 1987. A form developed in the emergency ward at the city Hospital of Reykjavik has been used for the registration. The following issues have been registered: the type and the seriousness of the injury, treatment, place of accident and time of accident. Health centres in Iceland have been computerized from 1976. At the time being about half of the health centres participate in the registration with the information included in the form as the source. Every health center has its well defined district. The accidents among the inhabitants in each district is registered, while accidents among other people, e.g. tourists, is registered separately. At this moment 183,000 out of a total number of 259,000 inhabitants are covered by the registration, i.e. 71% of the population. In 1989 the frequency of accidents was 198 per 100,000 inhabitants. 26% of the accidents occurred at home, 11% at work, 9% during physical activity, 6% was traffic accidents, whereas the same proportion occurred at school. This registration system has been created as a result of annual conferences on accidents arranged by the Director General of public health since 1984. Representatives for the following parties have been invited; medical doctors working in hospitals and health centres, clinical nurses, physiotherapists, the National Insurance Service, other insurance companies, rescue and ambulance personal, fire departments, the Automobile Association, the communication Council. Local communities members of the parliament, voluntary organizations, e.g. Red Cross, the Sea Rescue Service and the Aviation Board. This activity has stimulated measures aiming at preventing accidents in several local communities.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:1285816

  10. ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chatelard, P., E-mail: patrick.chatelard@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Reinke, N.; Arndt, S. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Eckel, J. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Jacq, F.; Marchetto, C.; Mun, C.; Piar, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France)

    2014-06-01

    The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants.

  11. Quantitative assessment of seismic source performance: Feasibility of small and affordable seismic sources for long term monitoring at the Ketzin CO2 storage site, Germany

    Science.gov (United States)

    Sopher, Daniel; Juhlin, Christopher; Huang, Fei; Ivandic, Monika; Lueth, Stefan

    2014-08-01

    We apply a range of quantitative pre-stack analysis techniques to assess the feasibility of using smaller and cheaper seismic sources, than those currently used at the Ketzin CO2 storage site. Results from two smaller land sources are presented alongside those from a larger, more powerful source, typically utilized for seismic acquisition at the Ketzin. The geological target for the study is the Triassic Stuttgart Formation which contains a saline aquifer currently used for CO2 storage. The reservoir lies at a depth of approximately 630 m, equivalent to a travel time of 500 ms along the study profile. The three sources discussed in the study are the Vibsist 3000, Vibsist 500 (using industrial hydraulic driven concrete breaking hammers) and a drop hammer source. Data were collected for the comparison using the three sources in 2011, 2012 and 2013 along a 984 m long line with 24 m receiver spacing and 12 m shot spacing. Initially a quantitative analysis is performed of the noise levels between the 3 surveys. The raw shot gathers are then analyzed quantitatively to investigate the relative energy output, signal to noise ratio, penetration depth, repeatability and frequency content for the different sources. The performance of the sources is also assessed based on stacked seismic sections. Based on the results from this study it appears that both of the smaller sources are capable of producing good images of the target reservoir and can both be considered suitable as lower cost, less invasive sources for use at the Ketzin site or other shallow CO2 storage projects. Finally, the results from the various pre-stack analysis techniques are discussed in terms of how representative they are of the final stacked sections.

  12. Accidents in radiotherapy: Lack of quality assurance?

    International Nuclear Information System (INIS)

    About 150 radiological accidents, involving more than 3000 patients with adverse effects, 15 patient's fatalities and about 5000 staff and public exposures have been collected and analysed. Out of 67 analysed accidents in external beam therapy 22% has been caused by wrong calculation of the exposure time or monitor units, 13% by inadequate review of patient's chart, 12% by mistakes in the anatomical area to be treated. The remaining 35% can be attributed to 17 different causes. The most common mistakes in brachytherapy were wrong activities of sources used for treatment (20%), inadequate procedures for placement of sources applicators (14%), mistakes in calculating the treatment time (12%), etc. The direct and contributing causes of radiological accidents have been deduced from each event, when it was possible and categorized into 9 categories: mistakes in procedures (30%), professional mistakes (17%), communication mistakes (15%), lack of training (8.5%), interpretation mistakes (7%), lack of supervision (6%), mistakes in judgement (6%), hardware failures (5%), software and other mistakes (5.5%). Three types of direct and contributing causes responsible for almost 62% of all accidents are directly connected to the quality assurance of treatment. The lessons learnt from the accidents are related to frequencies of direct and contributing factors and show that most of the accident are caused by lack, non-application of quality assurance (QA) procedures or by underestimating of QA procedures. The international system for collection of accidents and dissemination of lessons learnt from the different accidents, proposed by IAEA, can contribute to better practice in many radiotherapy departments. Most of the accidents could have been avoided, had a comprehensive QA programme been established and properly applied in all radiotherapy departments, whatever the size. (author)

  13. An overview of severe accident modeling and analysis work for the ANS reactor conceptual safety analysis report

    International Nuclear Information System (INIS)

    ORNL's Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management has initiated severe accident analysis and related technology development efforts early-on in the design phase itself. Early consideration of severe accident issues will aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It will also provide a means for satisfying on- and off-site regulatory requirements and provide containment response and source term analyses for level-2 and -3 Probabilistic Risk Analyses (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions, and consequently provide insights for the development of strategies and design philosophies for accident management, mitigation, and emergency preparedness. This paper presents a perspective overview of the severe accident modeling and analysis work for the ANS Conceptual Safety Analysis Report (CSAR)

  14. Radiation accidents and defence of population

    International Nuclear Information System (INIS)

    Full text: Development of nuclear physics, the fundamental and the applied researches in the field of radioactive insured wide possibility for application of radionuclides and ionizing radiation source in the different fields of national economy. Application of radionuclides in chemical, metallurgical, food industry, in agriculture and etc. Fields provide a large economic profit. It's hard to apprise significance of ionizing radiation source using in medicine for diagnostics and treatment of different disease. Nuclear power engineering and nuclear industry are developing intensively. At same time nuclear power, ionizing radiation sources incur potential treat for surroundings and health of population. As even that stage of protective measure development: there is no possibility of that happening of radiation accidents. A radiation accident qualifies as loss of ionizing radiation sources direction, which provoked by disrepair equipment, natural calamity or other causes which could bring to unplanned irradiation of population or radioactive pollution of surroundings. At present some following typical cases connected with radiation accident have been chosen: Contentious using or keeping of ionizing radiation source with breach of established requires; Loss, theft of ionizing radiation sources or radiation plants, instruments; Leaving the sources of ionizing radiation in the holes; Refusal radiation technic exploited in industry, medicine, SRI and etc; Disrepair in nuclear transport means of conveyance; Crashes and accidents at NPP and at other enterprises of nuclear industry. The radiation accidents according to character, degree and scales have been divided into two groups: Radiation accidents not connected with NPP; Accidents in the nuclear engineering and industry; The radiation accidents not connected with NPP according their consequence divide into 5 groups; accidents which do not come to irradiation of personal, persons from population (more PN-permissible norm

  15. Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

    International Nuclear Information System (INIS)

    This guide was prepared to provide the experienced safety analyst with accident analysis guidance in greater detail than is possible in Department of Energy (DOE) Orders. The guide addresses analysis of postulated serious accidents considered in the siting and selection of major design features of DOE nuclear facilities. Its scope has been limited to radiological accidents at nonreactor nuclear facilities. The analysis steps addressed in the guide lead to evaluation of radiological dose to exposed persons for comparison with siting guideline doses. Other possible consequences considered are environmental contamination, population dose, and public health effects. Choices of models and parameters leading to estimation of source terms, release fractions, reduction and removal factors, dispersion and dose factors are discussed. Although requirements for risk analysis have not been established, risk estimates are finding increased use in siting of major nuclear facilities, and are discussed in the guide. 3 figs., 9 tabs

  16. Final safety analysis report for the Galileo Mission: Volume 2, Book 2: Accident model document: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-15

    This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

  17. Preliminary Analysis of Radiation Shielding for HIC Transport Package Under the Hypothetical Accident Conditions

    International Nuclear Information System (INIS)

    A radiation shielding analysis under the hypothetical accident condition has been conducted using a computer program MCNP5 for a B-type HIC (High Integrated Container) Transport Package, which contains HIC with radioactive waste or spent resin, for transportation from nuclear power plat sites to disposal repository. Radiation source term is first carefully determined from the safety analysis reports related to HIC for appropriate calculation. And then MCNP5 is performed to obtain the minimum crevice between package lid and body, which meets the dose rate limit under the hypothetical accident conditions. Standards and codes of radiation shielding analysis related to the hypothetical accident condition are prescribed in Korea Nuclear Law, IAEA Safety Standards Series for Radioactive Material Transport and US 10CFR Part 71

  18. Assessing the consequences in a nuclear accident scenario at Cernavoda NPP

    International Nuclear Information System (INIS)

    Having in view a possible nuclear incident, considerable planning is necessary to reduce at manageable levels the types of decisions leading to effective responses concerning the public protection. One of the most important parts of an emergency response plan is the computerized system which allows to predict the radiological impact of the accident and to provide information in a manageable and effective form for evaluating alternative countermeasure strategies in the various stages of the accident. In this paper the PC-COSYMA results for early containment failure of a CANDU reactor are presented. The deterministic health effects arising in nuclear accident situation are also presented. As source term we have used the core inventory obtained with ORIGEN computer code. The essential input parameters for PC-COSYMA computer code are also done. (authors)

  19. Two decades of radiological accidents direct causes, roots causes and consequences

    Directory of Open Access Journals (Sweden)

    Rozental Jose de Julio

    2002-01-01

    of Safety Performance and Safety Culture. Accident investigation is the first step toward avoiding future injures and financial losses, by prevention of recur recurrence. On the other hand, accident investigation is also essential for the establishment of the responsibilities and liability for the consequences. This document discuss the main accidents that have happened in the last two decades, in terms of causes, consequences, similarities and lessons learned when sealed sources have been damaged, lost, stolen and abandoned. In considerable majority death and serious injuries were resulted from failures in the safety system for radiation sources and for the security of radioactive materials.

  20. Determination of a source term for a time fractional diffusion equation with an integral type over-determining condition

    Directory of Open Access Journals (Sweden)

    Timurkhan S. Aleroev

    2013-12-01

    Full Text Available We consider a linear heat equation involving a fractional derivative in time, with a nonlocal boundary condition. We determine a source term independent of the space variable, and the temperature distribution for a problem with an over-determining condition of integral type. We prove the existence and uniqueness of the solution, and its continuous dependence on the data.

  1. Nonradioactive Environmental Emissions Chemical Source Term for the Double Shell Tank (DST) Vapor Space During Waste Retrieval Operations

    Energy Technology Data Exchange (ETDEWEB)

    MAY, T.H.

    2000-04-21

    A nonradioactive chemical vapor space source term for tanks on the Phase 1 and the extended Phase 1 delivery, storage, and disposal mission was determined. Operations modeled included mixer pump operation and DST waste transfers. Concentrations of ammonia, specific volatile organic compounds, and quantitative volumes of aerosols were estimated.

  2. Laser accidents: Being Prepared

    Energy Technology Data Exchange (ETDEWEB)

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

  3. Communication and industrial accidents

    OpenAIRE

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the concept of role distance, which emphasizes the organizational characteristics. The general hypothesis is that communication failures are a main cause of role distance and accident-proneness within orga...

  4. The Chernobyl accident consequences

    International Nuclear Information System (INIS)

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  5. Proceedings of the first part of a joint OECD(NEA)/CEC workshop on recent advances in reactor accident consequence assessment

    International Nuclear Information System (INIS)

    The first part of the Joint Workshop, organised by the NEA, is focused on the progress achieved in the work of CSNI's GRECA (Group of Experts on Accident Consequences). The program is composed of the following papers. Session 1: characteristics of the Chernobyl release and fallout that affect transport and behaviour of radioactive substances in the environment; Chernobyl accident and hot particles in the fallout; radionuclides associated with colloids and particles in the Chernobyl fallout; source term in the Chernobyl accident; long range transport of radionuclides; parameters in consequence calculations for an urban area. Session 2: review of evaluations concerning radionuclide transfer to foodstuffs via plants in view of the data available after the Chernobyl accident; GRECA review of Chernobyl data on transfer to animal products; Chernobyl accident radiometric data (Cs-137 in fresh water fishes of north Italy lakes); distribution of Cs-137 in water sediment and fish in the Ijsselmeer (Netherlands); uptake in the human body resulting from the Chernobyl accident; radioactivity of people in the nordic countries following the Chernobyl accident; preparations for an international study to evaluate long-range transport models against the Chernobyl accident

  6. Long term response stability of a well-type ionization chamber used in calibration of high dose rate brachytherapy sources

    Directory of Open Access Journals (Sweden)

    Vandana S

    2010-01-01

    Full Text Available Well-type ionization chamber is often used to measure strength of brachytherapy sources. This study aims to check long term response stability of High Dose Rate (HDR -1000 Plus well-type ionization chamber in terms of reference air kerma rate (RAKR of a reference 137 Cs brachytherapy source and recommend an optimum frequency of recalibration. An HDR-1000 Plus well-type ionization chamber, a reference 137 Cs brachytherapy source (CDCSJ5, and a MAX-4000 electrometer were used in this study. The HDR-1000 Plus well-type chamber was calibrated in terms of reference air kerma rate by the Standards Laboratory of the International Atomic Energy Agency (IAEA, Vienna. The response of the chamber was verified at regular intervals over a period of eight years using the reference 137 Cs source. All required correction factors were applied in the calculation of the RAKR of the 137 Cs source. This study reveals that the response of the HDR-1000 Plus well-type chamber was well within ±0.5% for about three years after calibration/recalibration. However, it shows deviations larger than ±0.5% after three years of calibration/recalibration and the maximum variation in response of the chamber during an eight year period was 1.71%. The optimum frequency of recalibration of a high dose rate well-type chamber should be three years.

  7. Long term response stability of a well-type ionization chamber used in calibration of high dose rate brachytherapy sources.

    Science.gov (United States)

    Vandana, S; Sharma, S D

    2010-04-01

    Well-type ionization chamber is often used to measure strength of brachytherapy sources. This study aims to check long term response stability of High Dose Rate (HDR)-1000 Plus well-type ionization chamber in terms of reference air kerma rate (RAKR) of a reference (137)Cs brachytherapy source and recommend an optimum frequency of recalibration. An HDR-1000 Plus well-type ionization chamber, a reference (137)Cs brachytherapy source (CDCSJ5), and a MAX-4000 electrometer were used in this study. The HDR-1000 Plus well-type chamber was calibrated in terms of reference air kerma rate by the Standards Laboratory of the International Atomic Energy Agency (IAEA), Vienna. The response of the chamber was verified at regular intervals over a period of eight years using the reference (137)Cs source. All required correction factors were applied in the calculation of the RAKR of the (137)Cs source. This study reveals that the response of the HDR-1000 Plus well-type chamber was well within +/-0.5% for about three years after calibration/recalibration. However, it shows deviations larger than +/-0.5% after three years of calibration/recalibration and the maximum variation in response of the chamber during an eight year period was 1.71%. The optimum frequency of recalibration of a high dose rate well-type chamber should be three years.

  8. Investigation and analysis of NORM source term in the phosphate industry based on the first nationwide pollution source survey

    International Nuclear Information System (INIS)

    China has launched the First Nationwide Pollution Source Survey (FNPSS) during 2006-2009. Ministry Environmental Protection (MEP) sponsored the campaign of measuring the natural radionuclide contents in all factories and mines nationwide in relation to phosphate, rare-earth, niobium/tantalum, zircon, tin, lead/zinc, copper, iron, coal, aluminum and vanadium. This paper analyzes mainly the data on the contents of U, 232Th and 226Ra in phosphate ore and solid waste produced by the phosphate industry in China, as one of a series of papers on naturally occurring radioactive materials (NORMs) investigation. It is concluded that the averages of U, 232Th and 226Ra in phosphate ore are 396 Bq/kg, 26 Bq/kg and 403.6 Bq/kg, respectively. The average of U and 226Ra contents in solid waste produced by the phosphate industry are both less than 200 Bq/kg, mostly. The range of U and 226Ra are 22.7-723.6 Bq/kg and 5.6-1042.1 Bq/kg, respectively. The 232Th content is very low. It is suggested that the phosphate industrial solid waste should be subject to sort management, and some phosphate industry factories and mines should carry out relevant investigation, radiation evaluation and research. (authors)

  9. Preliminary report about Goiania radiological accident, Brazil

    International Nuclear Information System (INIS)

    The events that originate the Goiania radiological accident involving the rupture of Cesium 137 source, the source characteristics, the medical aspects related to the triage of victims, the medical attendance, and the special measurements of decontamination in the Goiania General Hospital (HGG), are described. (M.C.K.)

  10. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    International Nuclear Information System (INIS)

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the 3H, 2.7% of the 90Sr, 15% of the 129I, 20% of the 131I, and 42% of the 137Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of 90Sr and 144Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement

  11. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  12. Investigation of Key Factors for Accident Severity at Railroad Grade Crossings by Using a Logit Model

    OpenAIRE

    Hu, Shou-Ren; Li, Chin-Shang; Lee, Chi-Kang

    2010-01-01

    Although several studies have used logit or probit models and their variants to fit data of accident severity on roadway segments, few have investigated accident severity at a railroad grade crossing (RGC). Compared to accident risk analysis in terms of accident frequency and severity of a highway system, investigation of the factors contributing to traffic accidents at an RGC may be more complicated because of additional highway–railway interactions. Because the proportional odds assumption ...

  13. Development and testing of a prototype instrumented bicycle model for the prevention of cyclist accidents

    OpenAIRE

    Miah, S.; Kaparias, I.; Liatsis, P.

    2015-01-01

    Cycling is an increasingly popular mode of travel in cities owing to the great advantages that it offers in terms of space consumption, health and environmental sustainability, and is therefore favoured and promoted by many city authorities worldwide. However, cycling is also perceived as relatively unsafe, and therefore it has yet to be adopted as a viable alternative to the private car. Rising accident numbers, unfortunately, confirm this perception as reality, with a particular source of h...

  14. An assessment of the radiological consequences of accidents in research reactors

    International Nuclear Information System (INIS)

    This work analyses the radiological consequences of accidents in two types of research reactors: a 5 MWt open pool reactor and a 50 MWt PWR reactor. Two siting cases have been considered: the reactor located near to a large population center and sited in a rural area. The influence of several factors such as source term, meteorological conditions and population distribution have been considered in the present analysis. (author)

  15. Fukushima nuclear power plant accident was preventable

    Science.gov (United States)

    Kanoglu, Utku; Synolakis, Costas

    2015-04-01

    On 11 March 2011, the fourth largest earthquake in recorded history triggered a large tsunami, which will probably be remembered from the dramatic live pictures in a country, which is possibly the most tsunami-prepared in the world. The earthquake and tsunami caused a major nuclear power plant (NPP) accident at the Fukushima Dai-ichi, owned by Tokyo Electric Power Company (TEPCO). The accident was likely more severe than the 1979 Three Mile Island and less severe than the Chernobyl 1986 accidents. Yet, after the 26 December 2004 Indian Ocean tsunami had hit the Madras Atomic Power Station there had been renewed interest in the resilience of NPPs to tsunamis. The 11 March 2011 tsunami hit the Onagawa, Fukushima Dai-ichi, Fukushima Dai-ni, and Tokai Dai-ni NPPs, all located approximately in a 230km stretch along the east coast of Honshu. The Onagawa NPP was the closest to the source and was hit by an approximately height of 13m tsunami, of the same height as the one that hit the Fukushima Dai-ichi. Even though the Onagawa site also subsided by 1m, the tsunami did not reach to the main critical facilities. As the International Atomic Energy Agency put it, the Onagawa NPP survived the event "remarkably undamaged." At Fukushima Dai-ichi, the three reactors in operation were shut down due to strong ground shaking. The earthquake damaged all offsite electric transmission facilities. Emergency diesel generators (EDGs) provided back up power and started cooling down the reactors. However, the tsunami flooded the facilities damaging 12 of its 13 EDGs and caused a blackout. Among the consequences were hydrogen explosions that released radioactive material in the environment. It is unfortunately clear that TEPCO and Japan's principal regulator Nuclear and Industrial Safety Agency (NISA) had failed in providing a professional hazard analysis for the plant, even though their last assessment had taken place only months before the accident. The main reasons are the following. One

  16. A Methodology for a Comprehensive Probabilistic Tsunami Hazard Assessment: Multiple Sources and Short-Term Interactions

    OpenAIRE

    Grezio Anita; Roberto Tonini; Laura Sandri; Simona Pierdominici; Jacopo Selva

    2015-01-01

    We propose a methodological approach for a comprehensive and total probabilistic tsunami hazard assessment (TotPTHA), in which many different possible source types concur to the definition of the total tsunami hazard at given target sites. In a multi-hazard and multi-risk perspective, the approach allows us to consider all possible tsunamigenic sources (seismic events, slides, volcanic eruptions, asteroids, etc.). In this respect, we also formally introduce and discuss the treatment of intera...

  17. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs.

  18. Review of models applicable to accident aerosols

    International Nuclear Information System (INIS)

    Estimations of potential airborne-particle releases are essential in safety assessments of nuclear-fuel facilities. This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident-generated aerosol sources. Such characterization of the accident-generated aerosols is a necessary step toward estimating their eventual release in any accident scenario. Existing aerosol models can predict the size distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation, and other phenomena. Models developed in the fields of fluid mechanics, indoor air pollution, and nuclear-reactor accidents are reviewed with this nuclear fuel facility application in mind. The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity

  19. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  20. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  1. Accidents on ships in the Danish International Ship register

    DEFF Research Database (Denmark)

    Ádám, Balázs; Rasmussen, Hanna Barbara

    our study is to describe trend of accidents and their contributing factors, with special focus on nationality, occurring in ships under Danish flag in the period 2010-2012. The study used two independent data sources, the Danish Maritime Authority and the Danish Radio Medical. It is mandatory to...... report accidents causing at least one day off work beyond the day of accident but the first source contains several accidents not fulfilling this criterion, too. Radio Medical is an independent service where all Danish ships may seek medical advice. The data sets were merged by identification number to...... create a single database that has been studied by descriptive statistics and regression analysis. Findings show a stabilised number of accidents in the analysed period. The occurrence of accidents is influenced by nationality. There is a higher frequency of reported injuries found among Danish and other...

  2. National emergency plan for nuclear accidents

    International Nuclear Information System (INIS)

    The national emergency plan for nuclear accidents is a plan of action designed to provide a response to accidents involving the release or potential release of radioactive substances into the environment, which could give rise to radiation exposure to the public. The plan outlines the measures which are in place to assess and mitigate the effects of nuclear accidents which might pose a radiological hazard in ireland. It shows how accident management will operate, how technical information and monitoring data will be collected, how public information will be provided and what measures may be taken for the protection of the public in the short and long term. The plan can be integrated with the Department of Defence arrangements for wartime emergencies

  3. Communication and industrial accidents

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the conc

  4. Accidents - personal factors

    Energy Technology Data Exchange (ETDEWEB)

    Zaitsev, S.L.; Tsygankov, A.V.

    1982-03-01

    This paper evaluates influence of selected personal factors on accident rate in underground coal mines in the USSR. Investigations show that so-called organizational factors cause from 80 to 85% of all accidents. About 70% of the organizational factors is associated with social, personal and economic features of personnel. Selected results of the investigations carried out in Donbass mines are discussed. Causes of miner dissatisfaction are reviewed: 14% is caused by unsatisfactory working conditions, 21% by repeated machine failures, 16% by forced labor during days off, 14% by unsatisfactory material supply, 16% by hard physical labor, 19% by other reasons. About 25% of miners injured during work accidents are characterized as highly professionally qualified with automatic reactions, and about 41% by medium qualifications. About 60% of accidents is caused by miners with less than a 3 year period of service. About 15% of accidents occurs during the first month after a miner has returned from a leave. More than 30% of accidents occurs on the first work day after a day or days off. Distribution of accidents is also presented: 19% of accidents occurs during the first 2 hours of a shift, 36% from the second to the fourth hour, and 45% occurs after the fourth hour and before the shift ends.

  5. Accident investigation and analysis

    NARCIS (Netherlands)

    Kampen, J. van; Drupsteen, L.

    2013-01-01

    Many organisations and companies take extensive proactive measures to identify, evaluate and reduce occupational risks. However, despite these efforts things still go wrong and unintended events occur. After a major incident or accident, conducting an accident investigation is generally the next ste

  6. Modeling accident frequency in Denmark for improving road safety

    DEFF Research Database (Denmark)

    Lyckegaard, Allan; Hels, Tove; Kaplan, Sigal;

    the infrastructure characteristics and the traffic conditions of the road. The model can be used to point out high risk road segments and support road authorities in planning interventions for the improvement of road safety on Danish roads. The number of accidents on a road link was modeled using a count model after......Traffic accidents result in huge costs to society in terms of death, injury, lost productivity, and property damage. The main objective of the current study is the development of an accident frequency model that predicts the expected number of accidents on a given road segment, provided...... verifying the presence of overdispersion in the variance of the counts. The model relates the number of accidents to the characteristics of the infrastructure in terms of geometry and traffic in order to identify which risk factors and at which degree relate to the possibility of accident occurrence. Data...

  7. DOE modifications to the MAAP [Modular Accident Analysis Program] code

    International Nuclear Information System (INIS)

    This report presents an enhanced model for the MAAP code that addresses fuel-cladding interaction and core mass relocation during core degradation. The main purpose of this work is to assess the potential for in-vessel hydrogen production and to reduce the uncertainty in fission product source term evaluation. The model provides a description of fuel behavior in which the fuel comprises uranium dioxide, zirconium dioxide, and U-Zr-O compounds. The composition of the U-Zr-O compounds and their solidus and liquidus temperatures are calculated throughout the core melt transient. The interaction of control rod materials with fuel and cladding and the relocation of control rod materials are not addressed in this enhanced model. The enhanced core melt progression model has been applied to a hypothetical station blackout accident with a small break via the reactor coolant pump seals. The new model has been benchmarked against both the LOFT experiment LP-FP-2 and the TMI-2 accident prior to the B-loop pump restart. Although some uncertainties and deviations were seen, general agreement was obtained with the experimental data and with the TMI-2 accident. 21 refs., 30 figs

  8. Economic development and traffic accident mortality in the industrialized world, 1962-1990

    NARCIS (Netherlands)

    E.F. van Beeck (Ed); G.J.J.M. Borsboom (Gerard); J.P. Mackenbach (Johan)

    2000-01-01

    textabstractBACKGROUND: We examined the association between prosperity and traffic accident mortality in the industrialized world in a long-term perspective. METHODS: We calculated traffic accident mortality, traffic mobility and the fatal injury rate of 21 industrializ

  9. Long-term drift of the coronal source magnetic flux and the total solar irradiance

    OpenAIRE

    Lockwood, Mike; Stamper, R

    1999-01-01

    We test the method of Lockwood et al. [1999] for deriving the coronal source flux from the geomagnetic aa index and show it to be accurate to within 12% for annual means and 4.5% for averages over a sunspot cycle. Using data from four solar constant monitors during 1981-1995, we find a linear relationship between this magnetic flux and the total solar irradiance. From this correlation, we show that the 131% rise in the mean coronal source field over the interval 1901-1995 corresponds to a ris...

  10. Modelling and analysis of severe accidents for VVER-1000 reactors

    OpenAIRE

    Tusheva, Polina

    2013-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the re...

  11. 惠州市饮用水源地突发事故风险分析及应急备用水源初选%Risk analysis of accident of Huizhou drinking water source area and primary election of emergency standby water source

    Institute of Scientific and Technical Information of China (English)

    曾秀华

    2012-01-01

      针对惠州市各县区主要饮用水源地现状,结合流域历史水文资料、水利工程调度情况和水污染事件案例进行分析,指出惠州市饮用水源地突发事故风险主要来自水污染事故,分析其主要风险源,并在此基础上提出了惠州市应急备用水源初选方案,该成果对有效推进惠州市应急备用水源保障规划及工程建设、保障惠州市应急供水安全具有重要意义。%  Aimed at the status quo of the main drinking water source in Huizhou city and county district,combined with the watershed historical hydrological data, water conservancy project scheduling and water pollution incident case, Huizhou drinking water sources accident risk that mainly come from a water pollution accident were pointed out, the main risk source was analyzed, and on this basis, the huizhou emergency standby water primary plan was proposed, it was an important significance to effectively promote the Huizhou City emergency wellhead protection planning and construction, protection of Huizhou City emergency water supply security

  12. Long-Term Variability in Sugarcane Bagasse Feedstock Compositional Methods: Sources and Magnitude of Analytical Variability

    Energy Technology Data Exchange (ETDEWEB)

    Templeton, David W.; Sluiter, Justin B.; Sluiter, Amie; Payne, Courtney; Crocker, David P.; Tao, Ling; Wolfrum, Ed

    2016-10-18

    In an effort to find economical, carbon-neutral transportation fuels, biomass feedstock compositional analysis methods are used to monitor, compare, and improve biofuel conversion processes. These methods are empirical, and the analytical variability seen in the feedstock compositional data propagates into variability in the conversion yields, component balances, mass balances, and ultimately the minimum ethanol selling price (MESP). We report the average composition and standard deviations of 119 individually extracted National Institute of Standards and Technology (NIST) bagasse [Reference Material (RM) 8491] run by seven analysts over 7 years. Two additional datasets, using bulk-extracted bagasse (containing 58 and 291 replicates each), were examined to separate out the effects of batch, analyst, sugar recovery standard calculation method, and extractions from the total analytical variability seen in the individually extracted dataset. We believe this is the world's largest NIST bagasse compositional analysis dataset and it provides unique insight into the long-term analytical variability. Understanding the long-term variability of the feedstock analysis will help determine the minimum difference that can be detected in yield, mass balance, and efficiency calculations. The long-term data show consistent bagasse component values through time and by different analysts. This suggests that the standard compositional analysis methods were performed consistently and that the bagasse RM itself remained unchanged during this time period. The long-term variability seen here is generally higher than short-term variabilities. It is worth noting that the effect of short-term or long-term feedstock compositional variability on MESP is small, about $0.03 per gallon. The long-term analysis variabilities reported here are plausible minimum values for these methods, though not necessarily average or expected variabilities. We must emphasize the importance of training and

  13. Examination of some assumed severe reactor accidents at the Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Knowledge and analysis methods of severe accidents at nuclear power plants and of subsequent response of primary system and containment have been developed in last few years to the extent that realistic source tems of the specified accident sequences can be calculated for the Finnish nuclear power plants. The objective of this investigation was to calculate the source terms of off-site consequences brought about by some selected severe accident sequences initiated by the total loss of on-site and off-site AC power at the Olkiluoto nuclear power plant. The results describing the estimated off-site health risks are expressed as conditional assuming that the accident has taken place, because the probabilities of the occurence of the accident sequences considered have not been analysed in this study. The range and probabilities of occurence of health detriments are considered by calculating consequences in different weeather conditions and taking into account the annual frequency of each weather condition and statistical population distribution. The calculational results indicate that the reactor building provides and additional holdup and deposition of radioactive substance (except coble gases) released from the containment. Furthermore, the release fractions of the core inventory to the environment of volatile fission products such as iodine, cesium and tellurium remain under 0.03. No early health effects are predicted for the surrounding population in case the assumed short-tem countermeasures are performed effectively. Acute health effects are extremely improbable even without any active countermeasure. By reducing the long-term exposure from contaminated agricultural products, the collective dose from natural long-term background radiation, for instance in the sector of 30 degrees towards the southern Finland up to the distance of 300 kilometers, would be expected to increase with 2-20 percent depending on the release considered

  14. Numerical method of identification of an unknown source term in a heat equation

    Directory of Open Access Journals (Sweden)

    Fatullayev Afet Golayo?lu

    2002-01-01

    Full Text Available A numerical procedure for an inverse problem of identification of an unknown source in a heat equation is presented. Approach of proposed method is to approximate unknown function by polygons linear pieces which are determined consecutively from the solution of minimization problem based on the overspecified data. Numerical examples are presented.

  15. Evaluation of the Transient Hydrologic Source Term for the Cambric Underground Nuclear Test at Frenchman Flat, Nevada test Site

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-12

    The objective of Phase II HST work is to develop a better understanding of the evolution of the HST for 1,000 years at the CAMBRIC underground nuclear test site in Frenchman Flat at the NTS. This work provides a better understanding of activities as they actually occurred, incorporates improvements based on recent data acquisition, and provides a basis to use the CAMBRIC site for model validation and monitoring activities as required by the UGTA Project. CAMBRIC was the only test in Frenchman Flat detonated under the water table and best represents a fully saturated environment. These simulations are part of a broad Phase II Frenchman Flat Corrective Action Unit (CAU) flow and transport modeling effort being conducted by the Department of Energy (DOE) Underground Test Area (UGTA) Project. HST simulations provide, either directly or indirectly, the source term used in the CAU model to calculate a contaminant boundary. Work described in this report augments Phase I HST calculations at CAMBRIC conducted by Tompson et al. (1999) and Pawloski et al. (2001). Phase II HST calculations have been organized to calculate source terms under two scenarios: (1) A representation of the transient flow and radionuclide release behavior at the CAMBRIC site that is more specific than Tompson et al. (1999). This model reflects the influence of the background hydraulic gradient, residual test heat, pumping experiment, and ditch recharge, and takes into account improved data sources and modeling approaches developed since the previous efforts. Collectively, this approach will be referred to as the transient CAMBRIC source term. This report describes the development of the transient CAMBRIC HST. (2) A generic release model made under steady-state flow conditions, in the absence of any transient effects, at the same site with the same radiologic source term. This model is for use in the development of simpler release models for the other nine underground test sites in the Frenchman Flat

  16. Recovery of wild Pacific oyster, Crassostrea gigas in terms of reproduction and gametogenesis two-years after the Hebei Spirit Oil Spill Accident off the West Coast of Korea

    Science.gov (United States)

    Mondol, Mostafizur Rahman; Keshavmurthy, Shashank; Lee, Hee-Jung; Hong, Hyun-Ki; Park, Heung-Sik; Park, Sang-Rul; Kang, Chang-Keun; Choi, Kwang-Sik

    2015-12-01

    The Hebei Spirit oil spill in December 2007 at Taean off the west coast of Korea was the largest oil tanker accident in Korea. However, the impact of the spill on physiology of benthic animals remains largely unknown. Two-years after the accident, we compared reproductive effort and annual gametogenesis of the wild Pacific oyster Crassostrea gigas, residing at oil spill site with a control oyster population in Incheon Bay, North-West coast of Korea. Results showed that the oyster sampled from the oil spill site showed a significantly higher (279.0 mg standard animal-1, Pphysiological status to normal level after two years of the oil spill accident.

  17. Results of an aqueous source term model for a radiological risk assessment of the Drigg LLW Site, U.K.

    OpenAIRE

    Small, J. S.; Humphreys, P. N.; Johnstone, T. L.; Plant, R.; Randall, M. G.; Trivedi, D. P.

    1999-01-01

    A radionuclide source term model has been developed which simulates the biogeochemical evolution of the Drigg low level waste (LLW) disposal site. The DRINK (DRIgg Near field Kinetic) model provides data regarding radionuclide concentrations in groundwater over a period of 100,000 years, which are used as input to assessment calculations for a groundwater pathway. The DRINK model also provides input to human intrusion and gaseous assessment calculations through simulation of the solid radionu...

  18. Results of an aqueous source term model for a radiological risk assessment of the Drigg LLW site

    OpenAIRE

    Small, J.; Humphreys, Paul; Johnstone, T. J.; Plant, R.; Randall, M. G.; Trivedi, D. P.

    2000-01-01

    A radionuclide source term model has been developed which simulates the biogeochemical evolution of the Drigg low level waste (LLW) disposal site. The DRINK (DRIgg Near field Kinetic) model provides data regarding radionuclide concentrations in groundwater over a period of 100,000 years, which are used as inputs to safety assessment calculations. The DRINK model considers the coupled interaction of the effects of fluid flow, microbiology, corrosion, chemical reaction, sorption and radioactive...

  19. Integration of geological and macroseismic data to define probable seismic scenarios in terms of macroseismic intensity and related seismogenic source models. The Maiella 1706 earthquake (Abruzzo, Italy)

    OpenAIRE

    De Nardis, R.; Dipartimento della Protezione Civile; Galadini, F.; Istituto Nazionale di Geofisica e Vulcanologia, Sezione Milano-Pavia, Milano, Italia; Lavecchia, G.; Università G. D'Annunzio, Chieti; Marcucci, S.; Dipartimento della Protezione Civile, Ufficio Valutazione Prevenzione e Mitigazione del Rischio Sismico,Via Vitorchiano 2, 00189 Rome, Italy; Milana, G.; Istituto Nazionale di Geofisica e Vulcanologia, Sezione Roma1, Roma, Italia; Pace, B.; Università G. D'Annunzio, Chieti; Visini, F.; Università G. D'Annunzio, Chieti

    2007-01-01

    Integration of geological and macroseismic data to define probable seismic scenarios in terms of macroseismic intensity and related seismogenic source models. The Maiella 1706 earthquake (Abruzzo, Italy)

  20. Evaluating the maximum likelihood method for detecting short-term variability of AGILE γ-ray sources

    Science.gov (United States)

    Bulgarelli, A.; Chen, A. W.; Tavani, M.; Gianotti, F.; Trifoglio, M.; Contessi, T.

    2012-04-01

    Context. The AGILE space mission (whose instrument is sensitive to the energy ranges 18-60 keV, and 30 MeV-50 GeV) has been operating since 2007. Assessing the statistical significance of the time variability of γ-ray sources above 100 MeV is a primary task of the AGILE data analysis. In particular, it is important to verify the instrument sensitivity in terms of Poisson modeling of the data background, and to determine the post-trial confidence of detections. Aims: The goals of this work are: (i) to evaluate the distributions of the likelihood ratio test for both "empty" fields and regions of the Galactic plane, and (ii) to calculate the probability of false detections over multiple time intervals. Methods: We describe in detail the techniques used to search for short-term variability in the AGILE γ-ray source database. We describe the binned maximum likelihood method used for the analysis of AGILE data, and the numerical simulations that support the characterization of the statistical analysis. We apply our method to both Galactic and extragalactic transients, and provide a few examples. Results: After checking the reliability of the statistical description tested with the real AGILE data, we obtain the distribution of p-values for blind and specific source searches. We apply our results to the determination of the post-trial statistical significance of detections of transient γ-ray sources in terms of pre-trial values. Conclusions: The results of our analysis allow a precise determination of the post-trial significance of γ-ray sources detected by AGILE.

  1. Study of the source term of radiation of the CDTN GE-PET trace 8 cyclotron with the MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Benavente C, J. A.; Lacerda, M. A. S.; Fonseca, T. C. F.; Da Silva, T. A. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil); Vega C, H. R., E-mail: jhonnybenavente@gmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    Full text: The knowledge of the neutron spectra in a PET cyclotron is important for the optimization of radiation protection of the workers and individuals of the public. The main objective of this work is to study the source term of radiation of the GE-PET trace 8 cyclotron of the Development Center of Nuclear Technology (CDTN/CNEN) using computer simulation by the Monte Carlo method. The MCNPX version 2.7 code was used to calculate the flux of neutrons produced from the interaction of the primary proton beam with the target body and other cyclotron components, during 18F production. The estimate of the source term and the corresponding radiation field was performed from the bombardment of a H{sub 2}{sup 18}O target with protons of 75 μA current and 16.5 MeV of energy. The values of the simulated fluxes were compared with those reported by the accelerator manufacturer (GE Health care Company). Results showed that the fluxes estimated with the MCNPX codes were about 70% lower than the reported by the manufacturer. The mean energies of the neutrons were also different of that reported by GE Health Care. It is recommended to investigate other cross sections data and the use of physical models of the code itself for a complete characterization of the source term of radiation. (Author)

  2. Study of the source term of radiation of the CDTN GE-PET trace 8 cyclotron with the MCNPX code

    International Nuclear Information System (INIS)

    Full text: The knowledge of the neutron spectra in a PET cyclotron is important for the optimization of radiation protection of the workers and individuals of the public. The main objective of this work is to study the source term of radiation of the GE-PET trace 8 cyclotron of the Development Center of Nuclear Technology (CDTN/CNEN) using computer simulation by the Monte Carlo method. The MCNPX version 2.7 code was used to calculate the flux of neutrons produced from the interaction of the primary proton beam with the target body and other cyclotron components, during 18F production. The estimate of the source term and the corresponding radiation field was performed from the bombardment of a H218O target with protons of 75 μA current and 16.5 MeV of energy. The values of the simulated fluxes were compared with those reported by the accelerator manufacturer (GE Health care Company). Results showed that the fluxes estimated with the MCNPX codes were about 70% lower than the reported by the manufacturer. The mean energies of the neutrons were also different of that reported by GE Health Care. It is recommended to investigate other cross sections data and the use of physical models of the code itself for a complete characterization of the source term of radiation. (Author)

  3. Estimation of the SO2 source term for the Holuhraun event and its influence on central Europe air quality

    Science.gov (United States)

    Arnold, Delia; Iren Kristiansen, Nina; Theys, Nicolas; Brenot, Hugues; Maurer, Christian; Wotawa, Gerhard; Stebel, Kerstin; Holla, Robert; Gilge, Stefan; Flemming, Johannes; Stohl, Andreas; Hirtl, Marcus

    2015-04-01

    On 29 August 2014 a fissure eruption began in Holuhraun, Northeastern Iceland, associated with increased volcanic activity in the Bárdarbunga system. For more than 150 days, the eruption released large quantities of SO2 into the atmosphere affecting not only the local Icelandic air quality, but also leading to periods of increased ambient SO2 concentrations in parts of mainland Europe. During the second half of September, significant amounts of SO2 were rapidly transported southward by favourable meteorological conditions and several countries in Central Europe experienced high ground-level SO2 concentrations. The measured concentrations reached and even exceeded the EC directive health thresholds. In this work, we evaluate the air quality effects in Europe during this targeted period using both ground-based and satellite observations (GOME2B and OMI) as well as dispersion modelling with the Lagrangian particle model FLEXPART. We estimate the volcanic SO2 source emissions by comparing the satellite observations with atmospheric transport model simulations in an inverse modelling approach. The estimated source term is evaluated against independent ground-based observational data (e.g. MAX-DOAS, Brewer) and used as emission term in dispersion model forecasts for evaluating the air quality effects in Europe. In addition, the potential use of air quality data to perform the source term estimation by inversion with ground-based data will also be investigated.

  4. Modeling and assessment of accident consequences; development of RODOS, a real-time on-line decision support system for nuclear emergencies in Europe

    International Nuclear Information System (INIS)

    In cooperation with NRPB (UK), the first version 1.0 of PC COSYMA for use on advanced PCs has been released; during a training course in mid 1993, future users were educated in operating the software. The main frame version of the program package COSYMA has been up-dated with new dose conversion factors and fodd-chain data and was distributed to some 20 institutes in Europe and abroad. The comparative calculations performed within the international OECD(NEA)/CEC intercomparison of accident consequence assessment codes were analysed and documented in three reports. Furtheron, consequence assessments have been performed for the research reactor BER II (two source terms) and documented; the influence on individual doses and emergency actions of inplant accident management measures in future EPRs was quantified; within th scope of a EC/US-study on the external costs of the energy cycle, accident consequences were assessed for three source terms. (orig.)

  5. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  6. Shielding analysis of proton therapy accelerators: a demonstration using Monte Carlo-generated source terms and attenuation lengths.

    Science.gov (United States)

    Lai, Bo-Lun; Sheu, Rong-Jiun; Lin, Uei-Tyng

    2015-05-01

    Monte Carlo simulations are generally considered the most accurate method for complex accelerator shielding analysis. Simplified models based on point-source line-of-sight approximation are often preferable in practice because they are intuitive and easy to use. A set of shielding data, including source terms and attenuation lengths for several common targets (iron, graphite, tissue, and copper) and shielding materials (concrete, iron, and lead) were generated by performing Monte Carlo simulations for 100-300 MeV protons. Possible applications and a proper use of the data set were demonstrated through a practical case study, in which shielding analysis on a typical proton treatment room was conducted. A thorough and consistent comparison between the predictions of our point-source line-of-sight model and those obtained by Monte Carlo simulations for a 360° dose distribution around the room perimeter showed that the data set can yield fairly accurate or conservative estimates for the transmitted doses, except for those near the maze exit. In addition, this study demonstrated that appropriate coupling between the generated source term and empirical formulae for radiation streaming can be used to predict a reasonable dose distribution along the maze. This case study proved the effectiveness and advantage of applying the data set to a quick shielding design and dose evaluation for proton therapy accelerators. PMID:25811254

  7. Real-time software for multi-isotopic source term estimation

    International Nuclear Information System (INIS)

    Consideration is given to development of software for one of crucial components of the RODOS - assessment of the source rate (SR) from indirect measurements. Four components of the software are described in the paper. First component is a GRID system, which allow to prepare stochastic meteorological and radioactivity fields using measured data. Second part is a model of atmospheric transport which can be adapted for emulation of practically any gamma dose/spectrum detectors. The third one is a method which allows space-time and quantitative discrepancies in measured and modelled data to be taken into account simultaneously. It bases on the preference scheme selected by an expert. Last component is a special optimization method for calculation of multi-isotopic SR and its uncertainties. Results of a validation of the software using tracer experiments data and Chernobyl source estimation for main dose-forming isotopes are enclosed in the paper

  8. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

    International Nuclear Information System (INIS)

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

  9. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  10. Identification of sources and long term trends for pollutants in the arctic using isentropic trajectory analysis

    International Nuclear Information System (INIS)

    The understanding of factors driving climate and ecosystem changes in the Arctic requires careful consideration of the sources, correlation and trends for anthropogenic pollutants. The database from the NOAA-CMDL Barrow Observatory (71deg.17'N, 156deg.47'W) is the longest and most complete record of pollutant measurements in the Arctic. It includes observations of carbon dioxide (CO2), methane (CH4), carbon monoxide (CO), ozone (O3), aerosol scattering coefficient (σsp), aerosol number concentration (NCasl), etc. The objectives of this study are to understand the role of long-range transport to Barrow in explaining: (1) the year-to-year variations, and (2) the trends in the atmospheric chemistry record at the NOAA-CMDL Barrow observatory. The key questions we try to answer are: 1. What is the relationship between various chemical species measured at Barrow Observatory, Alaska and transport pathways at various altitudes? 2. What are the trends of species and their relation to transport patterns from the source regions? 3. What is the impact of the Prudhoe Bay emissions on the Barrow's records? To answer on these questions we apply the following main research tools. First, it is an isentropic trajectory model used to calculate the trajectories arriving at Barrow at three altitudes of 0.5, 1.5 and 3 km above sea level. Second - clustering procedure used to divide the trajectories into groups based on source regions. Third - various statistical analysis tools such as the exploratory data analysis, two component correlation analysis, trend analysis, principal components and factor analysis used to identify the relationship between various chemical species vs. source regions as a function of time. In this study, we used the chemical data from the NOAA-CMDL Barrow observatory in combination with isentropic backward trajectories from gridded ECMWF data to understand the importance of various pollutant source regions on atmospheric composition in the Arctic. We calculated

  11. Identification of sources and long term trends for pollutants in the arctic using isentropic trajectory analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mahura, A.; Jaffe, D.; Harris, J.

    2003-07-01

    The understanding of factors driving climate and ecosystem changes in the Arctic requires careful consideration of the sources, correlation and trends for anthropogenic pollutants. The database from the NOAA-CMDL Barrow Observatory (71deg.17'N, 156deg.47'W) is the longest and most complete record of pollutant measurements in the Arctic. It includes observations of carbon dioxide (CO{sub 2}), methane (CH{sub 4}), carbon monoxide (CO), ozone (O{sub 3}), aerosol scattering coefficient ({sigma}{sub sp}), aerosol number concentration (NC{sub asl}), etc. The objectives of this study are to understand the role of long-range transport to Barrow in explaining: (1) the year-to-year variations, and (2) the trends in the atmospheric chemistry record at the NOAA-CMDL Barrow observatory. The key questions we try to answer are: 1. What is the relationship between various chemical species measured at Barrow Observatory, Alaska and transport pathways at various altitudes? 2. What are the trends of species and their relation to transport patterns from the source regions? 3. What is the impact of the Prudhoe Bay emissions on the Barrow's records? To answer on these questions we apply the following main research tools. First, it is an isentropic trajectory model used to calculate the trajectories arriving at Barrow at three altitudes of 0.5, 1.5 and 3 km above sea level. Second - clustering procedure used to divide the trajectories into groups based on source regions. Third - various statistical analysis tools such as the exploratory data analysis, two component correlation analysis, trend analysis, principal components and factor analysis used to identify the relationship between various chemical species vs. source regions as a function of time. In this study, we used the chemical data from the NOAA-CMDL Barrow observatory in combination with isentropic backward trajectories from gridded ECMWF data to understand the importance of various pollutant source regions on

  12. Severe accident risks from external events

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

    This paper reviews the early development of design requirements for seismic events in USA early developing nuclear electric generating fleet.Notable safety studies,including WASH-1400,Sandia Siting Study and the NUREG-1150 probabilistic risk study,are briefly reviewed in terms of their relevance to extreme accidents arising from seismic and other severe accident initiators.Specific characteristic about the nature of severe accidents in nuclear power plant (NPP) are reviewed along with present day state-of-art analysis methodologies (methods for estimation of leakages and consequences of releases (MELCOR) and MELCOR accident consequence code system (MACCS)) that are used to evaluate severe accidents and to optimize mitigative and protective actions against such accidents.It is the aim of this paper to make nuclear operating nations aware of the risks that accompany a much needed energy resource and to identify some of the tools,techniques and landmark safety studies that serve to make the technology safer and to maintain vigilance and adequate safety culture for the responsible management of this valuable but unforgiving technology.

  13. Elements to diminish radioactive accidents

    International Nuclear Information System (INIS)

    In this work it is presented an application of the cause-effect diagram method or Ichikawa method identifying the elements that allow to diminish accidents when the radioactive materials are transported. It is considered the transport of hazardous materials which include radioactive materials in the period: December 1996 until March 1997. Among the identified elements by this method it is possible to mention: the road type, the radioactive source protection, the grade driver responsibility and the preparation that the OEP has in the radioactive material management. It is showed the differences found between the country inner roads and the Mexico City area. (Author)

  14. Cause Analysis of Wuhan Tianheng Building Pile Accident

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The geological condition and the original structure feature and foundation design of Wuhan Tianheng building are described. The accident appearance of pile foundation in the construction execution of work is illustrated. The generating source of this pile foundation accident is analyzed in great details.``

  15. An Iterative Regularization Method for Identifying the Source Term in a Second Order Differential Equation

    Directory of Open Access Journals (Sweden)

    Fairouz Zouyed

    2015-01-01

    Full Text Available This paper discusses the inverse problem of determining an unknown source in a second order differential equation from measured final data. This problem is ill-posed; that is, the solution (if it exists does not depend continuously on the data. In order to solve the considered problem, an iterative method is proposed. Using this method a regularized solution is constructed and an a priori error estimate between the exact solution and its regularized approximation is obtained. Moreover, numerical results are presented to illustrate the accuracy and efficiency of this method.

  16. Long-term X-ray Observations of Galactic Superluminal Sources with GRANAT/WATCH

    DEFF Research Database (Denmark)

    Sazonov, S.Y.; Sunyaev, R.; Lund, Niels

    1996-01-01

    The authors present X-ray time histories for the radio-jet sources GRS 1915+105 and GRO J1655-40 observed by the GRANAT/WATCH all-sky monitor at 8-20 keV. GRS 1915+105 is extremely variable on the time scales of months to years. The analysis of a 3-year data set gives no evidence for periodicity...... in its X-ray intensity. The light curve of GRO J1655-40 consists of strong outbursts alternating with periods of low flux....

  17. Accident resistant transport container

    Science.gov (United States)

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  18. Boating Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  19. FATAL ACCIDENT REPORTING SYSTEM (FARS)

    Science.gov (United States)

    The Fatal Accident Reporting System (FARS) database consist of three relational tables, containing data on automobile accidents on public U.S. roads that resulted in the death of one or more people within 30 days of the accident. Truck and trailer accidents are also included.

  20. Traffic Accidents on Slippery Roads

    DEFF Research Database (Denmark)

    Fonnesbech, J. K.; Bolet, Lars

    2014-01-01

    Police registrations from 65 accidents on slippery roads in normally Danish winters have been studied. The study showed: • 1 accident per 100 km when using brine spread with nozzles • 2 accidents per 100 km when using pre wetted salt • 3 accidents per 100 km when using kombi spreaders The results...

  1. ESTRO Breur Gold Medal Award Lecture 2001: Irradiation accidents - lessons for oncology?

    International Nuclear Information System (INIS)

    Considering the number of radioactive sources in use all over the world (both in industry and Medicine), irradiation accidents are exceedingly rare, as demonstrated by the main databases registering such cases: UNSCEAR, IAEA, REAC/TS (Oak Ridge, USA), the German group in Ulm and the Paris Institut Curie. The precise causes of most accidents have been openly analyzed, allowing to reduce the risk of subsequent identical accidental exposures. In addition, a rapid retrospective overview shows that positive lessons could be drawn from such accidents: - Lessons for patient management: one should keep in mind that the first ever allogeneic bone marrow transplantations were performed in 1958, on scientists from Yugoslavia who had been severely irradiated in a nuclear Research laboratory. Apart from what was learned from such accidents for the management of severe aplasia, the treatment of superficial accidental exposures has also benefited radiotherapy patients in certain specific situations. - Lessons for technology: the efforts to improve safety in nuclear plants are well known; the (successful) efforts to reduce the once-elevated risks when changing the therapeutic Cobalt 60 sources are less well known. Today, most irradiation accidents (by far) are related to misuse or loss of radioactive sources from industrial radiography sets. However, here again, various technological improvements significantly reduced the risks. - Lessons for radiobiology: the need for more and more sophisticated biological dosimetry has led to studies allowing better understanding of the short- and long-term effects of radiation on human cells. Analyses of samples taken in areas which were heavily accidentally irradiated also helped to identify, in particular, the cardinal role of TGF beta and TNF alpha in the development of fibrosis and necrosis after irradiation. - Lessons for prevention of accidents in radiotherapy: only three large-scale accidents involving external radiotherapy have been

  2. Asymptotically and exactly energy balanced augmented flux-ADER schemes with application to hyperbolic conservation laws with geometric source terms

    Science.gov (United States)

    Navas-Montilla, A.; Murillo, J.

    2016-07-01

    In this work, an arbitrary order HLL-type numerical scheme is constructed using the flux-ADER methodology. The proposed scheme is based on an augmented Derivative Riemann solver that was used for the first time in Navas-Montilla and Murillo (2015) [1]. Such solver, hereafter referred to as Flux-Source (FS) solver, was conceived as a high order extension of the augmented Roe solver and led to the generation of a novel numerical scheme called AR-ADER scheme. Here, we provide a general definition of the FS solver independently of the Riemann solver used in it. Moreover, a simplified version of the solver, referred to as Linearized-Flux-Source (LFS) solver, is presented. This novel version of the FS solver allows to compute the solution without requiring reconstruction of derivatives of the fluxes, nevertheless some drawbacks are evidenced. In contrast to other previously defined Derivative Riemann solvers, the proposed FS and LFS solvers take into account the presence of the source term in the resolution of the Derivative Riemann Problem (DRP), which is of particular interest when dealing with geometric source terms. When applied to the shallow water equations, the proposed HLLS-ADER and AR-ADER schemes can be constructed to fulfill the exactly well-balanced property, showing that an arbitrary quadrature of the integral of the source inside the cell does not ensure energy balanced solutions. As a result of this work, energy balanced flux-ADER schemes that provide the exact solution for steady cases and that converge to the exact solution with arbitrary order for transient cases are constructed.

  3. Proceedings of the first OECD (NEA) CSNI-Specialist Meeting on Instrumentation to Manage Severe Accidents

    International Nuclear Information System (INIS)

    OECD member countries have adopted various accident management measures and procedures. To initiate these measures and control their effectiveness, information on the status of the plant and on accident symptoms is necessary. This information includes physical data (pressure, temperatures, hydrogen concentrations, etc.) but also data on the condition of components such as pumps, valves, power supplies, etc. In response to proposals made by the CSNI - PWG 4 Task Group on Containment Aspects of Severe Accident Management (CAM) and endorsed by PWG 4, CSNI has decided to sponsor a Specialist Meeting on Instrumentation to Manage Severe Accidents. The knowledge-basis for the Specialist Meeting was the paper on 'Instrumentation for Accident Management in Containment'. This technical document (NEA/CSNI/R(92)4) was prepared by the CSNI - Principle Working Group Number 4 of experts on January 1992. The Specialist Meeting was structured in the following sessions: I. Information Needs for Managing Severe Accidents, II. Capabilities and Limitations of Existing Instrumentation, III. Unconventional Use and Further Development of Instrumentation, IV. Operational Aids and Artificial Intelligence. The Specialist Meeting concentrated on existing instrumentation and its possible use under severe accident conditions; it also examined developments underway and planed. Desirable new instrumentation was discussed briefly. The interactions and discussions during the sessions were helpful to bring different perspectives to bear, thus sharpening the thinking of all. Questions were raised concerning the long-term viability of current (or added) instrumentation. It must be realized that the subject of instrumentation to manage severe accidents is very new, and that no international meeting on this topic was held previously. One of the objectives was to bring this important issue to the attention of both safety authorities and experts. It could be seen from several of the presentations and from

  4. Accident management information needs

    International Nuclear Information System (INIS)

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  5. Accident and emergency management

    International Nuclear Information System (INIS)

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  6. The influence of user effect in quality assurance of accident analyses

    International Nuclear Information System (INIS)

    In this paper one identifies and discusses some of the sources by which a code user might influence the predicted results of an accident analysis. It presents some methods by which the magnitude of the user effect might be reduced and also presents some aspects of the user effect when one uses the SOPHAEROS module, as part of Accident Source Term Evaluation Code - ASTEC V2.0. Four models, in order to show the user effect in the final results (the masses of fission products retained in the Primary Heat Transport System and transferred into containment) are presented. The results of the analyzed cases show that the influence of the user decision on data, might lead to relevant differences in the final results. The choosing is determined by the user experience and the available data. (authors)

  7. Analysis of the different source terms of natural radionuclides in a river affected by NORM (Naturally Occurring Radioactive Materials) activities

    Energy Technology Data Exchange (ETDEWEB)

    Baeza, A.; Corbacho, J.A.; Guillen, J.; Salas, A.; Mora, J.C. [University of Extremadura, Caceres (Spain)

    2011-05-15

    The present work studied the radioactivity impact of a coal-fired power plant (CFPP), a NORM industry, on the water of the Regallo river which the plant uses for cooling. Downstream, this river passes through an important irrigated farming area, and it is a tributary of the Ebro, one of Spain's largest rivers. Although no alteration of the Po-210 or Th-232 content was detected, the U-234, U-238 and Ra-226 contents of the water were significantly greater immediately below CFPP's discharge point. The Ra-226 concentration decreased progressively downstream from the discharge point, but the uranium content increased significantly again at two sampling points 8 km downstream from the CFPP's effluent. This suggested the presence of another, unexpected uranium source term different from the CFPP. The input from this second uranium source term was even greater than that from the CFPP. Different hypotheses were tested (a reservoir used for irrigation, remobilization from sediments, and the effect of fertilizers used in the area), with it finally being demonstrated that the source was the fertilizers used in the adjacent farming areas.

  8. Controlling temporal solitary waves in the generalized inhomogeneous coupled nonlinear Schrödinger equations with varying source terms

    Science.gov (United States)

    Yang, Yunqing; Yan, Zhenya; Mihalache, Dumitru

    2015-05-01

    In this paper, we study the families of solitary-wave solutions to the inhomogeneous coupled nonlinear Schrödinger equations with space- and time-modulated coefficients and source terms. By means of the similarity reduction method and Möbius transformations, many types of novel temporal solitary-wave solutions of this nonlinear dynamical system are analytically found under some constraint conditions, such as the bright-bright, bright-dark, dark-dark, periodic-periodic, W-shaped, and rational wave solutions. In particular, we find that the localized rational-type solutions can exhibit both bright-bright and bright-dark wave profiles by choosing different families of free parameters. Moreover, we analyze the relationships among the group-velocity dispersion profiles, gain or loss distributions, external potentials, and inhomogeneous source profiles, which provide the necessary constraint conditions to control the emerging wave dynamics. Finally, a series of numerical simulations are performed to show the robustness to propagation of some of the analytically obtained solitary-wave solutions. The vast class of exact solutions of inhomogeneous coupled nonlinear Schrödinger equations with source terms might be used in the study of the soliton structures in twin-core optical fibers and two-component Bose-Einstein condensates.

  9. Study of safety features and accident scenarios in a fusion DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: •This paper reports progress in the fusion DEMO safety research conducted under the Broader Approach DEMO Design Activities. •Hazards of a reference DEMO concept have been assessed. •Reference accident event sequences in the reference DEMO in this study have been analyzed based on the master logic diagram (MLD) and the functional failure mode and effect analysis (FFMEA) techniques. •Accident events of particular concern in the DEMO have been selected based on the MLD and FFMEA analysis. -- Abstract: After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li2TiO3) and beryllium–titanium (Be12Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment

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

    International Nuclear Information System (INIS)

    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

  11. Cyclical Fluctuations in Workplace Accidents

    OpenAIRE

    Boone, J.; van Ours, J.C.

    2002-01-01

    This Paper presents a theory and an empirical investigation on cyclical fluctuations in workplace accidents. The theory is based on the idea that reporting an accident dents the reputation of a worker and raises the probability that he is fired. Therefore a country with a high or an increasing unemployment rate has a low (reported) workplace accident rate. The empirical investigation concerns workplace accidents in OECD countries. The analysis confirms that workplace accident rates are invers...

  12. Imbalance of energy and momentum source terms of the sea wave transfer equation for fully developed seas

    Science.gov (United States)

    Caudal, G. V.

    2012-12-01

    In the concept of full development, the sea wave spectrum is regarded as a nearly stationary solution of the wave transfer equation, where source and sink terms should be in balance with respect to both energy and momentum. Using a two-dimensional empirical sea wave spectral model at full development, this paper performs an assessment of the compatibility of the energy and momentum budgets of sea waves over the whole spectral range. Among the various combinations of model functions for wave breaking and wind source terms tested, not one is found to fulfill simultaneously the energy and momentum balance of the transfer equation. Based on experimental and theoretical grounds, wave breaking is known to contribute to frequency downshift of a narrow-banded wave spectrum when the modulational instability is combined with wave breaking. On those grounds, it is assumed that, in addition to dissipation, wave breaking produces a spectral energy flux directed toward low wavenumbers. I show that it is then possible to remove the energy and momentum budget inconsistency, and correspondingly the required strength of this spectral flux is estimated. Introducing such a downward spectral flux permits fulfilling both energy and momentum balance conditions. Meanwhile, the consistency between the transfer equation and empirical spectra, estimated by means of a cost function K, is either improved or slightly reduced, depending upon the wave breaking and wind source terms chosen. Other tests are performed in which it is further assumed that wave breaking would also be associated with azimuthal diffusion of the spectral energy. This would correspondingly reduce the required downward spectral flux by a factor of up to 5, although it would not be able to remove it entirely.

  13. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  14. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    CERN Document Server

    Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P

    2002-01-01

    A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.

  15. Long-term monitoring of airborne nickel (Ni) pollution in association with some potential source processes in the urban environment.

    Science.gov (United States)

    Kim, Ki-Hyun; Shon, Zang-Ho; Mauulida, Puteri T; Song, Sang-Keun

    2014-09-01

    The environmental behavior and pollution status of nickel (Ni) were investigated in seven major cities in Korea over a 13-year time span (1998-2010). The mean concentrations of Ni measured during the whole study period fell within the range of 3.71 (Gwangju: GJ) to 12.6ngm(-3) (Incheon: IC). Although Ni values showed a good comparability in a relatively large spatial scale, its values in most cities (6 out of 7) were subject to moderate reductions over the study period. To assess the effect of major sources on the long-term distribution of Ni, the relationship between their concentrations and the potent source processes like non-road transportation sources (e.g., ship and aircraft emissions) were examined from some cities with port and airport facilities. The potential impact of long-range transport of Asian dust particles in controlling Ni levels was also evaluated. The overall results suggest that the Ni levels were subject to gradual reductions over the study period irrespective of changes in such localized non-road source activities. The pollution of Ni at all the study sites was maintained well below the international threshold (Directive 2004/107/EC) value of 20ngm(-3).

  16. Extraction of temporal networks from term co-occurrences in online textual sources.

    Directory of Open Access Journals (Sweden)

    Marko Popović

    Full Text Available A stream of unstructured news can be a valuable source of hidden relations between different entities, such as financial institutions, countries, or persons. We present an approach to continuously collect online news, recognize relevant entities in them, and extract time-varying networks. The nodes of the network are the entities, and the links are their co-occurrences. We present a method to estimate the significance of co-occurrences, and a benchmark model against which their robustness is evaluated. The approach is applied to a large set of financial news, collected over a period of two years. The entities we consider are 50 countries which issue sovereign bonds, and which are insured by Credit Default Swaps (CDS in turn. We compare the country co-occurrence networks to the CDS networks constructed from the correlations between the CDS. The results show relatively small, but significant overlap between the networks extracted from the news and those from the CDS correlations.

  17. Measurement and apportionment of radon source terms for modeling indoor environments

    Energy Technology Data Exchange (ETDEWEB)

    Harley, N.H.

    1992-01-01

    During the present 2 1/2 year contract period, we have made significant Progress in modeling the source apportionment of indoor [sup 222]Rn and in [sup 222]Rn decay product dosimetry. Two additional areas were worked on which we believe are useful for the DOE Radon research Program. One involved an analysis of the research house data, grouping the hourly house [sup 222]Rn measurements into 2 day, 7 day and 90 day intervals to simulate the response of passive monitors. Another area requiring some attention resulted in a publication of 3 years of our indoor/outdoor measurements in a high-rise apartment. Little interest has been evinced in apartment measurements yet 20% of the US population lives in multiple-family dwellings, not in contact with the ground. These data together with a summary of all other published data on apartments showed that apartments have only about 50% greater [sup 222]Rn concentration than the measured outdoor [sup 222]Rn. Apartment dwellers generally represent a low risk group regarding [sup 222]Rn exposure. The following sections describe the main projects in some detail.

  18. Measurement and apportionment of radon source terms for modeling indoor environments

    International Nuclear Information System (INIS)

    During the present 2 1/2 year contract period, we have made significant Progress in modeling the source apportionment of indoor 222Rn and in 222Rn decay product dosimetry. Two additional areas were worked on which we believe are useful for the DOE Radon research Program. One involved an analysis of the research house data, grouping the hourly house 222Rn measurements into 2 day, 7 day and 90 day intervals to simulate the response of passive monitors. Another area requiring some attention resulted in a publication of 3 years of our indoor/outdoor measurements in a high-rise apartment. Little interest has been evinced in apartment measurements yet 20% of the US population lives in multiple-family dwellings, not in contact with the ground. These data together with a summary of all other published data on apartments showed that apartments have only about 50% greater 222Rn concentration than the measured outdoor 222Rn. Apartment dwellers generally represent a low risk group regarding 222Rn exposure. The following sections describe the main projects in some detail

  19. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-09-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  20. Estimating Uncertainty in Long Term Total Ozone Records from Multiple Sources

    Science.gov (United States)

    Frith, Stacey M.; Stolarski, Richard S.; Kramarova, Natalya; McPeters, Richard D.

    2014-01-01

    Total ozone measurements derived from the TOMS and SBUV backscattered solar UV instrument series cover the period from late 1978 to the present. As the SBUV series of instruments comes to an end, we look to the 10 years of data from the AURA Ozone Monitoring Instrument (OMI) and two years of data from the Ozone Mapping Profiler Suite (OMPS) on board the Suomi National Polar-orbiting Partnership satellite to continue the record. When combining these records to construct a single long-term data set for analysis we must estimate the uncertainty in the record resulting from potential biases and drifts in the individual measurement records. In this study we present a Monte Carlo analysis used to estimate uncertainties in the Merged Ozone Dataset (MOD), constructed from the Version 8.6 SBUV2 series of instruments. We extend this analysis to incorporate OMI and OMPS total ozone data into the record and investigate the impact of multiple overlapping measurements on the estimated error. We also present an updated column ozone trend analysis and compare the size of statistical error (error from variability not explained by our linear regression model) to that from instrument uncertainty.