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Sample records for maccs reactor accident

  1. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion.

  2. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk.

  3. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, water ingestion dose, milk growing season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meat, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of 1-131 at which disposal of crops will be initiated due to accidents that occur during the growing season.

  4. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. ); Rollstin, J.A. ); Chanin, D.I. )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  5. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect

    Chanin, D.I. ); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  6. MACCS usage at Rocky Flats Plant for consequence analysis of postulated accidents

    SciTech Connect

    Foppe, T.L.; Peterson, V.L.

    1993-10-01

    The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and health effects to the public from postulated plutonium releases and from postulated criticalities. These applications were conducted to support deterministic and probabilistic accident analyses for safety analyses for safety analysis reports, radiological sabotage studies, and other regulatory requests.

  7. A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications

    SciTech Connect

    Young, M.

    1995-02-01

    MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC`s probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions.

  8. Reactor Accident Consequence Code

    SciTech Connect

    2015-11-02

    MACCS1.5 performs probabilistic calculations of potential off site consequences of the atmospheric releases of radioactive material in reactor accidents. The principal phenomena considered in MACCS are atmospheric transport, environmental contamination, emergency response, long term mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. MACCS can be used for a variety of applications including probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost benefit analysis. The time scale after the accident is divided into three phases: emergency, intermediate, and long term. The region surrounding the reactor is divided into a polar-coordinate grid, with the reactor located at the center, for the calculations. Two preprocessors, MAXGC and DOSFAC, are included. MAXGC generates the maximum allowable ground concentrations based on protective action guide (PAG) dose levels. DOSFAC generates the dose conversion data used by MACCS.

  9. Reactor Accident Consequence Code

    Energy Science and Technology Software Center (ESTSC)

    2015-11-02

    MACCS1.5 performs probabilistic calculations of potential off site consequences of the atmospheric releases of radioactive material in reactor accidents. The principal phenomena considered in MACCS are atmospheric transport, environmental contamination, emergency response, long term mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. MACCS can be used for a variety of applications including probabilistic risk assessment (PRA) ofmore » nuclear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost benefit analysis. The time scale after the accident is divided into three phases: emergency, intermediate, and long term. The region surrounding the reactor is divided into a polar-coordinate grid, with the reactor located at the center, for the calculations. Two preprocessors, MAXGC and DOSFAC, are included. MAXGC generates the maximum allowable ground concentrations based on protective action guide (PAG) dose levels. DOSFAC generates the dose conversion data used by MACCS.« less

  10. Input-output model for MACCS nuclear accident impacts estimation¹

    SciTech Connect

    Outkin, Alexander V.; Bixler, Nathan E.; Vargas, Vanessa N

    2015-01-27

    Since the original economic model for MACCS was developed, better quality economic data (as well as the tools to gather and process it) and better computational capabilities have become available. The update of the economic impacts component of the MACCS legacy model will provide improved estimates of business disruptions through the use of Input-Output based economic impact estimation. This paper presents an updated MACCS model, bases on Input-Output methodology, in which economic impacts are calculated using the Regional Economic Accounting analysis tool (REAcct) created at Sandia National Laboratories. This new GDP-based model allows quick and consistent estimation of gross domestic product (GDP) losses due to nuclear power plant accidents. This paper outlines the steps taken to combine the REAcct Input-Output-based model with the MACCS code, describes the GDP loss calculation, and discusses the parameters and modeling assumptions necessary for the estimation of long-term effects of nuclear power plant accidents.

  11. Quality assurance and verification of the MACCS (MELCOR Accident Consequence Code System) code, Version 1. 5

    SciTech Connect

    Dobbe, C.A.; Carlson, E.R.; Marshall, N.H.; Marwil, E.S.; Tolli, J.E. )

    1990-02-01

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs.

  12. Documentation for RISKIN: A risk integration code for MACCS (MELCOR Accident Consequence Code System) output

    SciTech Connect

    Rollstin, J.A. ); Hong, Kou-John )

    1990-11-01

    This document has been prepared as a user's guide for the computer program RISKIN developed at Sandia National Laboratories. The RISKIN code generates integrated risk tables and the weighted mean risk associated with a user-selected set of consequences from up to five output files generated by the MELCOR Accident Consequence Code System (MACCS). Each MACCS output file can summarize the health and economic consequences resulting from up to 60 distinct severe accident source terms. Since the accident frequency associated with these source terms is not included as a MACCS input parameter a postprocessor is required to derived results that must incorporate accident frequency. The RISKIN code is such a postprocessor. RISKIN will search the MACCS output files for the mean and peak consequence values and the complementary cumulative distributive function (CCDF) tables for each requested consequence. Once obtained, RISKIN combines this data with accident frequency data to produce frequency weighted results. A postprocessor provides RISKIN an interface to the proprietary DISSPLA plot package. The RISKIN code has been written using ANSI Standard FORTRAN 77 to maximize its portability.

  13. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    SciTech Connect

    Neymotin, L.

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.

  14. MACCS2 development and verification efforts

    SciTech Connect

    Young, M.; Chanin, D.

    1997-03-01

    MACCS2 represents a major enhancement of the capabilities of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, released in 1987, was developed to estimate the potential impacts to the surrounding public of severe accidents at nuclear power plants. The principal phenomena considered in MACCS/MACCS2 are atmospheric transport and deposition under time-variant meteorology, short-term and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. MACCS2 was developed as a general-purpose analytical tool applicable to diverse reactor and nonreactor facilities. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. In addition, errors that had been identified in MACCS version1.5.11.1 were corrected, including an error that prevented the code from providing intermediate-phase results. MACCS2 version 1.10 beta test was released to the beta-test group in May, 1995. In addition, the University of New Mexico (UNM) has completed an independent verification study of the code package. Since the beta-test release of MACCS2 version 1.10, a number of minor errors have been identified and corrected, and a number of enhancements have been added to the code package. The code enhancements added since the beta-test release of version 1.10 include: (1) an option to allow the user to input the {sigma}{sub y} and {sigma}{sub z} plume expansion parameters in a table-lookup form for incremental downwind distances, (2) an option to define different initial dimensions for up to four segments of a release, (3) an enhancement to the COMIDA2 food-chain model preprocessor to allow the user to supply externally calculated tables of tritium food-chain dose per unit deposition on farmland to support analyses of tritium releases, and (4) the capability to calculate direction-dependent doses.

  15. A simplified model for calculating early offsite consequences from nuclear reactor accidents

    SciTech Connect

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1988-07-01

    A personal computer-based model, SMART, has been developed that uses an integral approach for calculating early offsite consequences from nuclear reactor accidents. The solution procedure uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast-running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarking and detailed sensitivity/uncertainty analyses using SMART are presented. 34 refs., 21 figs., 24 tabs.

  16. Review of the chronic exposure pathways models in MACCS (MELCOR Accident Consequence Code System) and several other well-known probabilistic risk assessment models

    SciTech Connect

    Tveten, U. )

    1990-06-01

    The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above.

  17. Use of post-Chernobyl data from Norway to validate the long-term exposure pathway models in the accident consequence code MACCS

    SciTech Connect

    Tveten, U. )

    1994-03-01

    This paper describes a task performed for the US Nuclear Regulatory Commission (NRC), consisting of using post-Chernobyl data from Norway to verify or find areas for possible improvement in the chronic exposure pathway models utilized in the NRC's program for probabilistic risk analysis, level 3, of the MELCOR accident consequence code system (MACCS), developed at Sandia National Laboratories, Albuquerque, New Mexico. Because of unfortunate combinations of weather conditions, the levels of Chernobyl fallout in parts of Norway were quite high, with large areas contaminated to more than 100 kBq/m[sup 2] of radioactive cesium. Approximately 6% of the total amount of radioactive cesium released from Chernobyl is deposited on Norwegian territory, according to a countrywide survey performed by the Norwegian National Institute for Radiation Hygiene. Accordingly, a very large monitoring effort was carried out in Norway, and some of the results of this effort have provided important new insights into the ways in which radioactive cesium behaves in the environment. In addition to collection and evaluation of post-Chernobyl monitoring results, some experiments were also performed as part of the task. Some experiments performed pre-Chernobyl were also relevant, and some conclusions could be drawn from these. In most connections, the data available show the models and data in MACCS to be appropriate. A few areas where the data indicate that the MACCS approach is inadequate are, however, also pointed out in the paper.

  18. Code manual for MACCS2: Volume 1, user`s guide

    SciTech Connect

    Chanin, D.I.; Young, M.L.

    1997-03-01

    This report describes the use of the MACCS2 code. The document is primarily a user`s guide, though some model description information is included. MACCS2 represents a major enhancement of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, distributed by government code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the surrounding public. The principal phenomena considered are atmospheric transport and deposition under time-variant meteorology, short- and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. No other U.S. code that is publicly available at present offers all these capabilities. MACCS2 was developed as a general-purpose tool applicable to diverse reactor and nonreactor facilities licensed by the Nuclear Regulatory Commission or operated by the Department of Energy or the Department of Defense. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency-response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. Other improvements are in the areas of phenomenological modeling and new output options. Initial installation of the code, written in FORTRAN 77, requires a 486 or higher IBM-compatible PC with 8 MB of RAM.

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

    SciTech Connect

    Sprung, J.L.; Jow, H-N ); Rollstin, J.A. ); Helton, J.C. )

    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.

  20. Analysis of Credible Accidents for Argonaut Reactors

    SciTech Connect

    Hawley, S. C.; Kathern, R. L.; Robkin, M. A.

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.

  1. Accident analysis for US fast burst reactors

    SciTech Connect

    Paternoster, R.; Flanders, M.; Kazi, H.

    1994-09-01

    In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards.

  2. (Severe accident technology of BWR (Boiling Water Reactor) reactors)

    SciTech Connect

    Ott, L.J.

    1989-10-23

    The traveler attended the 1989 CORA Workshop at KfK, FRG. Participation included the presentation included the presentation of three papers on work performed by the Boiling Water Reactor Severe Accident Technology (BWRSAT) program at Oak Ridge National Laboratory (ORNL) in Boiling Water Reactor (BWR) severe accident analyses. The Statement of Work (June 1989) for the BWRSAT Program provides for code analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that BWRSAT personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to arrange for acquisition of detailed CORA facility drawings, experimental data, and related engineering. 17 refs.

  3. On-site worker-risk calculations using MACCS

    SciTech Connect

    Peterson, V.L.

    1993-05-01

    We have revised the latest version of MACCS for use with the calculation of doses and health risks to on-site workers for postulated accidents at the Rocky Flats Plant (RFP) in Colorado. The modifications fall into two areas: (1) an improved estimate of shielding offered by buildings to workers that remain indoors; and, (2) an improved treatment of building-wake effects, which affects both indoor and outdoor workers. Because the postulated accident can be anywhere on plant site, user-friendly software has been developed to create those portions of the (revised) MACCS input data files that are specific to the accident site.

  4. Radioactive fallout from the Chernobyl nuclear reactor accident

    SciTech Connect

    Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

    1987-08-01

    This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US. (TEM)

  5. Loss of pumping accident limit calculation for Savannah River Reactor

    SciTech Connect

    Paul, P.K.; Barbour, K.L. )

    1992-01-01

    For the Savannah River Site production reactors, the design basis accident reactor power limit ensures that if a double-ended guillotine break (DEGB) in a secondary cooling water pipe were to occur, the reactor will shut down safely. The primary reactor coolant is heavy water (D{sub 2}O) with secondary light water (H{sub 2}O) cooling. The accident scenario is a DEGB in one of two secondary coolant inlet header pipes with several assumed single failures. The recycled primary coolant loses its cooling, and the reactor core temperature begins to rise. Another possible accident is a DEGB in one of two heat exchanger secondary coolant effluent header pipes. The inlet header break is slightly more limiting than the effluent header break. Upon break detection, emergency shutdown begins and the emergency cooling system (ECS) activates. The accident scenario was constructed with regard to physical, mechanical, and human factors. The computer code TRAC simulates the accident.

  6. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  7. PKL reactor tank bottom pressures in accident scenarios

    SciTech Connect

    Tudor, A.A.

    1987-03-10

    Nuclear Engineering Division requested estimates of the maximum PKL reactor tank pressures associated with postulated reactor accidents. Tank bottom pressures calculated in establishing confinement protection limits (CPL) in Mark 16B-31 and Mark 22 reactor charges are given in this document.

  8. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

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

    SciTech Connect

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. ); Medford, G.T. )

    1990-06-01

    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.

  10. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  11. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  12. Global risk of radioactive fallout after major nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2012-05-01

    Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  13. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  14. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    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.

  15. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-09

    ... COMMISSION Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors..., ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' DATES: Submit comments by... No. ML081430087) concerning the review of probabilistic risk assessment (PRA) information and...

  16. Global impact of the Chernobyl reactor accident

    SciTech Connect

    Anspaugh, L.R.; Catlin, R.J.; Goldman, M.

    1988-12-16

    Radioactive material was deposited throughout the Northern Hemisphere as a result of the accident at the Chernobyl Nuclear Power Station on 26 April 1986. On the basis of a large amount of environmental data and new integrated dose assessment and risk models, the collective dose commitment to the approximately 3 billion inhabitants is calculated to be 930,000 person-gray, with 97% in the western Soviet Union and Europe. The best estimates for the lifetime expectation of fatal radiogenic cancer would increase the risk from 0 to 0.02% in Europe and 0 to 0.003% in the Northern Hemisphere. By means of an integration of the environmental data, it is estimated that approximately 100 petabecquerels of cesium-137 (1 PBq = 10(15) Bq) were released during and subsequent to the accident.

  17. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Kunkel, D.; Lelieveld, J.; Lawrence, M. G.

    2012-04-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90 % of emitted 137Cs would be transported beyond 50 km and about 50 % beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  18. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2011-11-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50km and about 50% beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  19. The TOPAZ II space reactor response under accident conditions

    SciTech Connect

    Voss, S.S.

    1993-12-31

    The TOPAZ II is a single-cell thermionic space reactor power system developed by the Russians during the period of time from {approximately}1969 to 1989. The TOPAZ II has never been flight demonstrated, but the system was extensively tested on the ground. As part of the development and test program, the response of the TOPAZ II under accident conditions was analyzed and characterized. The US TOPAZ II team has been working closely with the Russian specialists to understand the TOPAZ II system, its operational characteristics, and its response under potential accident conditions. The purpose of the technical exchange is to enable a potential launch of a TOPAZ II by the US. The information is required to integrate the system with a US spacecraft and to support the safety review process. The purpose of this paper is to provide a brief overview of the system and its response under actual and postulated accident conditions.

  20. Radioactive fallout from the Chernobyl nuclear reactor accident

    SciTech Connect

    Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

    1987-03-23

    Following the accident at the nuclear reactor at Chernobyl, in the Soviet Union on April 26, 1986, we performed a variety of measurements to determine the level of the radioactive fallout on the western United States. We used gamma-spectroscopy to analyze air filters from the areas around Lawrence Livermore National Laboratory (LLNL), California, and Barrow and Fairbanks, Alaska. Milk from California and imported vegetables were also analyzed. The levels of the various fission products detected were far below the maximum permissible concentration levels.

  1. Severe accident sequence assessment for boiling water reactors: program overview

    SciTech Connect

    Fontana, M. H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case.

  2. Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

    SciTech Connect

    Tusheva, P.; Schaefer, F.; Kliem, S.

    2012-07-01

    The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

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

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; 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 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.

  4. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  5. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-06

    ... COMMISSION Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors... comment period. SUMMARY: On October 9, 2012 (77 FR 61446), the U.S. Nuclear Regulatory Commission (NRC or...), Section 19.0 ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' The NRC...

  6. BESAFE II: Accident safety analysis code for MFE reactor designs

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

  7. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  8. Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

    SciTech Connect

    Kroeger, P.G.; Chan, B.C.

    1985-01-01

    Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the CB atmosphere temperature and pressure, as overpressurization is the most likely failure mode.

  9. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    SciTech Connect

    Arcieri, W.C.; Hanson, D.J. )

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents.

  10. ALTERNATIVES OF MACCS2 IN LANL DISPERSION ANALYSIS FOR ONSITE AND OFFSITE DOSES

    SciTech Connect

    Wang, John HC

    2012-05-01

    In modeling atmospheric dispersion to determine accidental release of radiological material, one of the common statistical analysis tools used at Los Alamos National Laboratory (LANL) is MELCOR Accident Consequence Code System, Version 2 (MACCS2). MACCS2, however, has some limitations and shortfalls for both onsite and offsite applications. Alternative computer codes, which could provide more realistic calculations, are being investigated for use at LANL. In the Yucca Mountain Project (YMP), the suitability of MACCS2 for the calculation of onsite worker doses was a concern; therefore, ARCON96 was chosen to replace MACCS2. YMP's use of ARCON96 provided results which clearly demonstrated the program's merit for onsite worker safety analyses in a wide range of complex configurations and scenarios. For offsite public exposures, the conservatism of MACCS2 on the treatment of turbulence phenomena at LANL is examined in this paper. The results show a factor of at least two conservatism in calculated public doses. The new EPA air quality model, AERMOD, which implements advanced meteorological turbulence calculations, is a good candidate for LANL applications to provide more confidence in the accuracy of offsite public dose projections.

  11. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  12. A Web Server for MACCS Magnetometer Data

    NASA Technical Reports Server (NTRS)

    Engebretson, Mark J.

    1998-01-01

    NASA Grant NAG5-3719 was provided to Augsburg College to support the development of a web server for the Magnetometer Array for Cusp and Cleft Studies (MACCS), a two-dimensional array of fluxgate magnetometers located at cusp latitudes in Arctic Canada. MACCS was developed as part of the National Science Foundation's GEM (Geospace Environment Modeling) Program, which was designed in part to complement NASA's Global Geospace Science programs during the decade of the 1990s. This report describes the successful use of these grant funds to support a working web page that provides both daily plots and file access to any user accessing the worldwide web. The MACCS home page can be accessed at http://space.augsburg.edu/space/MaccsHome.html.

  13. Thyroid Consequences of the Fukushima Nuclear Reactor Accident

    PubMed Central

    Nagataki, Shigenobu

    2012-01-01

    Background A special report, ‘The Fukushima Accident’, was delivered at the 35th Annual Meeting of the European Thyroid Association in Krakow on September 11, 2011, and this study is the follow-up of the special report. Objectives To present a preliminary review of potential thyroid consequences of the 2011 Fukushima nuclear reactor accident. Methods Numerous new data have been presented in Japanese, and most of them are available on the website from each research institute and/or from each municipality. The review was made using these data from the website. Results When individual radiation doses were expressed as values in more than 99% of residents, radiation doses by behavior survey in evacuation and deliberate evacuation areas were less than 10 mSv in the first 4 months, and internal radiation doses measured by whole body counters were less than 1 mSv/year. Individual thyroid radiation doses were less than 50 mSv (intervention levels) even in evacuation areas. As for health consequences, no one died and no one suffered from acute effects. The thyroid ultrasound examination is in progress and following examination of almost 40,000 children, 35% of them have nodules and/or cysts but no cancers. Conclusions Countermeasures against radiation must consider current individual measured values, although every effort must be taken to reconstruct radiation doses as precisely as possible. At present, the difference of thyroid radiation dose between Chernobyl and Fukushima appears to be due to the strict control of milk started within a week after the accident in Fukushima. Since the iodine-131 plume moved around in wide areas and for a long time, the method of thyroid protection must be reconsidered. PMID:24783014

  14. Metal-fueled HWR (heavy water reactors) severe accident issues: Differences and similarities to commercial LWRs (light water reactors)

    SciTech Connect

    Ellison, P.G.; Hyder, M.L.; Monson, P.R. ); Coryell, E.W. )

    1990-01-01

    Differences and similarities in severe accident progression and phenomena between commercial Light Water Reactors (LWR) and metal-fueled isotopic production Heavy Water Reactors (HWR) are described. It is very important to distinguish between accident progression in the two systems because each reactor type behaves in a unique manner to a fuel melting accident. Some of the lessons learned as a result of the extensive commercial severe accident research are not applicable to metal-fueled heavy water reactors. A direct application of severe accident phenomena developed from oxide-fueled LWRs to metal-fueled HWRs may lead to large errors or substantial uncertainties. In general, the application of severe accident LWR concepts to HWRs should be done with the intent to define the relevant issues, define differences, and determine areas of overlap. This paper describes the relevant differences between LWR and metal-fueled HWR severe accident phenomena. Also included in the paper is a description of the phenomena that govern the source term in HWRs, the areas where research is needed to resolve major uncertainties, and areas in which LWR technology can be directly applied with few modifications.

  15. Code System for Calculating Early Offsite Consequences from Nuclear Reactor Accidents.

    Energy Science and Technology Software Center (ESTSC)

    1992-06-10

    SMART calculates early offsite consequences from nuclear reactor accidents. Once the air and ground concentrations of the radionuclide are estimated, the early dose to an individual is calculated via three pathways: cloudshine, short-term groundshine, and inhalation.

  16. Contamination of surface-water bodies after reactor accidents by the erosion of atmospherically deposited radionuclides.

    PubMed

    Helton, J C; Muller, A B; Bayer, A

    1985-06-01

    Reactor safety analyses usually do not consider the population risk which might result from the contamination of surface-water bodies after reactor accidents by the erosion of atmospherically deposited radionuclides. This paper is intended to provide perspective on the reasonableness of this omission. Data are presented which are suggestive of the rates at which atmospherically deposited radionuclides might erode into surface-water bodies. These rates are used in the calculation of potential health effects resulting from surface-water contamination due to such erosion. These health effects are compared with predicted health effects due to atmospheric and terrestrial pathways after reactor accidents. The presented results support the belief that the contamination of surface-water bodies after reactor accidents by the erosion of atmospherically deposited radionuclides is not a major contributor to the risk associated with such accidents. PMID:3997527

  17. Action Plan for updated Chapter 15 Accident Analysis in the SRS Production Reactor SAR

    SciTech Connect

    Hightower, N.T. III; Burnett, T.W.

    1989-11-15

    This report describes the Action Plan for the upgrade of the Chapter 15 Accident Analysis in the SRS Production Reactor SAR required for K-Restart. This Action Plan will be updated periodically to reflect task accomplishments and issue resolutions.

  18. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  19. An idealized transient model for melt dispersal from reactor cavities during pressurized melt ejection accident scenarios

    SciTech Connect

    Tutu, N.K.

    1991-06-01

    The direct Containment Heating (DCH) calculations require that the transient rate at which the melt is ejected from the reactor cavity during hypothetical pressurized melt ejection accident scenarios be calculated. However, at present no models, that are able to predict the available melt dispersal data from small scale reactor cavity models, are available. In this report, a simple idealized model of the melt dispersal process within a reactor cavity during a pressurized melt ejection accident scenario is presented. The predictions from the model agree reasonably well with the integral data obtained from the melt dispersal experiments using a small scale model of the Surry reactor cavity. 17 refs., 15 figs.

  20. Thermalhydraulic processes in the reactor coolant system of a BWR (boiling water reactor) under severe accident conditions

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MSIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a brief discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermalhydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are briefly described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is presented. 12 refs., 4 figs., 1 tab.

  1. Implications for accident management of adding water to a degrading reactor core

    SciTech Connect

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J.

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  2. The health impact of major nuclear accidents: the case of Greece.

    PubMed

    Kollas, J G

    1993-10-01

    An assessment of the radiological consequences that would result for the population of Greece from postulated major nuclear accidents in the Kozloduy nuclear power station in Bulgaria is performed. Kozloduy lies at a distance of 225 km from the northern borders of Greece and contains six reactors, all of the Russian WWER type. The postulated accidents that are classified as level 7 accidents on the International Nuclear Event Scale, involve significant releases of radioactive materials into the environment, and widespread health and environmental effects. The analysis is performed by the MACCS code. The estimated consequences are compared to the corresponding actual impact of the Chernobyl accident in Greece. The results of the analysis indicate that, under the conservative assumptions adopted, the radiological consequences of the most severe accidents considered would be about 1.5 orders of magnitude larger than the actual radiological consequences of the Chernobyl accident. PMID:8259439

  3. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema

    None

    2014-03-11

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  4. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    SciTech Connect

    2013-09-25

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  5. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  6. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  7. Guidelines for Exposure Assessment in Health Risk Studies Following a Nuclear Reactor Accident

    PubMed Central

    Bouville, André; Linet, Martha S.; Hatch, Maureen; Mabuchi, Kiyohiko

    2013-01-01

    Background: Worldwide concerns regarding health effects after the Chernobyl and Fukushima nuclear power plant accidents indicate a clear need to identify short- and long-term health impacts that might result from accidents in the future. Fundamental to addressing this problem are reliable and accurate radiation dose estimates for the affected populations. The available guidance for activities following nuclear accidents is limited with regard to strategies for dose assessment in health risk studies. Objectives: Here we propose a comprehensive systematic approach to estimating radiation doses for the evaluation of health risks resulting from a nuclear power plant accident, reflected in a set of seven guidelines. Discussion: Four major nuclear reactor accidents have occurred during the history of nuclear power production. The circumstances leading to these accidents were varied, as were the magnitude of the releases of radioactive materials, the pathways by which persons were exposed, the data collected afterward, and the lifestyle factors and dietary consumption that played an important role in the associated radiation exposure of the affected populations. Accidents involving nuclear reactors may occur in the future under a variety of conditions. The guidelines we recommend here are intended to facilitate obtaining reliable dose estimations for a range of different exposure conditions. We recognize that full implementation of the proposed approach may not always be feasible because of other priorities during the nuclear accident emergency and because of limited resources in manpower and equipment. Conclusions: The proposed approach can serve as a basis to optimize the value of radiation dose reconstruction following a nuclear reactor accident. Citation: Bouville A, Linet MS, Hatch M, Mabuchi K, Simon SL. 2014. Guidelines for exposure assessment in health risk studies following a nuclear reactor accident. Environ Health Perspect 122:1–5; http://dx.doi.org/10

  8. Estimation Of 137Cs Using Atmospheric Dispersion Models After A Nuclear Reactor Accident

    NASA Astrophysics Data System (ADS)

    Simsek, V.; Kindap, T.; Unal, A.; Pozzoli, L.; Karaca, M.

    2012-04-01

    Nuclear energy will continue to have an important role in the production of electricity in the world as the need of energy grows up. But the safety of power plants will always be a question mark for people because of the accidents happened in the past. Chernobyl nuclear reactor accident which happened in 26 April 1986 was the biggest nuclear accident ever. Because of explosion and fire large quantities of radioactive material was released to the atmosphere. The release of the radioactive particles because of accident affected not only its region but the entire Northern hemisphere. But much of the radioactive material was spread over west USSR and Europe. There are many studies about distribution of radioactive particles and the deposition of radionuclides all over Europe. But this was not true for Turkey especially for the deposition of radionuclides released after Chernobyl nuclear reactor accident and the radiation doses received by people. The aim of this study is to determine the radiation doses received by people living in Turkish territory after Chernobyl nuclear reactor accident and use this method in case of an emergency. For this purpose The Weather Research and Forecasting (WRF) Model was used to simulate meteorological conditions after the accident. The results of WRF which were for the 12 days after accident were used as input data for the HYSPLIT model. NOAA-ARL's (National Oceanic and Atmospheric Administration Air Resources Laboratory) dispersion model HYSPLIT was used to simulate the 137Cs distrubition. The deposition values of 137Cs in our domain after Chernobyl Nuclear Reactor Accident were between 1.2E-37 Bq/m2 and 3.5E+08 Bq/m2. The results showed that Turkey was affected because of the accident especially the Black Sea Region. And the doses were calculated by using GENII-LIN which is multipurpose health physics code.

  9. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab.

  10. Analysis of reactivity-insertion accidents in the TREAT Upgrade reactor

    SciTech Connect

    Rudolph, R.R.; Bhattacharyya, S.K.

    1983-01-01

    The expansion of the experimental capabilities of the TREAT Upgrade (TU) reactor also tends to increase the potential risks associated with off-normal reactivity insertion incidents compared to the TREAT reactor. To provide adequate prtection for the public and the facility, while meeting experimenter's requirements, a specialized Reactor Trip System (RTS) with energy-dependent scram trips on reactor power and period has been developed. With this protection strategy, the consequences of reactivity insertion accidents in the TU reactor have been analyzed using a general methodology developed earlier. Results of these analyses are presented.

  11. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  12. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  13. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Cleveland, J.; Krueger, K.; Kernforschungsanlage Juelich G.m.b.H. . Arbeitsgemeinschaft Versuchsreaktor)

    1989-01-01

    Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have been conducted with the German pebble-bed High-Temperature Reactor AVR. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The tests demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into modular HTGR designs. The main LOCA test lasted for 5 d. After the test began, core temperatures increased for {approximately}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded accident levels of decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components. 3 refs., 9 figs., 1 tab.

  14. Estimates of the financial consequences of nuclear-power-reactor accidents

    SciTech Connect

    Strip, D.R.

    1982-09-01

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

  15. Nuclear reactor accidents: Chernobyl, TMI, and Windscale. (Latest citations from Pollution abstracts). Published Search

    SciTech Connect

    1995-12-01

    The bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout; the resultant runoff into rivers, lakes, and the sea; the radiation effects on people; and the transfrontier radioactive contamination of the environment. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  16. Nuclear reactor accidents: Chernobyl, TMI, and windscale. (Latest citations from Pollution Abstracts). Published Search

    SciTech Connect

    Not Available

    1994-11-01

    The bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout; the resultant runoff into rivers, lakes, and the sea; the radiation effects on people; and the transfrontier radioactive contamination of the environment. (Contains 250 citations and includes a subject term index and title list.)

  17. Loss-of-coolant accident experiment at the AVR (Arbeitsgemeinschaft Versuchsreaktor) gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1989-11-01

    Loss of coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into modular gas-cooled reactor designs, loss-of-coolant accident (LOCA) simulation tests were conducted with the 15-MW(electric), 46-MW(thermal), pebble-bed, high-temperature Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany (FRG). This is the only nuclear power plant ever to have been intentionally subjected to LOCa conditions. Oak Ridge National Laboratory participation in the preparation and conduct of the tests was carried out within the U.S./FRG Agreement for Cooperation in Gas-Cooled Reactor Development.

  18. Global Reactive Gases in the MACC project

    NASA Astrophysics Data System (ADS)

    Schultz, M. G.

    2012-04-01

    In preparation for the planned atmospheric service component of the European Global Monitoring for Environment and Security (GMES) initiative, the EU FP7 project Monitoring of Atmospheric Composition and Climate (MACC) developed a preoperational data assimilation and modelling system for monitoring and forecasting of reactive gases, greenhouse gases and aerosols. The project is coordinated by the European Centre for Medium-Range Weather Forecast (ECMWF) and the system is built on ECMWF's Integrated Forecasting System (IFS) which has been coupled to the chemistry transport models MOZART-3 and TM5. In order to provide daily forecasts of up to 96 hours for global reactive gases, various satellite retrieval products for ozone (total column and profile data), CO, NO2, CH2O and SO2 are either actively assimilated or passively monitored. The MACC system is routinely evaluated with in-situ data from ground-based stations, ozone sondes and aircraft measurements, and with independent satellite retrievals. Global MACC reactive gases forecasts are used in the planning and analysis of large international field campaigns and to provide dynamical chemical boundary conditions to regional air quality models worldwide. Several case studies of outstanding air pollution events have been performed, and they demonstrate the strengths and weaknesses of chemical data assimilation based on current satellite data products. Besides the regular analyses and forecasts of the tropospheric chemical composition, the MACC system is also used to monitor the evolution of stratospheric ozone. A comprehensive reanalysis simulation from 2003 to 2010 provides new insights into the interannual variability of the atmospheric chemical composition.

  19. The global impact of the Chernobyl reactor accident.

    PubMed

    Anspaugh, L R; Catlin, R J; Goldman, M

    1988-12-16

    Radioactive material was deposited throughout the Northern Hemisphere as a result of the accident at the Chernobyl Nuclear Power Station on 26 April 1986. On the basis of a large amount of environmental data and new integrated dose assessment and risk models, the collective dose commitment to the approximately 3 billion inhabitants is calculated to be 930,000 person-gray, with 97% in the western Soviet Union and Europe. The best estimates for the lifetime expectation of fatal radiogenic cancer would increase the risk from 0 to 0.02% in Europe and 0 to 0.003% in the Northern Hemisphere. By means of an integration of the environmental data, it is estimated that approximately 100 petabecquerels of cesium-137 (1 PBq = 10(15) Bq) were released during and subsequent to the accident. PMID:3201240

  20. Computer program predicts thermal and flow transients experienced in a reactor loss- of-flow accident

    NASA Technical Reports Server (NTRS)

    Hale, C. J.

    1967-01-01

    Program analyzes the consequences of a loss-of-flow accident in the primary cooling system of a heterogeneous light-water moderated and cooled nuclear reactor. It produces a temperature matrix 36 x 41 /x,y/ which includes fuel surface temperatures relative to the time the pump power was lost.

  1. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  2. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    SciTech Connect

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-12-23

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  3. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    NASA Astrophysics Data System (ADS)

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-12-01

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  4. TRAC loss-of-coolant accident analyses of the Savannah River production reactors

    SciTech Connect

    Lime, J.F.; Motley, F.E. )

    1990-06-01

    TRAC loss-of-coolant accident (LOCA) analyses were performed as part of the independent safety review of the US Department of Energy's Savannah River (SR) production reactors. The double-ended guillotine break in a coolant loop is a design-basis LOCA for the SR reactors. Three break locations were analyzed to determine the worst break location: (1) at the pump-suction flange; (2) at the pump discharge flange; or (3) at the plenum inlet. The plenum-inlet break was shown to be the most severe in terms of minimum flow delivered to each fuel assembly in the reactor core.

  5. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    SciTech Connect

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-03-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of {minus}3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept.

  6. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    SciTech Connect

    Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

    2011-06-21

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE

  7. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  8. Potassium iodide for thyroid blockade in a reactor accident: administrative policies that govern its use.

    PubMed

    Becker, D V; Zanzonico, P

    1997-04-01

    A marked increase in thyroid cancer among young children who were in the vicinity of the Chernobyl nuclear power plant at the time of the 1986 accident strongly suggests a possible causal relationship to the large amounts of radioactive iodine isotopes in the resulting fallout. Although remaining indoors, restricting consumption of locally produced milk and foodstuffs, and evacuation are important strategies in a major breach-of-containment accident, stable potassium iodide (KI) prophylaxis given shortly before or immediately after exposure can reduce greatly the thyroidal accumulation of radioiodines and the resulting radiation dose. Concerns about possible side effects of large-scale, medically unsupervised KI consumption largely have been allayed in light of the favorable experience in Poland following the Chernobyl accident; 16 million persons received single administrations of KI with only rare occurrence of side effects and with a probable 40% reduction in projected thyroid radiation dose. Despite the universal acceptance of KI as an effective thyroid protective agent, supplies of KI in the US are not available for public distribution in the event of a reactor accident largely because government agencies have argued that stockpiling and distribution of KI to other than emergency workers cannot be recommended in light of difficult distribution logistics, problematic administrative issue, and a calculated low cost-effectiveness. However, KI in tablet form is expensive and has a long shelf life, and many countries have largely stockpiles and distribution programs. The World Health Organization recognizes the benefits of stable KI and urges its general availability. At present there are 110 operating nuclear power plants in the US and more than 300 in the rest of the world. These reactors product 17% of the world's electricity and in some countries up to 60-70% of the total electrical energy. Almost all US nuclear power plants have multistage containment

  9. BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1988-01-01

    This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

  10. "What--me worry?" "Why so serious?": a personal view on the Fukushima nuclear reactor accidents.

    PubMed

    Gallucci, Raymond

    2012-09-01

    Infrequently, it seems that a significant accident precursor or, worse, an actual accident, involving a commercial nuclear power reactor occurs to remind us of the need to reexamine the safety of this important electrical power technology from a risk perspective. Twenty-five years since the major core damage accident at Chernobyl in the Ukraine, the Fukushima reactor complex in Japan experienced multiple core damages as a result of an earthquake-induced tsunami beyond either the earthquake or tsunami design basis for the site. Although the tsunami itself killed tens of thousands of people and left the area devastated and virtually uninhabitable, much concern still arose from the potential radioactive releases from the damaged reactors, even though there was little population left in the area to be affected. As a lifelong probabilistic safety analyst in nuclear engineering, even I must admit to a recurrence of the doubt regarding nuclear power safety after Fukushima that I had experienced after Three Mile Island and Chernobyl. This article is my attempt to "recover" my personal perspective on acceptable risk by examining both the domestic and worldwide history of commercial nuclear power plant accidents and attempting to quantify the risk in terms of the frequency of core damage that one might glean from a review of operational history. PMID:22394214

  11. MACCS version 1.5.11.1: A maintenance release of the code

    SciTech Connect

    Chanin, D.; Foster, J.; Rollstin, J.; Miller, L.

    1993-10-01

    A new version of the MACCS code (version 1.5.11.1) has been developed by Sandia National Laboratories under sponsorship of the US Nuclear Regulatory Commission. MACCS was developed to support evaluations of the off-site consequences from hypothetical severe accidents at commercial power plants. MACCS is the only current public domain code in the US that embodies all of the following modeling capabilities: (1) weather sampling using a year of recorded weather data; (2) mitigative actions such as evacuation, sheltering, relocation, decontamination, and interdiction; (3) economic costs of mitigative actions; (4) cloudshine, groundshine, and inhalation pathways as well as food and water ingestion; (5) calculation of both individual and societal doses to various organs; and (6) calculation of both acute (nonstochastic) and latent (stochastic) health effects and risks of health effects. All of the consequence measures may be fun generated in the form of a complementary cumulative distribution function (CCDF). The current version implements a revised cancer model consistent with recent reports such as BEIR V and ICRP 60. In addition, a number of error corrections and portability enhancements have been implemented. This report describes only the changes made in creating the new version. Users of the code will need to obtain the code`s original documentation, NUREG/CR-4691.

  12. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  13. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

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

    SciTech Connect

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

    1991-12-31

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

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

    SciTech Connect

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

    1991-01-01

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

  16. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    SciTech Connect

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

  17. Generation IV reactors and the ASTRID prototype: Lessons from the Fukushima accident

    NASA Astrophysics Data System (ADS)

    Gauché, François

    2012-05-01

    In France, the ASTRID prototype is a sodium-cooled fast neutron industrial demonstrator, fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for 3rd generation reactors, and take into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, take advantage of the large thermal inertia of SFRs for decay heat removal, and provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place.

  18. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Yoder, G.L. ); Wendel, M.W. )

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  19. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1990-01-01

    A landmark safety test has been conducted at the AVR-reactor, a high-temperature gas-cooled reactor (HTGR) in the Federal Republic of Germany owned by the Arbeitsgemeinschaft Versuchsreaktor, AVR in Juelich. The 46-MW(t), 15-MW(e) AVR reactor was subjected to a simulated loss-of-coolant accident (LOCA), a very severe occurrence in which the coolant escapes from the reactor core and no emergency system provides coolant flow to the core. The test, which demonstrated the inherently safe response of this reactor to a LOCA, marked the first time ever that a reactor has been intentionally subjected to loss-of-coolant conditions without emergency cooling. Oak Ridge National Laboratory (ORNL) and General Atomics participated in the test by working with AVR staff by jointly performing the analyses needed to obtain the license to conduct the test and by performing post test analyses. This participation was carried out under the cooperative AVR Subprogram which is conducted within the US/FRG Agreement for Cooperation in Gas-Cooled Reactor Development. 7 figs.

  20. Modeling & analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    SciTech Connect

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

    1992-10-01

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied.

  1. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  2. Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

    SciTech Connect

    White, J.E.; Roussin, R.W.; Gilpin, H.

    1988-12-01

    A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.

  3. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  4. A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

    SciTech Connect

    Brad J. Merrill; Richard L. Moore; J. Phillip Sharp

    2008-09-01

    A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium-steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.

  5. Using cost/risk procedures to establish recovery criteria following a nuclear reactor accident.

    PubMed

    Tawil, J J; Strenge, D L

    1987-02-01

    In the event of a major accidental release of radionuclides at a nuclear power plant, large populated areas could become seriously contaminated. Local officials would be responsible for establishing radiation recovery criteria that would permit the evacuated population to return safely to their jobs and homes. The range of acceptable criteria could imply variations in property losses in the billions of dollars. Given the likely public concern over the health consequences and the enormity of the potential property losses, a cost/risk analysis can provide important input to establishing the recovery criteria. This paper describes procedures for conducting a cost/risk analysis of a site radiologically contaminated by a nuclear power plant accident. The procedures are illustrated by analyzing a hypothetically contaminated site, using software developed for determining the property and health effects of major reactor accidents. PMID:3818283

  6. Large break loss-of-coolant accident analyses for the high flux isotope reactor

    SciTech Connect

    Taleyarkhan, R.P. )

    1989-01-01

    The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

  7. Chlorine Release from Hypalon Cable Insulation During Severe Nuclear Reactor Accidents

    SciTech Connect

    Auvinen, Ari; Zilliacus, Riitta; Jokiniemi, Jorma

    2005-02-15

    Pyrolytic dehydrochlorination of the electrical cable insulation Hypalon was studied as a function of time and temperature. The chlorine evolution was determined separately by means of on-line activity measurements and by neutron activation analysis in the temperature range 200 deg. C to 300 deg. C, with one test conducted at 500 deg. C. The object of the research was to determine the chlorine release and the chlorine release fraction as a function of temperature. The data obtained were needed to formulate conclusions regarding the release mechanisms of chlorine. Estimates of the amount of hydrochloric acid released to the containment building in a severe reactor accident were also calculated. It can be concluded that the amount of chlorine release from the Hypalon cable is significant and will have an effect on iodine behavior in a severe accident.

  8. Thermodynamic analysis of cesium and iodine behavior in severe light water reactor accidents

    NASA Astrophysics Data System (ADS)

    Minato, Kazuo

    1991-11-01

    In order to understand the release and transport behavior of cesium (Cs) and iodine (I) in severe light water reactor accidents, chemical forms of Cs and I in steam-hydrogen mixtures were analyzed thermodynamically. In the calculations reactions of boron (B) with Cs were taken into consideration. The analysis showed that B plays an important role in determining chemical forms of Cs. The main Cs-containing species are CsBO 2(g) and CsBO 2(l), depending on temperature. The contribution of CsOH(g) is minor. The main I-containing species are HI(g) and CsI(g) over the wide ranges of the parameters considered. Calculations were also carried out under the conditions of the Three Mile Island Unit 2 accident.

  9. Radionuclide monitoring in Northern Ireland of the Chernobyl nuclear reactor accident

    PubMed Central

    Gilmore, B J; Cranley, K

    1987-01-01

    Northern Ireland received higher radiation doses due to the radionuclide contamination from the Chernobyl nuclear reactor accident than did the south of England. Levels of radioactive iodine (131I) and caesium (137Cs) in cows' milk in Northern Ireland increased to 166 and 120 Bq/l respectively in May 1986, but had decreased by factors of one million, and of twenty-five, respectively, by 1 September 1986. The resultant radiation doses represent less than one per cent of those received by a Northern Ireland individual over a period of 40 years from natural background radiation sources. The added risk to any individual from the Chernobyl accident will therefore be very small and may best be judged in the context of the enormously greater risk of death due to potentially preventable diseases, such as smoking-related lung cancer, and coronary heart disease. PMID:3590387

  10. Air Ingress Accident in a High Temperature Reactor with Prismatic Fuel

    SciTech Connect

    Haque, H.; Brinkmann, G.

    2006-07-01

    In this paper, the safety behavior of the new generation high temperature reactors (HTRs) with prismatic fuels during air ingress accident conditions has been investigated. These reactors conceived primarily for the production of hydrogen, are characterized by their inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. Air ingress is an HTR specific event. The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 deg. C leading to reaction heat and graphite corrosion. A substantial amount of graphite burn-off can take place only if sufficient amount of air enters into the core. In order to better assess the phenomena of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is determined as a function of the break cross sections exposed above and below the core. This paper demonstrates the thermal behavior of the ANTARES reactor (operating inlet/outlet temperatures 450/850 deg. C) for various air flow rates with respect to graphite burn-off and maximum temperatures of fuel and bottom reflector region. It indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the pellets exposed. (authors)

  11. Radiocesium levels measured in breast milk one year after the reactor accident at Chernobyl

    SciTech Connect

    Assimakopoulos, P.A.; Ioannides, K.G.; Pakou, A.A.; Lolis, D.; Zikopoulos, K.; Dusias, B.

    1989-01-01

    One hundred-two samples of colostral milk, collected during spring of 1987, approximately one year after the reactor accident at Chernobyl, were measured for radiocesium contamination. The data showed a normal-type distribution with a mean contamination concentration of 16.4 Bq L-1. A weak correlation of the data to the mothers' diet was established by taking into account four of the main staples in the area. The corresponding transfer coefficient was deduced with a value of fm = 0.06 +/- 0.03 d L-1. The resultant effective dose received by breast-feeding infants was estimated, on the average, as 0.012 mrem d-1.

  12. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    SciTech Connect

    Al-Falahi, A.; Haennine, M.; Porkholm, K.

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  13. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect

    Rebak, Raul B.

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  14. Accident sequence analysis for a BWR (Boiling Water Reactor) during low power and shutdown operations

    SciTech Connect

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs.

  15. Sensitivity studies of loss-of-coolant accidents in the Savannah River production reactors

    SciTech Connect

    Edwards, J.N.; Motley, F.E.; Morgan, M.M.; Knight, T.D.; Fischer, S.R. )

    1990-01-01

    Loss-of-coolant accident (LOCA) analyses were completed using the Transient Reactor Analysis Code (TRAC) to support the U.S. Department of Energy efforts to restart the production reactors located at the Savannah River Site. The break location and pump operation after the LOCA were the parameters varied for these sensitivity studies. Three location of double-ended guillotine break were studied: plenum inlet, pump suction, and pump discharge. Three pump operation scenarios were also studied: continued operation of both ac and dc pumps, tripping of the ac motor at 2 s after the LOCA, and tripping of the ac motor at 200 s after the LOCA. The production reactors use low pressure and temperature heavy water as the process fluid. The reactor has a moderator tank that contains the fuel channels. Above the moderator tank is an upper plenum that distributes the heavy water to each fuel assembly. The heavy water flows down through the fuel channels and into the moderator tank. From the tank, the water is pumped back to the upper plenum through six loops. Each loop contains a pump and two heat exchangers. Four of the loops have an emergency core coolant system (ECCS) connection. This TRAC model has been benchmarked extensively against data taken in the actual reactors or in prototypical models of the components of the reactors. The calculations were completed using a version of TRAC-PF1/MOD 2 that was updated to include heavy water properties and other changes that are specific to the production reactors.

  16. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    SciTech Connect

    Heames, T.J. ); Williams, D.A.; Johns, N.A.; Chown, N.M. ); Bixler, N.E.; Grimley, A.J. ); Wheatley, C.J. )

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  17. Relative radiological impact from a reactor accident in the case of emerging nuclear fuels.

    PubMed

    Nicolaou, G

    2009-08-01

    An assessment has been carried out on the radiological impact on an area contaminated from an accident of a nuclear reactor loaded with different actinide fuels considered in transmutation and recycling schemes. The impact of these schemes is compared to reference cases of commercial UO2 and MOX fuels. The effective dose equivalent delivered to permanent residents has been calculated using the RESRAD code and used as an index for the assessment purposes. The highest and lowest doses would be delivered from the self-generating recycling of actinides in fast and thermal reactors, respectively. External irradiation is the main contributor to the dose delivered to the target population in comparison to ingestion and inhalation. The external dose delivered would be attributed for the first few years to 134Cs and for the following several tens of years to 137Cs. PMID:19590275

  18. Radiological protection issues arising during and after the Fukushima nuclear reactor accident.

    PubMed

    González, Abel J; Akashi, Makoto; Boice, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Sakai, Kazuo; Weiss, Wolfgang; Yamashita, Shunichi; Yonekura, Yoshiharu

    2013-09-01

    Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with 'contamination' of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of potential

  19. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-03-15

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was <0.3 kg/(m{sup 3}s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to {approx}3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged.

  20. Radiation protection: an analysis of thyroid blocking. [Effectiveness of KI in reducing radioactive uptake following potential reactor accident

    SciTech Connect

    Aldrich, D C; Blond, R M

    1980-01-01

    An analysis was performed to provide guidance to policymakers concerning the effectiveness of potassium iodide (KI) as a thyroid blocking agent in potential reactor accident situations, the distance to which (or area within which) it should be distributed, and its relative effectiveness compared to other available protective measures. The analysis was performed using the Reactor Safety Study (WASH-1400) consequence model. Four categories of accidents were addressed: gap activity release accident (GAP), GAP without containment isolation, core melt with a melt-through release, and core melt with an atmospheric release. Cost-benefit ratios (US $/thyroid nodule prevented) are given assuming that no other protective measures are taken. Uncertainties due to health effects parameters, accident probabilities, and costs are assessed. The effects of other potential protective measures, such as evacuation and sheltering, and the impact on children (critical population) are evaluated. Finally, risk-benefit considerations are briefly discussed.

  1. Radioactivity in persons exposed to fallout from the Chernobyl reactor accident

    SciTech Connect

    Schlenker, R.A.; Oltman, B.G.; Lucas, H.F.

    1987-01-01

    Measurements of fallout radioactivity were made in the thyroid region, abdomen, whole body, or urine of 96 persons who were in eastern Europe at the time of the Chernobyl reactor accident or who went there shortly afterward. The most frequently encountered radionuclides were /sup 131/I, sup 134,137/Cs, and /sup 103/Ru//sup 103/Rh. The median /sup 131/I activity in the thyroids of 42 subjects in whom radioiodine was detected and who were in Europe when the accident began was projected as 42 nCi the day the accident began. The median total body activity of /sup 134/Cs in 40 subjects in which it was detected was 1.7 nCi upon arrival in the US. For 51 subjects with detectable /sup 137/Cs burdens, the total body activity was 4.6 nCi. The risk of fatal thyroid cancer is less than 3 x 10/sup -6/ for nearly all subjects in this series. The risk of fatal cancer from /sup 134,137/Cs for subjects with cesium exposures similar to the ones observed by us, but who remained in Europe, is estimated as 1.4 x 10/sup -6/ to 4.2 x 10/sup -5/ with 95% of the risk attributable to /sup 137/Cs. 5 refs., 4 tabs.

  2. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  3. Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

    SciTech Connect

    White, M.D.; Lombardo, N.J.; Heard, F.J.; Ogden, D.M.; Quapp, W.J.

    1988-04-01

    Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and for uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.

  4. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

    SciTech Connect

    Himes, D.A.

    1997-11-17

    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.

  5. Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

    NASA Astrophysics Data System (ADS)

    Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

    2012-06-01

    The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High

  6. European Pressurized water Reactor (EPR) SAR ATWS Accident Analyses by using 3D Code Internal Coupling Method

    SciTech Connect

    Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.

    2004-07-01

    Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)

  7. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    SciTech Connect

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs.

  8. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-09-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios.

  9. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. )

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.

  10. Search for ^90Sr from the Fukushima Reactor Accident in San Francisco Bay Area Rainwater

    NASA Astrophysics Data System (ADS)

    Lo, B. T.; Chodash, P. A.; Thomas, K. J.; Norman, E. B.

    2012-10-01

    Shortly after the Fukushima reactor accident, we collected rainwater samples in the San Francisco Bay area. Subsequent gamma-ray counting revealed the presence of volatile short-lived fission fragments such as ^131, 132I, ^132Te, and ^134,137 Cs [1]. Recently, we have searched for the presence of the long-lived fission fragment ^90Sr in these same rainwater samples. To chemically separate Sr, a small amount of stable Sr carrier was dissolved in each rainwater sample. Potassium carbonate was then added to precipitate SrCO3. The precipitate was filtered, dried, and then beta counted using a planar Ge detector. Results from these measurements will be presented and compared to the levels of other fission fragments previously observed in the rainwater. [4pt] [1] E. B. Norman, C. T. Angell, P. A. Chodash, PLoSONE 6(9): e24330. Doi:10.1371/journal.pone.0024330.

  11. Scoping studies of vapor behavior during a severe accident in a metal-fueling reactor

    NASA Astrophysics Data System (ADS)

    Spencer, B. W.; Marchaterre, J. F.

    1985-04-01

    The consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel were examined. The principal gas and vapor species released are shown to be Xe, Cs, and bond sodium contained within the fuel porosity. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core. If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the ability of vapor expansion to perform appreciable work on the system and the ability of an expanding vapor bubble to transport fuel and fission produce species to the cover gas region where they may be released to the containment are largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

  12. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1994-01-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 210 citations and includes a subject term index and title list.)

  13. Chernobyl Nuclear Reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1996-11-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  14. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS Bibliographic database). Published Search

    SciTech Connect

    1993-09-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 208 citations and includes a subject term index and title list.)

  15. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1995-02-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 247 citations and includes a subject term index and title list.)

  16. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1996-01-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  17. Parametric study of recriticality in a boiling water reactor severe accident

    SciTech Connect

    Shamoun, B.I.; Witt, R.J. . Dept. of Nuclear Engineering and Engineering Physics)

    1994-08-01

    Recriticality is possible in a severe accident if unborated or low boron concentration water is added to a damaged core after control rod melting but before fuel melting. Recriticality in a severe accident in a boiling water reactor was parametrically investigated using the TWODANT code. Eigenvalue calculations for a unit central fuel cell with reflective boundary conditions were performed by solving the two-dimensional multigroup steady-state Boltzman transport equation using TWODANT. Two sets of calculations were performed in this work. The first set of calculations was carried out under three types of normal operating conditions to provide reference values for the accident calculations: (a) cold rodded condition, (b) cold unrodded condition, and (c) hot full-power condition. The eigenvalues at these conditions were found to be 1.055, 1.208, and 1.098, respectively. The second set of calculations was carried out after the melting of the control element and during the reflood phase, under the following reflood conditions: (a) reflood with unborated water and (b) reflood with borated water. For the reflood case with unborated water, five values of void fractions were considered (100, 60, 40, 20, and 0%). Decreasing void fractions represent greater refill levels during the reflood process. The system pressure was taken to be 7 MPa, while the moderator temperature was set to 560 K. Plotting the eigenvalue compared with the fraction of control materials lost indicates recriticality is only possible if nearly 100% of the control material is lost from the core. Eigenvalue calculations were repeated for short- and long-term recovery conditions of the reflood phase corresponding to maximum moderator density at 4 MPa pressure and 525 K moderator temperature and for 1 MPa pressure and 325 K moderator temperature, respectively. Recriticality was again observed to be a concern only after losing 95% ore more of control materials from the unit cell.

  18. A simplified model for calculating atmospheric radionuclide transport and early health effects from nuclear reactor accidents

    SciTech Connect

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1995-11-01

    During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences.

  19. Observations of fallout from the Fukushima reactor accident in Cienfuegos, Cuba.

    PubMed

    Alonso-Hernandez, Carlos M; Guillen-Arruebarrena, Aniel; Cartas-Aguila, Hector; Morera-Gomez, Yasser; Diaz-Asencio, Misael

    2012-05-01

    Following the recent accident at the Fukushima Daiichi nuclear power plant in Japan, radioactive contamination was observed near the reactor site. As a contribution towards the understanding of the worldwide impact of the accident, we collected fallout samples in Cienfuegos, Cuba, and examined them for the presence of above normal amounts of radioactivity. Gamma ray spectra measured from these samples showed clear evidence of fission products (131)I and (137)Cs. However, the fallout levels measured for these isotopes (135 ± 4.78 mBq m(-2) day(-1) for (131)I and 10.7 ± 0.38 mBq m(-2) day(-1)for (137)Cs) were very low and posed no health risk to the public. The doses received as consequence to the Fukushima fallout by the Cienfuegos population's (0.002 mSv per year) don't overcome the limit of dose (1 mSv per year) fixed for the public in Cuba. PMID:22310844

  20. A review of source term and dose estimation for the TMI-2 reactor accident

    SciTech Connect

    Gudiksen, P.H.; Dickerson, M.H.

    1990-09-01

    The TMI-2 nuclear reactor accident, which occurred on March 28, 1979 in Harrisburg, Pennsylvania, produced environmental releases of noble gases and small quantities of radioiodine. The releases occurred over a roughly two week period with almost 90% of the noble gases being released during the first three days after the initiation of the accident. Meteorological conditions during the prolonged release period varied from strong synoptic driven flows that rapidly transported the radioactive gases out of the Harrisburg area to calm situations that allowed the radioactivity to accumulate within the low lying river area and to subsequently slowly disperse within the immediate vicinity of the reactor. The results reported by various analysts, revealed that approximately 2.4--10 million curies of noble gases (mainly Xe-133), and about 14 curies of I-131 were released. During the first two days, when most of the noble gas release occurred, the plume was transported in a northerly direction causing the most exposed area to lie within a northwesterly to northeasterly direction from TMI. Changing surface winds caused the plume to be subsequently transported in a southerly direction, followed by an easterly direction. The calculated maximum whole body dose due to plume passage exceeded 100 mrem over an area extending several kilometers north of the plant, although the highest measured dose was 75 mrem. The collective dose equivalent (within a radius of 80 km) due to the noble gas exposure ranged over several orders of magnitude with a central estimate of 3300 person-rem. The small I-131 release produced barely detectable levels of activity in air and milk samples. This may have produced thyroid doses of a few milirem to a small segment of the population. 7 refs., 4 figs., 2 tabs.

  1. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  2. Material distribution in light water reactor-type bundles tested under severe accident conditions

    SciTech Connect

    Noack, V.; Hagen, S.J.L.; Hofmann, P.; Schanz, G.; Sepold, L.K.

    1997-02-01

    Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorber and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.

  3. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  4. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    SciTech Connect

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  5. LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor: Appendix A-4

    SciTech Connect

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.

  6. Sandia National Laboratories results for the 2010 criticality accident dosimetry exercise, at the CALIBAN reactor, CEA Valduc France.

    SciTech Connect

    Ward, Dann C.

    2011-09-01

    This document describes the personal nuclear accident dosimeter (PNAD) used by Sandia National Laboratories (SNL) and presents PNAD dosimetry results obtained during the Nuclear Accident Dosimeter Intercomparison Study held 20-23 September, 2010, at CEA Valduc, France. SNL PNADs were exposed in two separate irradiations from the CALIBAN reactor. Biases for reported neutron doses ranged from -15% to +0.4% with an average bias of -7.7%. PNADs were also exposed on the back side of phantoms to assess orientation effects.

  7. The ENEA criticality accident dosimetry system: a contribution to the 2002 international intercomparison at the SILENE reactor.

    PubMed

    Gualdrini, G; Bedogni, R; Fantuzzi, E; Mariotti, F

    2004-01-01

    The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (Silène reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescent detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user. PMID:15353692

  8. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. )

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  9. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    SciTech Connect

    Heams, T J; Williams, D A; Johns, N A; Mason, A; Bixler, N E; Grimley, A J; Wheatley, C J; Dickson, L W; Osborn-Lee, I; Domagala, P; Zawadzki, S; Rest, J; Alexander, C A; Lee, R Y

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  10. An analysis of thermionic space nuclear reactor power system: I. Effect of disassembling radial reflector, following a reactivity initiated accident

    SciTech Connect

    El-Genk, M.S.; Paramonov, D. )

    1993-01-10

    An analysis is performed to determine the effect of disassembling the radial reflector of the TOPAZ-II reactor, following a hypothetical severe Reactivity Initiated Accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the drums to rotate the full 180[degree] outward at their maximum speed of 1.4[degree]/s. Results indicate that disassembling only three of twelve radial reflector panels would successfully shutdown the reactor, with little overheating of the fuel and the moderator.

  11. Launch Vehicle Fire Accident Preliminary Analysis of a Liquid-Metal Cooled Thermionic Nuclear Reactor: TOPAZ-II

    NASA Astrophysics Data System (ADS)

    Hu, G.; Zhao, S.; Ruan, K.

    2012-01-01

    In this paper, launch vehicle propellant fire accident analysis of TOPAZ-II reactor has been done by a thermionic reactor core analytic code-TATRHG(A) developed by author. When a rocket explodes on a launch pad, its payload-TOPAZ-II can be subjected to a severe thermal environment from the resulting fireball. The extreme temperatures associated with propellant fires can create a destructive environment in or near the fireball. Different kind of propellants - liquid propellant and solid propellant which will lead to different fire temperature are considered. Preliminary analysis shows that the solid propellant fires can melt the whole toxic beryllium radial reflector.

  12. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    SciTech Connect

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y.; Rashid, Y.R.

    1998-08-01

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  13. Criticality accident dosimetry systems: an international intercomparison at the SILENE reactor in 2002.

    PubMed

    Médioni, R; Asselineau, B; Verrey, B; Trompier, F; Itié, C; Texier, C; Muller, H; Pelcot, G; Clairand, I; Jacquet, X; Pochat, J L

    2004-01-01

    In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue. PMID:15353686

  14. Evaluation of graphite/steam interactions for ITER (International Thermonuclear Experimental Reactor) accident scenarios

    SciTech Connect

    Smolik, G.R.; Merrill, B.J.; Piet, S.J.; Holland, D.F.

    1990-01-01

    This paper presents the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during an accident involving coolant leakage into the plasma chamber of the International Thermonuclear Experimental Reactor (ITER). This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 and 1700{degree}C. The effects of steam flow rate and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We used the relationships for GraphNOL N3M in thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 16 refs., 7 figs., 1 tab.

  15. Atmospheric radionuclides from the Fukushima Dai-ichi nuclear reactor accident observed in Vietnam.

    PubMed

    Long, N Q; Truong, Y; Hien, P D; Binh, N T; Sieu, L N; Giap, T V; Phan, N T

    2012-09-01

    Radionuclides from the reactor accident at the Fukushima Dai-ichi Nuclear Power Plant were observed in the surface air at stations in Hanoi, Dalat, and Ho Chi Minh City (HCMC) in Vietnam, about 4500 km southwest of Japan, during the period from March 27 to April 22, 2011. The maximum activity concentrations in the air measured at those three sites were 193, 33, and 37 μBq m(-3) for (131)I, (13)(4)Cs, and (13)(7)Cs, respectively. Peaks of radionuclide concentrations in the air corresponded to arrival of the air mass from Fukushima to Vietnam after traveling for 8 d over the Pacific Ocean. Cesium-134 was detected with the (134)Cs/(137)Cs activity ratio of about 0.85 in line with observations made elsewhere. The (131)I/(137)Cs activity ratio was observed to decrease exponentially with time as expected from radioactive decay. The ratio at Dalat, where is 1500 m high, was higher than those at Hanoi and HCMC in low lands, indicating the relative enrichment of the iodine in comparison to cesium at high altitudes. The time-integrated surface air concentrations of the Fukushima-derived radionuclides in the Southeast Asia showed exponential decrease with distance from Fukushima. PMID:22200554

  16. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  17. Assessment of possible consequences of a hypothetical reactivity accident associated with a {open_quotes}Topaz-2{close_quotes} spacecraft reactor entering water

    SciTech Connect

    Glushkov, E.S.; Ermoshin, M.Yu.; Ponomarev-Stepnoi; Skorlygin, V.V.

    1994-12-01

    An accident analysis for a Russian Topaz-2 nuclear reactor is summarized. The accident scenario involves emergency return from orbit, severe damage to reactor structural elements, and subsequent falling of the reactor core into the ocean. The thermionic converter reactor, used in spacecraft, has a large neutron leakage which decreases when water enters the inner core cavity. Preliminary results of numerical modeling, summarized in the article, show that the possible consequences of the hypothetical accidental submersion are limited. 8 refs., 2 figs., 2 tabs.

  18. Nuclear-reactor accidents: Chernobyl, TMI, and Windscale. January 1974-September 1988 (Citations from Pollution Abstracts). Report for January 1974-September 1988

    SciTech Connect

    Not Available

    1988-11-01

    This bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear-reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout, the resultant runoff into rivers, lakes, and the sea, the radiation effects on people, and the transfrontier radioactive contamination of the environment. (Contains 105 citations fully indexed and including a title list.)

  19. Significant increase in trisomy 21 in Berlin nine months after the Chernobyl reactor accident: temporal correlation or causal relation?

    PubMed Central

    Sperling, K.; Pelz, J.; Wegner, R. D.; Dörries, A.; Grüters, A.; Mikkelsen, M.

    1994-01-01

    OBJECTIVE--To assess whether the increased prevalence of trisomy 21 in West Berlin in January 1987 might have been causally related to exposure to ionising radiation as a result of the Chernobyl reactor accident or was merely a chance event. DESIGN--Analysis of monthly prevalence of trisomy 21 in West Berlin from January 1980 to December 1989. SETTING--Confines of West Berlin. RESULTS--Owing to the former "island" situation of West Berlin and its well organised health services, ascertainment of trisomy 21 was thought to be almost complete. A cluster of 12 cases occurred in January 1987 as compared with two or three expected. After exclusion of factors that might have explained the increase, including maternal age distribution, only exposure to radiation as a result of the Chernobyl reactor accident remained. In six of seven cases that could be studied cytogenetically the extra chromosome was of maternal origin, confirming that nondisjunction had occurred at about the time of conception. CONCLUSION--On the basis of two assumptions--(a) that maternal meiosis is an error prone process susceptible to exogenous factors at the time of conception; (b) that owing to the high prevalence of iodine deficiency in Berlin a large amount of iodine-131 would have been accumulated over a short period--it is concluded that the increased prevalence of trisomy 21 in West Berlin in January 1987 was causally related to a short period of exposure to ionising radiation as a result of the Chernobyl reactor accident. PMID:8044094

  20. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    SciTech Connect

    Gauntt, Randall O.; Mattie, Patrick D.

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  1. Activity ratios in soil contaminated by the source of different reactor condition in the FDNPP accident

    NASA Astrophysics Data System (ADS)

    Satou, Yukihiko; Sueki, Keisuke; Sasa, Kimikazu; Matsunaka, Tetsuya; Shibayama, Nao; Takahashi, Tsutomu; Kinoshita, Norikazu

    2014-05-01

    The Fukushima Dai-ichi Nuclear power plant (FDNPP) accident caused radioactive contamination on the surface soil at Fukushima and its adjacent prefectures. Substantial contamination has been found in the northwestern area from the FDNPP, according to the airborne monitoring and ground base survey by the Japanese government. Activity ratios would have characteristic information on emission sources because each relevant reactor had different amount of radionuclide and different activity ratio. The ratios can be used to clarify more detailed source and process in the contamination. We have addressed to consider them in Namie town, northwestern region from the FDNPP. This study focused on the gamma-ray emitting radionuclides of 134Cs, 137Cs, and 110mAg. The activities were decay-corrected as of 11th March, 2011 when all nuclear reactors scrammed. Data of activity ratios by our results and the Japanese official report classified the investigated northwestern region into 3 groups. Ratios of 0.02 for 110mAg/137Cs and 0.90 for 134Cs/137Cs were observed in the northern region of 15 km inside from the FDNPP. On the other hand, two kinds of 110mAg/137Cs ratios of 0.005 and 0.002 were distributed broadly in the region 60 km away from the plant. The 134Cs/137Cs ratio was 0.98 there. The activity ratios of 110mAg/137Cs and 134Cs/137Cs in the northern region from the FDNPP correspond to those of nuclear fuel in Unit 1 according to estimation using the ORIGEN code. The 134Cs/137Cs in the northwestern area from FDNPP agrees with that of Unit 2 and 3. The 110mAg/137Cs ratios of 0.005 and0.002 are 1/5 - 1/10 of the Unit 2 and 3. Official report has announced that discharges of the radionuclides from Unit 2 and 3 occurred on 14th March, 2011. It is known that contamination in the northwestern region from the FDNPP took place on 15th March, 2011. Plausible species for silver in reactor core, metal, and halide etc. have higher boiling point than those species for cesium. The core would

  2. MACC1 upregulation promotes gastric cancer tumor cell metastasis and predicts a poor prognosis*

    PubMed Central

    Xie, Qiu-ping; Xiang, Cheng; Wang, Gang; Lei, Ke-feng; Wang, Yong

    2016-01-01

    In various studies, metastasis associated with colon cancer 1 (MACC1) has been frequently reported to be abnormally highly expressed in human lung cancer, colon cancer, and hepatocellular carcinoma. Our study focuses on the association of MACC1 expression with gastric cancer (GC). During our experiment, the MACC1 expression was tested in 105 GC samples using an immunohistochemical (IHC) method. The clinical characteristics and prognosis of these patients were summarized. During analysis, MACC1 distribution in GC samples with distant metastasis was higher than that in normal samples and in tumors with no dissemination. Subsequently, a lower 5-year survival rate had a strong correlation with high MACC1 expression. As a consequence, the present results suggest that MACC1 is more frequently expressed in a poor prognosis phenotype of GC and acts as a promising prognostic prediction parameter for GC. PMID:27143263

  3. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    SciTech Connect

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.

  4. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    NASA Astrophysics Data System (ADS)

    Osadchaya, D. Yu.; Fuks, R. L.

    2014-04-01

    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  5. Effects of control system failures on transients and accidents at a 3-loop Westinghouse pressurized water reactor. Volume 2. Appendices

    SciTech Connect

    Bruske, S.J.; Davis, C.B.; Ogden, D.M.; Ransom, C.B.; Stitt, B.D.; Stromberg, H.M.; Waterman, M.E.

    1985-10-01

    Safety Implications of Control Systems (A-47) was approved as an Unresolved Safety Issue (USI) by the Nuclear Regulatory Commission (NRC) in December of 1980. USI A-47 is concerned with the potential for transients or accidents being made more severe than previously analyzed as a result of control system failures. This report describes the work performed on the effects of control system failures on transients and accidents at a Westinghouse 3-loop pressurized water reactor. In this volume, the appendices contain detailed information consisting of the FMEA (failure mode and analysis) results, an in-depth description of the computer model, the deterministic computer analyses, and responses to comments made by Carolina Power and Light Company and Westinghouse Electric Corporation.

  6. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    SciTech Connect

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  7. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N.; Kannan, S.E.

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  8. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  9. MLAM assessment of air concentration, deposition, and dose for Chernobyl reactor accident

    SciTech Connect

    Olsen, A.R.; Davis, W.E.; Didier, B.T.; Soldat, J.K.; Napier, B.A.; Peloquin, R.A.

    1989-12-01

    The purpose of this report is to provide estimates for the areas in Europe affected by the accident involving Unit 4 of the Chernobylskaya Atomic Energy Station which resulted in the release of radioactive material to the atmosphere.

  10. Hypothetical accident scenario analyses for a 250-MW(T) modular high temperature gas-cooled reactor

    SciTech Connect

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1985-01-01

    This paper describes calculations performed at Oak Ridge National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission's HTGR Safety Research Program, to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e. upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced helium primary coolant circulation, loss of primary coolant pressurization, and loss of heat sink were studied and are discussed. Comparisons of calculated and measured response for a flow reduction test on the German reactor AVR are also presented.

  11. EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

    1981-04-01

    Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

  12. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    SciTech Connect

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization.

  13. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

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

    SciTech Connect

    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

  15. GOSAT BESD XCO2 for MACC-II: Current Status

    NASA Astrophysics Data System (ADS)

    Heymann, Jens; Reuter, Maximilian; Hilker, Michael; Buchwitz, Michael; Schneising, Oliver; Bovensmann, Heinrich; Burrows, John P.

    2014-05-01

    Carbon dioxide (CO2) is the most important anthropogenic greenhouse gas contributing to global warming. Near-surface sensitive measurements from satellite instruments such as SCIAMACHY on-board ENVISAT and TANSO on-board GOSAT can provide important missing global information on the regional sources and sinks of CO2. However, this requires to meet challenging accuracy requirements. An algorithm to retrieve the column-averaged dry air mole fraction of CO2 ("XCO2") from satellite measurements is the Bremen Optimal Estimation DOAS (BESD) retrieval algorithm. BESD was originally developed to retrieve XCO2 from SCIAMACHY measurements. In the framework of the MACC-II project, the SCIAMACHY BESD XCO2 product was delivered for MACC-II for delayed mode production and monitoring by University of Bremen. After the loss of ENVISAT in April 2012, it was decided that University of Bremen shall deliver GOSAT XCO2 instead of SCIAMACHY XCO2. To achieve this, the BESD algorithm has been modified. Consistency of long-term XCO2 products derived from different satellites is important for climate applications and using the same algorithm contributes to minimize inconsistencies. Here, we present results from these activities.

  16. BENCHMARKING UPGRADED HOTSPOT DOSE CALCULATIONS AGAINST MACCS2 RESULTS

    SciTech Connect

    Brotherton, Kevin

    2009-04-30

    The radiological consequence of interest for a documented safety analysis (DSA) is the centerline Total Effective Dose Equivalent (TEDE) incurred by the Maximally Exposed Offsite Individual (MOI) evaluated at the 95th percentile consequence level. An upgraded version of HotSpot (Version 2.07) has been developed with the capabilities to read site meteorological data and perform the necessary statistical calculations to determine the 95th percentile consequence result. These capabilities should allow HotSpot to join MACCS2 (Version 1.13.1) and GENII (Version 1.485) as radiological consequence toolbox codes in the Department of Energy (DOE) Safety Software Central Registry. Using the same meteorological data file, scenarios involving a one curie release of {sup 239}Pu were modeled in both HotSpot and MACCS2. Several sets of release conditions were modeled, and the results compared. In each case, input parameter specifications for each code were chosen to match one another as much as the codes would allow. The results from the two codes are in excellent agreement. Slight differences observed in results are explained by algorithm differences.

  17. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  18. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    SciTech Connect

    Weiss, A. J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  19. Study of light water reactor containments under important severe accident conditions

    SciTech Connect

    Hofmayer, C.H.; Pratt, W.T.; Bagchi, G.; Noonan, V.S.

    1985-01-01

    The US Nuclear Regulatory Commission has sponsored studies to develop a ''LEAKAGE-BEFORE-FAILURE'' model for use in severe accident risk assessments to provide a means of accounting for significant containment leakage prior to reaching the containment threshold pressure. Six containment types have been studied (large dry, subatmospheric, ice condenser, Mark I, II, and III). Potential leak paths through major containment penetration assemblies were investigated and upper-bound estimates of leak areas established. These leak areas may result from increasing internal pressure and degradation of nonmetallic seal materials due to severe accident conditions. This paper describes the approach and summarizes the results and conclusions of this study.

  20. Liquid metal reactions under postulated accident conditions for fission and fusion reactors

    SciTech Connect

    Muhlestein, L.D.

    1980-04-01

    Sodium and lithium reactions are considered in the context of a postulated breach of a coolant boundary. Specific topics addressed are coolant-atmosphere and coolant-material reactions which may contribute to the overall consequence of a postulated accident scenario, and coolant reaction extinguishment and effluent control which may be desirable for containment of the spilled coolant.

  1. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect

    Simpson, D.B.; Greene, S.R.

    1990-01-01

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

  2. Assessment of severe accident prevention and mitigation features: BWR (boiling water reactor), Mark I containment design

    SciTech Connect

    Pratt, W.T.; Eltawila, F.; Perkins, K.R.; Fitzpatrick, R.G.; Luckas, W.J.; Lehner, J.R.; Davis, P.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark I containments (BWR Mark I's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Peach Bottom plant and from assessment of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark I to severe accident containment loads were also identified. In addition, those features of a BWR Mark I, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Peach Bottom and other Mark I plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance.

  3. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: UPDATE 2, VOLUME III

    EPA Science Inventory

    The original report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem St...

  4. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: UPDATE 2, VOLUME II

    EPA Science Inventory

    The original report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem St...

  5. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: UPDATE 2, VOLUME I

    EPA Science Inventory

    The original report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem St...

  6. Risk Analysis for Public Consumption: Media Coverage of the Ginna Nuclear Reactor Accident.

    ERIC Educational Resources Information Center

    Dunwoody, Sharon; And Others

    Researchers have determined that the lay public makes risk judgments in ways that are very different from those advocated by scientists. Noting that these differences have caused considerable concern among those who promote and regulate health and safety, a study examined media coverage of the accident at the Robert E. Ginna nuclear power plant…

  7. Single channel flow blockage accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    SciTech Connect

    Popov, N.K.; Abdul-Razzak, A.; Snell, V.G.; Langman, V.; Sills, H.

    2004-07-01

    The Advanced Candu Reactor (ACRTM) is an evolutionary advancement of the current Candu 6{sup R} reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by a heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper documents the results of Phenomena Identification and Ranking Table (PIRT) results for a very limited frequency, beyond design basis event of the ACR design. This PIRT is developed in a highly structured process of expert elicitation that is well supported by experimental data and analytical results. The single-channel flow blockage event in an ACR reactor assumes a severe flow blockage of one of the reactor fuel channels, which leads to a reduction of the flow in the affected channel, leading to fuel cladding and fuel temperature increase. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the flow blockage phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the finalized PIRT tables. (authors)

  8. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    SciTech Connect

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  9. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    SciTech Connect

    Powers, Jeffrey J.; George, Nathan; Maldonado, G. Ivan; Worrall, Andrew

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

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

    SciTech Connect

    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.

  11. The foaming of U-Al fuel under simulated reactor accident conditions

    SciTech Connect

    Neimark, L.A.; Liu, Y.Y.

    1993-03-01

    Postirradiation heating tests were conducted on segments of UAl{sub 4}/Al dispersion fuel plates clad with Al to scope the foaming (rapid swelling) behavior of such fuels during beyond-design-basis accident scenarios. Four tests investigated maximum temperature, ramp rate, and duration with a liquid phase as parameters in foam formation and stability. Real-time fission-gas release was also determined during the foaming process. Ramp-rate had the most noticeable effect of foam formation and collapse.

  12. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

  13. MACC-1 Promotes Endothelium-Dependent Angiogenesis in Gastric Cancer by Activating TWIST1/VEGF-A Signal Pathway

    PubMed Central

    Zhao, Yang; Dong, Shaoting; Zhang, Jingwen; Luo, Yuhao; Huang, Na; Shi, Min; Bin, Jianping; Liao, Yulin; Liao, Wangjun

    2016-01-01

    Endothelium-dependent angiogenesis is thought to be a crucial step in cancer progression. We previously reported that metastasis-associated in colon cancer-1 (MACC1) contributed to the vasculogenic mimicry in gastric cancer (GC), but it remains unknown whether MACC1 promotes endothelium-dependent angiogenesis of GC and whether TWIST1 is involved in this process. In the present study, we detected MACC1 expression and microvessel density (MVD) by immunohistochemistry in 159 patients with stage I-III GC, and investigated the role of TWIST1 and vascular endothelial growth factor A (VEGF-A) in MACC1-induced endothelium-dependent angiogenesis using nude mice with GC xenografts, and human umbilical vein endothelial cells (HUVECs) that were co-cultured with conditioned media from overexpression and interference MACC1 GC cells. We found that MACC1 expression was positively correlated with an increased MVD and tumor recurrence in GC patients. In GC xenograft models, MACC1 elevated MVD and upregulated the expression of VEGF-A as well as accelerated tumor growth. In addition, MACC1 obviously increased the expression of TWIST1 and induced tube-like formation of HUVECs, whereas attenuation of TWIST1 suppressed the protein expression of VEGF-A and repealed the effect of MACC1 on tube formation. Our findings shed light on the function of MACC1 in endothelium-dependent angiogenesis of GC and suggest potential prognostic and therapeutic value. PMID:27280289

  14. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    SciTech Connect

    Yu, D.; Kastenberg, W.E.; Okrent, D. . Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-06-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities.

  15. Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors

    SciTech Connect

    Spencer, B.W.; Marchaterre, J.F.; Anderson, R.P.; Armstrong, D.R.; Baker, L.; Cho, D.H.; Gabor, J.D.; Pedersen, D.R.; Sienicki, J.J.; Stein, R.P.

    1986-01-01

    Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO/sub 2//sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included.

  16. Assessment of light water reactor fuel damage during a reactivity initiated accident

    SciTech Connect

    MacDonald, P.E.; Seiffert, S.L.; Martinson, Z.R.; McCardell, R.K.; Owen, D.E.; Fukuda, S.K.

    1980-01-01

    This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO/sub 2/. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies.

  17. MACC1 mediates acetylcholine-induced invasion and migration by human gastric cancer cells

    PubMed Central

    Xia, Jianling; Zhou, Rui; Wu, Zhenzhen; Zhao, Yang; Shi, Min

    2016-01-01

    The neurotransmitter acetylcholine (ACh) promotes the growth and metastasis of several cancers via its M3 muscarinic receptor (M3R). Metastasis-associated in colon cancer-1 (MACC1) is an oncogene that is overexpressed in gastric cancer (GC) and plays an important role in GC progression, though it is unclear how MACC1 activity is regulated in GC. In this study, we demonstrated that ACh acts via M3Rs to promote GC cell invasion and migration as well as expression of several markers of epithelial-mesenchymal transition (EMT). The M3R antagonist darifenacin inhibited GC cell activity in both the presence and absence of exogenous ACh, suggesting GC cells secrete endogenous ACh, which then acts in an autocrine fashion to promote GC cell migration/invasion. ACh up-regulated MACC1 in GC cells, and MACC1 knockdown using siRNA attenuated the effects of ACh on GC cells. AMP-activated protein kinase (AMPK) served as an intermediate signal between ACh and MACC1. These findings suggest that ACh acts via a M3R/AMPK/MACC1 signaling pathway to promote GC cell invasion/migration, which provides insight into the mechanisms underlying GC growth and metastasis and may shed light on new targets for GC treatment. PMID:26919111

  18. Containment performance of prototypical reactor containments subjected to severe accident conditions

    SciTech Connect

    Klamerus, E.W.; Bohn, M.P.; Wesley, D.A.; Krishnaswamy, C.N.

    1996-12-01

    In SECY-90-016, the NTRC proposed a safety goal of a conditional containment failure probability (CCFP) of 0.1 and the alternative acceptance criteria allowed for steel containments, which specifies that the stresses should not exceed ASNE Level C allowables for severe accident pressures and temperatures. In this work, the need for an equivalent criterion for concrete containments was studied. Six surrogate containments were designed and analyzed in order to compare the margins between design pressure, pressure resulting in exceedance of Level C (or yield) stress limits, and ultimate pressure. For comparability, each containment has an identical internal volume and design pressure. Results from the analysis showed margins to yield are comparable and display a similar margin for both steel and concrete containments. In addition, the margin to failure, although slightly higher in the steel containments, were also comparable. Finally, a CCFP for code design was determined based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies. The resulting CCFP`s were less then 0.02 (or 2%) for all the surrogate containments studied, showing that these containment designs all achieved the NRC safety goal.

  19. Interactive Rapid Dose Assessment Model (IRDAM): reactor-accident assessment methods. Vol. 2

    SciTech Connect

    Poeton, R.W.; Moeller, M.P.; Laughlin, G.J.; Desrosiers, A.E.

    1983-05-01

    As part of the continuing emphasis on emergency preparedness, the US Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). This system, the Interactive Rapid Dose Assessment Model (IRDAM) is a micro-computer based program for rapidly assessing the radiological impact of accidents at nuclear power plants. This document describes the technical bases for IRDAM including methods, models and assumptions used in calculations. IRDAM calculates whole body (5-cm depth) and infant thyroid doses at six fixed downwind distances between 500 and 20,000 meters. Radionuclides considered primarily consist of noble gases and radioiodines. In order to provide a rapid assessment capability consistent with the capacity of the Osborne-1 computer, certain simplifying approximations and assumptions are made. These are described, along with default values (assumptions used in the absence of specific input) in the text of this document. Two companion volumes to this one provide additional information on IRDAM. The user's Guide (NUREG/CR-3012, Volume 1) describes the setup and operation of equipment necessary to run IRDAM. Scenarios for Comparing Dose Assessment Models (NUREG/CR-3012, Volume 3) provides the results of calculations made by IRDAM and other models for specific accident scenarios.

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

    SciTech Connect

    Not Available

    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.

  1. A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

    NASA Astrophysics Data System (ADS)

    Bertch, Timothy Creston

    1998-12-01

    Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release

  2. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  3. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    SciTech Connect

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  4. Analysis of station blackout accidents for the Bellefonte pressurized water reactor

    SciTech Connect

    Gasser, R D; Bieniarz, P P; Tills, J L

    1986-09-01

    An analysis has been performed for the Bellefonte PWR Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis which include the effects of direct heating on containment loading, and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating which involves more than about 50% of the core can fail the Bellefonte containment, but natural convection in the RCS may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach due to natural circulation and after vessel breach due to reevolution of retained fission products by fission product heating of RCS structures.

  5. The feasibility of using 129I to reconstruct 131I deposition from the Chernobyl reactor accident.

    PubMed

    Straume, T; Marchetti, A A; Anspaugh, L R; Khrouch, V T; Gavrilin YuI; Shinkarev, S M; Drozdovitch, V V; Ulanovsky, A V; Korneev, S V; Brekeshev, M K; Leonov, E S; Voigt, G; Panchenko, S V; Minenko, V F

    1996-11-01

    Radioiodine released to the atmosphere from the accident at the Chernobyl nuclear power station in the spring of 1986 resulted in large-scale thyroid-gland exposure of populations in Ukraine, Belarus, and Russia. Because of the short half life of 131I (8.04 d), adequate data on the intensities and patterns of iodine deposition were not collected, especially in the regions where the incidence of childhood-thyroid cancer is now increasing. Results are presented from a feasibility study that show that accelerator-mass-spectrometry measurements of 129I (half life 16 x 106 y) in soil can be used to reconstruct 131I-deposition density and thus help in the thyroid-dosimetry effort that is now urgently needed to support epidemiologic studies of childhood-thyroid cancer in the affected regions. PMID:8887520

  6. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    SciTech Connect

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-11

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ∼5% of maximum value by the final moment of the stage of linear increase in the temperature.

  7. A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors

    SciTech Connect

    S. Khericha

    2011-06-01

    This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of

  8. Nuclear reactor accidents--the use of KI as a blocking agent against radioiodine uptake in the thyroid--a review.

    PubMed

    Crocker, D G

    1984-06-01

    This paper is intended for those people who are responsible for the public health and safety in the event of a nuclear reactor accident. The possibilities and consequences of a radioiodine release are examined briefly. The possible side effects of stable iodine, the prognosis for radiation-induced thyroid disease, and the alternative protective measures are put into perspective and assessed for their individual risks and benefits. It is recommended that all appropriate counter-radiation measures be considered in the case of a reactor accident, and that the harmful side effects of the various actions be weighed carefully. Definitive guidelines as to when the use of KI is justified must be decided upon before implementing mass distribution. PMID:6373674

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

    SciTech Connect

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

    1994-09-01

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

  10. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

    SciTech Connect

    Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

    1981-09-01

    A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

  11. Validation of reactive gases and aerosols in the MACC global analysis and forecast system

    NASA Astrophysics Data System (ADS)

    Eskes, H.; Huijnen, V.; Arola, A.; Benedictow, A.; Blechschmidt, A.-M.; Botek, E.; Boucher, O.; Bouarar, I.; Chabrillat, S.; Cuevas, E.; Engelen, R.; Flentje, H.; Gaudel, A.; Griesfeller, J.; Jones, L.; Kapsomenakis, J.; Katragkou, E.; Kinne, S.; Langerock, B.; Razinger, M.; Richter, A.; Schultz, M.; Schulz, M.; Sudarchikova, N.; Thouret, V.; Vrekoussis, M.; Wagner, A.; Zerefos, C.

    2015-11-01

    The European MACC (Monitoring Atmospheric Composition and Climate) project is preparing the operational Copernicus Atmosphere Monitoring Service (CAMS), one of the services of the European Copernicus Programme on Earth observation and environmental services. MACC uses data assimilation to combine in situ and remote sensing observations with global and regional models of atmospheric reactive gases, aerosols, and greenhouse gases, and is based on the Integrated Forecasting System of the European Centre for Medium-Range Weather Forecasts (ECMWF). The global component of the MACC service has a dedicated validation activity to document the quality of the atmospheric composition products. In this paper we discuss the approach to validation that has been developed over the past 3 years. Topics discussed are the validation requirements, the operational aspects, the measurement data sets used, the structure of the validation reports, the models and assimilation systems validated, the procedure to introduce new upgrades, and the scoring methods. One specific target of the MACC system concerns forecasting special events with high-pollution concentrations. Such events receive extra attention in the validation process. Finally, a summary is provided of the results from the validation of the latest set of daily global analysis and forecast products from the MACC system reported in November 2014.

  12. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    SciTech Connect

    Glumac, B; Ravnik, M.; Logar, M.

    1997-02-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10{sup {minus}6} per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool.

  13. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  14. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME V. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  15. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME I. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  16. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: UPDATE. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  17. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME IV. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  18. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME III. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  19. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME VI. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  20. THREE MILE ISLAND NUCLEAR REACTOR ACCIDENT OF MARCH 1979. ENVIRONMENTAL RADIATION DATA: VOLUME II. A REPORT TO THE PRESIDENT'S COMMISSION ON THE ACCIDENT AT THREE MILE ISLAND

    EPA Science Inventory

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corp...

  1. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    DOE PAGESBeta

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the

  2. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    SciTech Connect

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to

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

    SciTech Connect

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    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.

  4. The use of Sentinel satellite data in the MACC-II Copernicus pre-operational atmosphere service

    NASA Astrophysics Data System (ADS)

    Engelen, Richard; Flemming, Johannes; Benedetti, Angela; Inness, Antje; Massart, Sebastien; Parrington, Mark; Peuch, Vincent-Henri

    2014-05-01

    The Monitoring Atmospheric Composition and Climate (MACC-II) project is the current pre-operational atmosphere service of the European Copernicus programme. MACC-II provides data records on atmospheric composition for recent years, data for monitoring present conditions and forecasts of the distribution of key constituents for a few days ahead. MACC-II combines state-of-the-art atmospheric modelling with Earth observation data to provide information services covering European Air Quality, Global Atmospheric Composition, Climate, and UV and Solar Energy. MACC-II uses a wide array of satellite and in-situ data observing both meteorological and atmospheric composition variables to provide a best estimate of the current state of the atmosphere on a daily basis. These analyses are then used as initial conditions for 5-day global forecasts of atmospheric composition and 4-day European air quality forecasts. This presentation will provide an overview of the MACC-II pre-operational monitoring/forecasting system focusing on the use of current and future satellite data. The Sentinel missions will provide crucial new information on atmospheric composition, both for operational and research purposes, and we will show how MACC-II is preparing for these new observations and how MACC-II can provide important feedback about the data quality through careful data monitoring in the comprehensive data assimilation system. The latter information will be very beneficial for the scientific community to make more optimal use of the Sentinel data.

  5. Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

    NASA Astrophysics Data System (ADS)

    Ha, Taesung

    A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

  6. Model sensitivity to MACC anthropogenic and biogenic emissions: Global simulations and evaluation for reactive gases

    NASA Astrophysics Data System (ADS)

    Stein, O.; Schultz, M. G.; Bouarar, I.; Clark, H.; Katragkou, E.; Leitao, J.; Heil, A.

    2012-04-01

    The EU projects MACC (Monitoring Atmospheric Composition and Climate, 2009-2011) and MACC-II (2011-2014) prepare for the operational Global Monitoring for Environment and Security (GMES) atmospheric core service which is envisaged to start in 2014. Besides global service lines for greenhouse gases and aerosols, emphasis is put also on global monitoring and forecasting of reactive gases. The MACC reanalysis and forecast simulations benefit from the multi-sensor approach for data assimilation of ozone, CO and NO2 observations. Currently the Integrated Forecast System (IFS) of the European Centre for Medium-range Weather Forecasts (ECMWF) is coupled to the chemical transport model MOZART-3 to represent in detail the chemical conversion as well as major source and sink processes. A global emission inventory for reactive gases has been developed as part of the MACC project. Based upon the ACCMIP emissions for the year 2000 these emissions are extrapolated for years after 2000 with the Representative Concentration Pathway RCP8.5 scenario and extended for VOCs and several other species. This inventory composes the MACCity anthropogenic emission inventory (Granier et al. 2011). During the MACC project it became apparent that using the MACCity emissions in reanalysis simulations for recent years led to an underestimation of CO concentrations in the Northern Hemisphere when compared to independent observations. In order to give insight into the reasons for this behavior we conducted MOZART offline simulations for the year 2008 to test the sensitivity of the chemical transport model to the varying emissions. Therefore we ran MOZART with different sets of emissions: 1. MACCity emissions, 2. The GEMS/RETRO emission inventory, 3. MACCity emissions, but with increased traffic CO emissions. While using the emission inventory developed in the RETRO and GEMS projects gives quite reasonable tropospheric concentrations for the key species, the MACCity emissions are too low

  7. RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

    SciTech Connect

    Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

    1990-05-01

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

  8. Molecular Characterization of MaCCS, a Novel Copper Chaperone Gene Involved in Abiotic and Hormonal Stress Responses in Musa acuminata cv. Tianbaojiao

    PubMed Central

    Feng, Xin; Chen, Fanglan; Liu, Weihua; Thu, Min Kyaw; Zhang, Zihao; Chen, Yukun; Cheng, Chunzhen; Lin, Yuling; Wang, Tianchi; Lai, Zhongxiong

    2016-01-01

    Copper/zinc superoxide dismutases (Cu/ZnSODs) play important roles in improving banana resistance to adverse conditions, but their activities depend on the copper chaperone for superoxide dismutase (CCS) delivering copper to them. However, little is known about CCS in monocots and under stress conditions. Here, a novel CCS gene (MaCCS) was obtained from a banana using reverse transcription PCR and rapid-amplification of cDNA ends (RACE) PCR. Sequence analyses showed that MaCCS has typical CCS domains and a conserved gene structure like other plant CCSs. Alternative transcription start sites (ATSSs) and alternative polyadenylation contribute to the mRNA diversity of MaCCS. ATSSs in MaCCS resulted in one open reading frame containing two in-frame start codons to form two protein versions, which is supported by the MaCCS subcellular localization of in both cytosol and chloroplasts. Furthermore, MaCCS promoter was found to contain many cis-elements associated with abiotic and hormonal responses. Quantitative real-time PCR analysis showed that MaCCS was expressed in all tested tissues (leaves, pseudostems and roots). In addition, MaCCS expression was significantly induced by light, heat, drought, abscisic acid and indole-3-acetic acid, but inhibited by relatively high concentrations of CuSO4 and under cold treatment, which suggests that MaCCS is involved in abiotic and hormonal responses. PMID:27023517

  9. Molecular Characterization of MaCCS, a Novel Copper Chaperone Gene Involved in Abiotic and Hormonal Stress Responses in Musa acuminata cv. Tianbaojiao.

    PubMed

    Feng, Xin; Chen, Fanglan; Liu, Weihua; Thu, Min Kyaw; Zhang, Zihao; Chen, Yukun; Cheng, Chunzhen; Lin, Yuling; Wang, Tianchi; Lai, Zhongxiong

    2016-01-01

    Copper/zinc superoxide dismutases (Cu/ZnSODs) play important roles in improving banana resistance to adverse conditions, but their activities depend on the copper chaperone for superoxide dismutase (CCS) delivering copper to them. However, little is known about CCS in monocots and under stress conditions. Here, a novel CCS gene (MaCCS) was obtained from a banana using reverse transcription PCR and rapid-amplification of cDNA ends (RACE) PCR. Sequence analyses showed that MaCCS has typical CCS domains and a conserved gene structure like other plant CCSs. Alternative transcription start sites (ATSSs) and alternative polyadenylation contribute to the mRNA diversity of MaCCS. ATSSs in MaCCS resulted in one open reading frame containing two in-frame start codons to form two protein versions, which is supported by the MaCCS subcellular localization of in both cytosol and chloroplasts. Furthermore, MaCCS promoter was found to contain many cis-elements associated with abiotic and hormonal responses. Quantitative real-time PCR analysis showed that MaCCS was expressed in all tested tissues (leaves, pseudostems and roots). In addition, MaCCS expression was significantly induced by light, heat, drought, abscisic acid and indole-3-acetic acid, but inhibited by relatively high concentrations of CuSO₄ and under cold treatment, which suggests that MaCCS is involved in abiotic and hormonal responses. PMID:27023517

  10. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 1: Main report

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harrison, J.D.; Harper, F.T.; Hora, S.C.

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models.

  11. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harper, F.T.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models.

  12. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Haskin, F.E.; Harper, F.T.; Goossens, L.H.J.; Kraan, B.C.P.; Grupa, J.B.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models.

  13. DIANA: A multi-phase, multi-component hydrodynamic model for the analysis of severe accidents in heavy water reactors with multiple-tube assemblies

    SciTech Connect

    Tentner, A.M.

    1994-03-01

    A detailed hydrodynamic fuel relocation model has been developed for the analysis of severe accidents in Heavy Water Reactors with multiple-tube Assemblies. This model describes the Fuel Disruption and Relocation inside a nuclear fuel assembly and is designated by the acronym DIANA. DIANA solves the transient hydrodynamic equations for all the moving materials in the core and treats all the relevant flow regimes. The numerical solution techniques and some of the physical models included in DIANA have been developed taking advantage of the extensive experience accumulated in the development and validation of the LEVITATE (1) fuel relocation model of SAS4A [2, 3]. The model is designed to handle the fuel and cladding relocation in both voided and partially voided channels. It is able to treat a wide range of thermal/ hydraulic/neutronic conditions and the presence of various flow regimes at different axial locations within the same hydrodynamic channel.

  14. Thermal state of the safety system, reactor, side reflector and shielding of the {open_quote}{open_quote}TOPAZ-2{close_quote}{close_quote} system under conditions of fire caused by a launcher accident at the launch pad

    SciTech Connect

    Grinberg, E.I.; Doschatov, V.V.; Nikolaev, V.S.; Sokolov, N.S.; Usov, V.A.

    1996-03-01

    The paper presents some results of calculational analyses performed to determine thermal state of the TOPAZ II safety system structure, radiation shielding, reactor without the side reflector, rods and inserts of the side reflector under conditions of fire at the launch pad when an accident occurs to a launch vehicle. {copyright} {ital 1996 American Institute of Physics.}

  15. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  16. [Regularities of changes in 137Cs content in milk in the long term after the Chernobyl nuclear reactor accident].

    PubMed

    Fesenko, S V; Pakhomov, A Iu; Pasternak, A D; Goriainov, V A; Fesenko, G A; Panov, A V

    2004-01-01

    Regularities of changes in 137Cs content in cattle milk in the long term after the Chernobyl accident have been analyzed. Contamination levels of haylands and pastures, soil properties, specific features of agricultural production and time after the fallout play a crucial role in 137Cs concentration changes in animal products. Trends have been studied that reflect the influence of these factors and their significance assessed. The half-life periods of 137Cs decay in milk vary over the period of 1994 to 2000 between 7.1 and 14.8 years and approach similar periods calculated for the long term after global radioactive fallout from nuclear weapons tests. PMID:15287266

  17. U.S./Belarus/Ukraine joint research on the biomedical effects of the Chernobyl Reactor Accident. Final report

    SciTech Connect

    Bruce Wachholz

    2000-06-20

    The National Cancer Institute has negotiated with the governments of Belarus and Ukraine (Ministers/Ministries of Health, institutions and scientists) to develop scientific research protocols to study the effects of radioactive iodine released by the Chernobyl accident upon thyroid anatomy and function in defined cohorts of persons under the age of 19 years at the time of the accident. These studies include prospective long term medical follow-up of the cohort and the reconstruction of the radiation dose to each cohort subject's thyroid. The protocol for the study in Belarus was signed by the US and Belorussian governments in May 1994 and the protocol for the study in Ukraine was signed by the US and Ukraine in May 1995. A second scientific research protocol also was negotiated with Ukraine to study the feasibility of a long term study to follow the development of leukemia and lymphoma among Ukrainian cleanup workers; this protocol was signed by the US and Ukraine in October 1996.

  18. Evaluation of near-surface ozone over Europe from the MACC reanalysis

    NASA Astrophysics Data System (ADS)

    Katragkou, E.; Zanis, P.; Tsikerdekis, A.; Kapsomenakis, J.; Melas, D.; Eskes, H.; Flemming, J.; Huijnen, V.; Inness, A.; Schultz, M. G.; Stein, O.; Zerefos, C. S.

    2015-07-01

    This work is an extended evaluation of near-surface ozone as part of the global reanalysis of atmospheric composition, produced within the European-funded project MACC (Monitoring Atmospheric Composition and Climate). It includes an evaluation over the period 2003-2012 and provides an overall assessment of the modeling system performance with respect to near-surface ozone for specific European subregions. Measurements at rural locations from the European Monitoring and Evaluation Program (EMEP) and the European Air Quality Database (AirBase) were used for the evaluation assessment. The fractional gross error of near-surface ozone reanalysis is on average 24 % over Europe, the highest found over Scandinavia (27 %) and the lowest over the Mediterranean marine stations (21 %). Near-surface ozone shows mostly a negative bias in winter and a positive bias during warm months. Assimilation reduces the bias in near-surface ozone in most of the European subregions - with the exception of Britain and Ireland and the Iberian Peninsula and its impact is mostly notable in winter. With respect to the seasonal cycle, the MACC reanalysis reproduces the photochemically driven broad spring-summer maximum of surface ozone of central and south Europe. However, it does not capture adequately the early spring peak and the shape of the seasonality at northern and north-eastern Europe. The diurnal range of surface ozone, which is as an indication of the local photochemical production processes, is reproduced fairly well, with a tendency for a small overestimation during the warm months for most subregions (especially in central and southern Europe). Possible reasons leading to discrepancies between the MACC reanalysis and observations are discussed.

  19. MACCS : Multi-Mission Atmospheric Correction and Cloud Screening tool for high-frequency revisit data processing

    NASA Astrophysics Data System (ADS)

    Petrucci, B.; Huc, M.; Feuvrier, T.; Ruffel, C.; Hagolle, O.; Lonjou, V.; Desjardins, C.

    2015-10-01

    For the production of Level2A products during Sentinel-2 commissioning in the Technical Expertise Center Sentinel-2 in CNES, CESBIO proposed to adapt the Venus Level-2 , taking advantage of the similarities between the two missions: image acquisition at a high frequency (2 days for Venus, 5 days with the two Sentinel-2), high resolution (5m for Venus, 10, 20 and 60m for Sentinel-2), images acquisition under constant viewing conditions. The Multi-Mission Atmospheric Correction and Cloud Screening (MACCS) tool was born: based on CNES Orfeo Toolbox Library, Venμs processor which was already able to process Formosat2 and VENμS data, was adapted to process Sentinel-2 and Landsat5-7 data; since then, a great effort has been made reviewing MACCS software architecture in order to ease the add-on of new missions that have also the peculiarity of acquiring images at high resolution, high revisit and under constant viewing angles, such as Spot4/Take5 and Landsat8. The recursive and multi-temporal algorithm is implemented in a core that is the same for all the sensors and that combines several processing steps: estimation of cloud cover, cloud shadow, water, snow and shadows masks, of water vapor content, aerosol optical thickness, atmospheric correction. This core is accessed via a number of plug-ins where the specificity of the sensor and of the user project are taken into account: products format, algorithmic processing chaining and parameters. After a presentation of MACCS architecture and functionalities, the paper will give an overview of the production facilities integrating MACCS and the associated specificities: the interest for this tool has grown worldwide and MACCS will be used for extensive production within the THEIA land data center and Agri-S2 project. Finally the paper will zoom on the use of MACCS during Sentinel-2 In Orbit Test phase showing the first Level-2A products.

  20. Radiation accidents.

    PubMed

    Saenger, E L

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity. PMID:3526994

  1. Radiation accidents

    SciTech Connect

    Saenger, E.L.

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity.

  2. The feasibility of using {sup 129}I to reconstruct {sup 131}I desposition from the Chernobyl reactor accident

    SciTech Connect

    Straume, T.; Marchetti, A.A.; Anspaugh, L.R.

    1996-11-01

    Radioiodine released to the atmosphere from the accident at the Chernobyl nuclear power station in the spring of 1986 resulted in large-scale thyroid-gland exposure of populations in Ukraine, Belarus, and Russia. Because of the short half life of {sup 131}I (8.04 d), adequate data on the intensities and patterns of iodine deposition were not collected, especially in the regions where the incidence of childhood-thyroid cancer is now increasing. Results are presented from a feasibility study that show that accelerator-mass-spectrometry measurements of {sup 129}I (half life 16 {times} 10{sup 6}y) in soil can be used to reconstruct {sup 131}I-deposition density and thus help in the thyroid-dosimetry effort that is now urgently needed to support epidemiologic studies of childhood-thyroid cancer in the affected regions. 32refs., 9 figs., 3 tabs.

  3. Measurement of Airborne Fission Products in Chapel Hill, NC, USA from the Kukushima Dai-ichi Reactor Accident

    SciTech Connect

    MacMullin, S.; Giovanetti, G. K.; Green, M. P.; Henning, R.; Holmes, R.; Vorren, K.

    2012-01-01

    We present measurement results of airborne fission products in Chapel Hill, NC, USA, from 62 d following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products 131I and 137Cs were measured with maximum activity concentrations of 4.2 0.6 mBq/m3 and 0.42 0.07 mBq/m3 respectively. Additional activity from 131,132I, 134,136,137Cs and 132Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  4. Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima Dai-ichi reactor accident.

    PubMed

    MacMullin, S; Giovanetti, G K; Green, M P; Henning, R; Holmes, R; Vorren, K; Wilkerson, J F

    2012-10-01

    We present measurement results of airborne fission products in Chapel Hill, NC, USA, from 62 d following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products (131)I and (137)Cs were measured with maximum activity concentrations of 4.2 ± 0.6 mBq/m(3) and 0.42 ± 0.07 mBq/m(3) respectively. Additional activity from (131,132)I, (134,136,137)Cs and (132)Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF). PMID:22348994

  5. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    SciTech Connect

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

  6. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    SciTech Connect

    Weiss, A.J.

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  7. A regional air quality forecasting system over Europe: the MACC-II daily ensemble production

    NASA Astrophysics Data System (ADS)

    Marécal, V.; Peuch, V.-H.; Andersson, C.; Andersson, S.; Arteta, J.; Beekmann, M.; Benedictow, A.; Bergström, R.; Bessagnet, B.; Cansado, A.; Chéroux, F.; Colette, A.; Coman, A.; Curier, R. L.; Denier van der Gon, H. A. C.; Drouin, A.; Elbern, H.; Emili, E.; Engelen, R. J.; Eskes, H. J.; Foret, G.; Friese, E.; Gauss, M.; Giannaros, C.; Guth, J.; Joly, M.; Jaumouillé, E.; Josse, B.; Kadygrov, N.; Kaiser, J. W.; Krajsek, K.; Kuenen, J.; Kumar, U.; Liora, N.; Lopez, E.; Malherbe, L.; Martinez, I.; Melas, D.; Meleux, F.; Menut, L.; Moinat, P.; Morales, T.; Parmentier, J.; Piacentini, A.; Plu, M.; Poupkou, A.; Queguiner, S.; Robertson, L.; Rouïl, L.; Schaap, M.; Segers, A.; Sofiev, M.; Tarasson, L.; Thomas, M.; Timmermans, R.; Valdebenito, Á.; van Velthoven, P.; van Versendaal, R.; Vira, J.; Ung, A.

    2015-09-01

    This paper describes the pre-operational analysis and forecasting system developed during MACC (Monitoring Atmospheric Composition and Climate) and continued in the MACC-II (Monitoring Atmospheric Composition and Climate: Interim Implementation) European projects to provide air quality services for the European continent. This system is based on seven state-of-the art models developed and run in Europe (CHIMERE, EMEP, EURAD-IM, LOTOS-EUROS, MATCH, MOCAGE and SILAM). These models are used to calculate multi-model ensemble products. The paper gives an overall picture of its status at the end of MACC-II (summer 2014) and analyses the performance of the multi-model ensemble. The MACC-II system provides daily 96 h forecasts with hourly outputs of 10 chemical species/aerosols (O3, NO2, SO2, CO, PM10, PM2.5, NO, NH3, total NMVOCs (non-methane volatile organic compounds) and PAN+PAN precursors) over eight vertical levels from the surface to 5 km height. The hourly analysis at the surface is done a posteriori for the past day using a selection of representative air quality data from European monitoring stations. The performance of the system is assessed daily, weekly and every 3 months (seasonally) through statistical indicators calculated using the available representative air quality data from European monitoring stations. Results for a case study show the ability of the ensemble median to forecast regional ozone pollution events. The seasonal performances of the individual models and of the multi-model ensemble have been monitored since September 2009 for ozone, NO2 and PM10. The statistical indicators for ozone in summer 2014 show that the ensemble median gives on average the best performances compared to the seven models. There is very little degradation of the scores with the forecast day but there is a marked diurnal cycle, similarly to the individual models, that can be related partly to the prescribed diurnal variations of anthropogenic emissions in the models

  8. A regional air quality forecasting system over Europe: the MACC-II daily ensemble production

    NASA Astrophysics Data System (ADS)

    Marécal, V.; Peuch, V.-H.; Andersson, C.; Andersson, S.; Arteta, J.; Beekmann, M.; Benedictow, A.; Bergström, R.; Bessagnet, B.; Cansado, A.; Chéroux, F.; Colette, A.; Coman, A.; Curier, R. L.; Denier van der Gon, H. A. C.; Drouin, A.; Elbern, H.; Emili, E.; Engelen, R. J.; Eskes, H. J.; Foret, G.; Friese, E.; Gauss, M.; Giannaros, C.; Guth, J.; Joly, M.; Jaumouillé, E.; Josse, B.; Kadygrov, N.; Kaiser, J. W.; Krajsek, K.; Kuenen, J.; Kumar, U.; Liora, N.; Lopez, E.; Malherbe, L.; Martinez, I.; Melas, D.; Meleux, F.; Menut, L.; Moinat, P.; Morales, T.; Parmentier, J.; Piacentini, A.; Plu, M.; Poupkou, A.; Queguiner, S.; Robertson, L.; Rouïl, L.; Schaap, M.; Segers, A.; Sofiev, M.; Thomas, M.; Timmermans, R.; Valdebenito, Á.; van Velthoven, P.; van Versendaal, R.; Vira, J.; Ung, A.

    2015-03-01

    This paper describes the pre-operational analysis and forecasting system developed during MACC (Monitoring Atmospheric Composition and Climate) and continued in MACC-II (Monitoring Atmospheric Composition and Climate: Interim Implementation) European projects to provide air quality services for the European continent. The paper gives an overall picture of its status at the end of MACC-II (summer 2014). This system is based on seven state-of-the art models developed and run in Europe (CHIMERE, EMEP, EURAD-IM, LOTOS-EUROS, MATCH, MOCAGE and SILAM). These models are used to calculate multi-model ensemble products. The MACC-II system provides daily 96 h forecasts with hourly outputs of 10 chemical species/aerosols (O3, NO2, SO2, CO, PM10, PM2.5, NO, NH3, total NMVOCs and PAN + PAN precursors) over 8 vertical levels from the surface to 5 km height. The hourly analysis at the surface is done a posteriori for the past day using a selection of representative air quality data from European monitoring stations. The performances of the system are assessed daily, weekly and 3 monthly (seasonally) through statistical indicators calculated using the available representative air quality data from European monitoring stations. Results for a case study show the ability of the median ensemble to forecast regional ozone pollution events. The time period of this case study is also used to illustrate that the median ensemble generally outperforms each of the individual models and that it is still robust even if two of the seven models are missing. The seasonal performances of the individual models and of the multi-model ensemble have been monitored since September 2009 for ozone, NO2 and PM10 and show an overall improvement over time. The change of the skills of the ensemble over the past two summers for ozone and the past two winters for PM10 are discussed in the paper. While the evolution of the ozone scores is not significant, there are improvements of PM10 over the past two winters

  9. A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

    SciTech Connect

    Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

    1994-09-01

    Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

  10. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  11. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  12. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    SciTech Connect

    Not Available

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

  13. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

    SciTech Connect

    Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harper, F.T.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  14. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harrison, J.D.; Harper, F.T.; Hora, S.C.

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  15. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Boardman, J.; Jones, J.A.; Harper, F.T.; Young, M.L.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  16. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

    SciTech Connect

    Haskin, F.E.; Harper, F.T.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  17. [Chernobyl nuclear power plant accident and Tokaimura criticality accident].

    PubMed

    Takada, Jun

    2012-03-01

    It is clear from inspection of historical incidents that the scale of disasters in a nuclear power plant accident is quite low level overwhelmingly compared with a nuclear explosion in nuclear war. Two cities of Hiroshima and Nagasaki were destroyed by nuclear blast with about 20 kt TNT equivalent and then approximately 100,000 people have died respectively. On the other hand, the number of acute death is 30 in the Chernobyl nuclear reactor accident. In this chapter, we review health hazards and doses in two historical nuclear incidents of Chernobyl and Tokaimura criticality accident and then understand the feature of the radiation accident in peaceful utilization of nuclear power. PMID:22514916

  18. A review of criticality accidents

    SciTech Connect

    Stratton, W R; Smith, D R

    1989-03-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Forty-one accidental power transients are reviewed. In each case where available, enough detail is given to help visualize the physical situation, the cause or causes of the accident, the history and characteristics of the transient, the energy release, and the consequences, if any, to personnel and property. Excursions associated with large power reactors are not included in this study, except that some information on the major accident at the Chernobyl reactor in April 1986 is provided in the Appendix. 67 refs., 21 figs., 2 tabs.

  19. Thermal hydraulic features of the TMI accident

    NASA Astrophysics Data System (ADS)

    Tolman, B.

    1985-10-01

    The Three Mile island (TMI)-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident scenario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiments. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors.

  20. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  1. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  2. A POTENTIAL APPLICATION OF UNCERTAINTY ANALYSIS TO DOE-STD-3009-94 ACCIDENT ANALYSIS

    SciTech Connect

    Palmrose, D E; Yang, J M

    2007-05-10

    The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG.

  3. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  4. Consistent evaluation of GOSAT, SCIAMACHY, CarbonTracker, and MACC through comparisons to TCCON

    NASA Astrophysics Data System (ADS)

    Kulawik, S. S.; Wunch, D.; O'Dell, C.; Frankenberg, C.; Reuter, M.; Oda, T.; Chevallier, F.; Sherlock, V.; Buchwitz, M.; Osterman, G.; Miller, C.; Wennberg, P.; Griffith, D. W. T.; Morino, I.; Dubey, M.; Deutscher, N. M.; Notholt, J.; Hase, F.; Warneke, T.; Sussmann, R.; Robinson, J.; Strong, K.; Schneider, M.; Wolf, J.

    2015-06-01

    Consistent validation of satellite CO2 estimates is a prerequisite for using multiple satellite CO2 measurements for joint flux inversion, and for establishing an accurate long-term atmospheric CO2 data record. We focus on validating model and satellite observation attributes that impact flux estimates and CO2 assimilation, including accurate error estimates, correlated and random errors, overall biases, biases by season and latitude, the impact of coincidence criteria, validation of seasonal cycle phase and amplitude, yearly growth, and daily variability. We evaluate dry air mole fraction (XCO2) for GOSAT (ACOS b3.5) and SCIAMACHY (BESD v2.00.08) as well as the CarbonTracker (CT2013b) simulated CO2 mole fraction fields and the MACC CO2 inversion system (v13.1) and compare these to TCCON observations (GGG2014). We find standard deviations of 0.9 ppm, 0.9, 1.7, and 2.1 ppm versus TCCON for CT2013b, MACC, GOSAT, and SCIAMACHY, respectively, with the single target errors 1.9 and 0.9 times the predicted errors for GOSAT and SCIAMACHY, respectively. When satellite data are averaged and interpreted according to error2 = a2+ b2 /n (where n are the number of observations averaged, a are the systematic (correlated) errors, and b are the random (uncorrelated) errors), we find that the correlated error term a = 0.6 ppm and the uncorrelated error term b = 1.7 ppm for GOSAT and a = 1.0 ppm, b = 1.4 ppm for SCIAMACHY regional averages. Biases at individual stations have year-to-year variability of ~ 0.3 ppm, with biases larger than the TCCON predicted bias uncertainty of 0.4 ppm at many stations. Using fitting software, we find that GOSAT underpredicts the seasonal cycle amplitude in the Northern Hemisphere (NH) between 46-53° N. In the Southern Hemisphere (SH), CT2013b underestimates the seasonal cycle amplitude. Biases are calculated for 3-month intervals and indicate the months that contribute to the observed amplitude differences. The seasonal cycle phase indicates

  5. From Fire Observations to Smoke Plume Forecasting in the MACC Services

    NASA Astrophysics Data System (ADS)

    Kaiser, Johannes W.; Benedetti, A.; Flemming, J.; Morcrette, J.-J.; Heil, A.; Schultz, M. G.; van der Werf, G. R.; Wooster, M. J.

    2011-01-01

    The MACC project is implementing atmospheric monitoring and forecasting services for the global and Euro- pean domains as part of the GMES programme. Smoke plumes are monitored by assimilating observations of aerosol optical depth and various trace gases. Biomass burning is monitored in real time by assimilating observations of fire radiative power (FRP) from five satellite- based instruments. The global monitoring capability is demonstrated with a near real time fire and smoke analysis for South America, where a threefold increase of biomass burning has been detected in 2010 compared to 2009. Furthermore, an anomalously flat diurnal cycle has been recorded for the Russian wildfires of July and August 2010. This can be interpreted as a characteristic of peaty soil burning, which entails particularly large emissions. The global aerosol service was able to forecast, with three days lead time, an air quality threshold transgression in Finland that resulted from the Russian fires.

  6. Consistent evaluation of GOSAT, SCIAMACHY, carbontracker, and MACC through comparisons to TCCON

    SciTech Connect

    Kulawik, S. S.; Wunch, D.; O'Dell, C.; Frankenberg, C.; Reuter, M.; Oda, T.; Chevallier, F.; Sherlock, V.; Buchwitz, M.; Osterman, G.; Miller, C.; Wennberg, P.; Griffith, D. W. T.; Morino, I.; Dubey, M.; Deutscher, N. M.; Notholt, J.; Hase, F.; Warneke, T.; Sussmann, R.; Robinson, J.; Strong, K.; Schneider, M.; Wolf, J.

    2015-06-22

    Consistent validation of satellite CO2 estimates is a prerequisite for using multiple satellite CO2 measurements for joint flux inversion, and for establishing an accurate long-term atmospheric CO2 data record. We focus on validating model and satellite observation attributes that impact flux estimates and CO2 assimilation, including accurate error estimates, correlated and random errors, overall biases, biases by season and latitude, the impact of coincidence criteria, validation of seasonal cycle phase and amplitude, yearly growth, and daily variability. We evaluate dry air mole fraction (XCO2) for GOSAT (ACOS b3.5) and SCIAMACHY (BESD v2.00.08) as well as the CarbonTracker (CT2013b) simulated CO2 mole fraction fields and the MACC CO2 inversion system (v13.1) and compare these to TCCON observations (GGG2014). We find standard deviations of 0.9 ppm, 0.9, 1.7, and 2.1 ppm versus TCCON for CT2013b, MACC, GOSAT, and SCIAMACHY, respectively, with the single target errors 1.9 and 0.9 times the predicted errors for GOSAT and SCIAMACHY, respectively. When satellite data are averaged and interpreted according to error2 = a2+ b2 /n (where n are the number of observations averaged, a are the systematic (correlated) errors, and b are the random (uncorrelated) errors), we find that the correlated error term a = 0.6 ppm and the uncorrelated error term b = 1.7 ppm for GOSAT and a = 1.0 ppm, b = 1.4 ppm for SCIAMACHY regional averages. Biases at individual stations have year-to-year variability of ~ 0.3 ppm, with biases larger than the TCCON predicted bias uncertainty of 0.4 ppm at many stations. Using fitting software, we find that GOSAT underpredicts the seasonal cycle amplitude in the Northern Hemisphere (NH) between 46–53° N. In the Southern Hemisphere (SH), CT2013b underestimates the

  7. Consistent evaluation of GOSAT, SCIAMACHY, carbontracker, and MACC through comparisons to TCCON

    DOE PAGESBeta

    Kulawik, S. S.; Wunch, D.; O'Dell, C.; Frankenberg, C.; Reuter, M.; Oda, T.; Chevallier, F.; Sherlock, V.; Buchwitz, M.; Osterman, G.; et al

    2015-06-22

    Consistent validation of satellite CO2 estimates is a prerequisite for using multiple satellite CO2 measurements for joint flux inversion, and for establishing an accurate long-term atmospheric CO2 data record. We focus on validating model and satellite observation attributes that impact flux estimates and CO2 assimilation, including accurate error estimates, correlated and random errors, overall biases, biases by season and latitude, the impact of coincidence criteria, validation of seasonal cycle phase and amplitude, yearly growth, and daily variability. We evaluate dry air mole fraction (XCO2) for GOSAT (ACOS b3.5) and SCIAMACHY (BESD v2.00.08) as well as the CarbonTracker (CT2013b) simulated CO2more » mole fraction fields and the MACC CO2 inversion system (v13.1) and compare these to TCCON observations (GGG2014). We find standard deviations of 0.9 ppm, 0.9, 1.7, and 2.1 ppm versus TCCON for CT2013b, MACC, GOSAT, and SCIAMACHY, respectively, with the single target errors 1.9 and 0.9 times the predicted errors for GOSAT and SCIAMACHY, respectively. When satellite data are averaged and interpreted according to error2 = a2+ b2 /n (where n are the number of observations averaged, a are the systematic (correlated) errors, and b are the random (uncorrelated) errors), we find that the correlated error term a = 0.6 ppm and the uncorrelated error term b = 1.7 ppm for GOSAT and a = 1.0 ppm, b = 1.4 ppm for SCIAMACHY regional averages. Biases at individual stations have year-to-year variability of ~ 0.3 ppm, with biases larger than the TCCON predicted bias uncertainty of 0.4 ppm at many stations. Using fitting software, we find that GOSAT underpredicts the seasonal cycle amplitude in the Northern Hemisphere (NH) between 46–53° N. In the Southern Hemisphere (SH), CT2013b underestimates the seasonal cycle amplitude. Biases are calculated for 3-month intervals and indicate the months that contribute to the observed amplitude differences. The seasonal cycle phase

  8. Reactor operation safety information document

    SciTech Connect

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  9. Accident investigation

    NASA Technical Reports Server (NTRS)

    Laynor, William G. Bud

    1987-01-01

    The National Transportation Safety Board (NTSB) has attributed wind shear as a cause or contributing factor in 15 accidents involving transport-categroy airplanes since 1970. Nine of these were nonfatal; but the other six accounted for 440 lives. Five of the fatal accidents and seven of the nonfatal accidents involved encounters with convective downbursts or microbursts. Of other accidents, two which were nonfatal were encounters with a frontal system shear, and one which was fatal was the result of a terrain induced wind shear. These accidents are discussed with reference to helping the aircraft to avoid the wind shear or if impossible to help the pilot to get through the wind shear.

  10. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  11. Accident management information needs

    SciTech Connect

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. )

    1990-04-01

    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.

  12. Cluster analysis of European surface ozone observations for evaluation of MACC reanalysis data

    NASA Astrophysics Data System (ADS)

    Lyapina, Olga; Schultz, Martin G.; Hense, Andreas

    2016-06-01

    The high density of European surface ozone monitoring sites provides unique opportunities for the investigation of regional ozone representativeness and for the evaluation of chemistry climate models. The regional representativeness of European ozone measurements is examined through a cluster analysis (CA) of 4 years of 3-hourly ozone data from 1492 European surface monitoring stations in the Airbase database; the time resolution corresponds to the output frequency of the model that is compared to the data in this study. K-means clustering is implemented for seasonal-diurnal variations (i) in absolute mixing ratio units and (ii) normalized by the overall mean ozone mixing ratio at each site. Statistical tests suggest that each CA can distinguish between four and five different ozone pollution regimes. The individual clusters reveal differences in seasonal-diurnal cycles, showing typical patterns of the ozone behavior for more polluted stations or more rural background. The robustness of the clustering was tested with a series of k-means runs decreasing randomly the size of the initial data set or lengths of the time series. Except for the Po Valley, the clustering does not provide a regional differentiation, as the member stations within each cluster are generally distributed all over Europe. The typical seasonal, diurnal, and weekly cycles of each cluster are compared to the output of the multi-year global reanalysis produced within the Monitoring of Atmospheric Composition and Climate (MACC) project. While the MACC reanalysis generally captures the shape of the diurnal cycles and the diurnal amplitudes, it is not able to reproduce the seasonal cycles very well and it exhibits a high bias up to 12 nmol mol-1. The bias decreases from more polluted clusters to cleaner ones. Also, the seasonal and weekly cycles and frequency distributions of ozone mixing ratios are better described for clusters with relatively clean signatures. Due to relative sparsity of CO and NOx

  13. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  14. MACC regional multi-model ensemble simulations of birch pollen dispersion in Europe

    NASA Astrophysics Data System (ADS)

    Sofiev, M.; Berger, U.; Prank, M.; Vira, J.; Arteta, J.; Belmonte, J.; Bergmann, K.-C.; Chéroux, F.; Elbern, H.; Friese, E.; Galan, C.; Gehrig, R.; Khvorostyanov, D.; Kranenburg, R.; Kumar, U.; Marécal, V.; Meleux, F.; Menut, L.; Pessi, A.-M.; Robertson, L.; Ritenberga, O.; Rodinkova, V.; Saarto, A.; Segers, A.; Severova, E.; Sauliene, I.; Siljamo, P.; Steensen, B. M.; Teinemaa, E.; Thibaudon, M.; Peuch, V.-H.

    2015-07-01

    This paper presents the first ensemble modelling experiment in relation to birch pollen in Europe. The seven-model European ensemble of MACC-ENS, tested in trial simulations over the flowering season of 2010, was run through the flowering season of 2013. The simulations have been compared with observations in 11 countries, all members of the European Aeroallergen Network, for both individual models and the ensemble mean and median. It is shown that the models successfully reproduced the timing of the very late season of 2013, generally within a couple of days from the observed start of the season. The end of the season was generally predicted later than observed, by 5 days or more, which is a known feature of the source term used in the study. Absolute pollen concentrations during the season were somewhat underestimated in the southern part of the birch habitat. In the northern part of Europe, a record-low pollen season was strongly overestimated by all models. The median of the multi-model ensemble demonstrated robust performance, successfully eliminating the impact of outliers, which was particularly useful since for most models this was the first experience of pollen forecasting.

  15. The Chornobyl Accident: A Comprehensive Risk Assessment

    SciTech Connect

    Poyarkov, Victor A.; Vargo, George J.; George J. Vargo

    2000-01-01

    This book provides a comprehensive of the April 1986 Chornobyl Nuclear Power Plant accident and its short and long-term effects in the fourteen years since the accident. Chapters include: cause and description of the accident; the Shelter constructed to contain the remains the destroyed reactor, radioactive wastes arising from the accident, environmental contamination, individual and collective radiation doses, societal aspects, economic impact and conclusions. Appendices on radiological units, the medical consequences of the accident, and a list of acronym and abbreviations are included.

  16. Analysis of containment performance and radiological consequences under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    SciTech Connect

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

    1994-01-01

    A severe accident study was conducted to evaluate conservatively scoped source terms and radiological consequences to support the Advanced Neutron Source (ANS) Conceptual Safety Analysis Report (CSAR). Three different types of severe accident scenarios were postulated with a view of evaluating conservatively scoped source terms. The first scenario evaluates maximum possible steaming loads and associated radionuclide transport, whereas the next scenario is geared towards evaluating conservative containment loads from releases of radionuclide vapors and aerosols with associated generation of combustible gases. The third scenario follows the prescriptions given by the 10 CFR 100 guidelines. It was included in the CSAR for demonstrating site-suitability characteristics of the ANS. Various containment configurations are considered for the study of thermal-hydraulic and radiological behaviors of the ANS containment. Severe accident mitigative design features such as the use of rupture disks were accounted for. This report describes the postulated severe accident scenarios, methodology for analysis, modeling assumptions, modeling of several severe accident phenomena, and evaluation of the resulting source term and radiological consequences.

  17. The long-term impact of a man-made disaster: An examination of a small town in the aftermath of the Three Mile Island Nuclear Reactor Accident.

    PubMed

    Goldsteen, R; Schorr, J K

    1982-03-01

    This paper explores the long-term effects of a nuclear accident on residents' perceptions of their physical and mental health, their trust of public officials, and their attitudes toward the future risks of nuclear power generation In their community. We find that in the period after the accident at Three Mile Island that there are constant or Increasing levels of distress reported by community residents. We conclude that the effects of a technological disaster may often be more enduring than those natural disaster and that greater research efforts should be made to Investigate the long-term consequences of man-made catastrophies of all types. PMID:20958512

  18. Utilization of 134Cs/137Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

    PubMed Central

    Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

    2016-01-01

    The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12–21, 2011 were identified individually by analyzing the combination of measured 134Cs/137Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of 134Cs/137Cs are different in reactor units owing to fuel burnup differences, the 134Cs/137Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2. PMID:27546490

  19. Utilization of (134)Cs/(137)Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident.

    PubMed

    Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

    2016-01-01

    The Fukushima Daiichi nuclear power reactor units that generated large amounts of airborne discharges during the period of March 12-21, 2011 were identified individually by analyzing the combination of measured (134)Cs/(137)Cs depositions on ground surfaces and atmospheric transport and deposition simulations. Because the values of (134)Cs/(137)Cs are different in reactor units owing to fuel burnup differences, the (134)Cs/(137)Cs ratio measured in the environment was used to determine which reactor unit ultimately contaminated a specific area. Atmospheric dispersion model simulations were used for predicting specific areas contaminated by each dominant release. Finally, by comparing the results from both sources, the specific reactor units that yielded the most dominant atmospheric release quantities could be determined. The major source reactor units were Unit 1 in the afternoon of March 12, 2011, Unit 2 during the period from the late night of March 14 to the morning of March 15, 2011. These results corresponded to those assumed in our previous source term estimation studies. Furthermore, new findings suggested that the major source reactors from the evening of March 15, 2011 were Units 2 and 3 and that the dominant source reactor on March 20, 2011 temporally changed from Unit 3 to Unit 2. PMID:27546490

  20. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  1. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Anshari, Rio

    2012-06-01

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  2. Probabilistic Accident Consequence Uncertainty - A Joint CEC/USNRC Study

    SciTech Connect

    Gregory, Julie J.; Harper, Frederick T.

    1999-07-28

    The joint USNRC/CEC consequence uncertainty study was chartered after the development of two new probabilistic accident consequence codes, MACCS in the U.S. and COSYMA in Europe. Both the USNRC and CEC had a vested interest in expanding the knowledge base of the uncertainty associated with consequence modeling, and teamed up to co-sponsor a consequence uncertainty study. The information acquired from the study was expected to provide understanding of the strengths and weaknesses of current models as well as a basis for direction of future research. This paper looks at the elicitation process implemented in the joint study and discusses some of the uncertainty distributions provided by eight panels of experts from the U.S. and Europe that were convened to provide responses to the elicitation. The phenomenological areas addressed by the expert panels include atmospheric dispersion and deposition, deposited material and external doses, food chain, early health effects, late health effects and internal dosimetry.

  3. Consistent evaluation of ACOS-GOSAT, BESD-SCIAMACHY, CarbonTracker, and MACC through comparisons to TCCON

    NASA Astrophysics Data System (ADS)

    Kulawik, Susan; Wunch, Debra; O'Dell, Christopher; Frankenberg, Christian; Reuter, Maximilian; Oda, Tomohiro; Chevallier, Frederic; Sherlock, Vanessa; Buchwitz, Michael; Osterman, Greg; Miller, Charles E.; Wennberg, Paul O.; Griffith, David; Morino, Isamu; Dubey, Manvendra K.; Deutscher, Nicholas M.; Notholt, Justus; Hase, Frank; Warneke, Thorsten; Sussmann, Ralf; Robinson, John; Strong, Kimberly; Schneider, Matthias; De Mazière, Martine; Shiomi, Kei; Feist, Dietrich G.; Iraci, Laura T.; Wolf, Joyce

    2016-02-01

    Consistent validation of satellite CO2 estimates is a prerequisite for using multiple satellite CO2 measurements for joint flux inversion, and for establishing an accurate long-term atmospheric CO2 data record. Harmonizing satellite CO2 measurements is particularly important since the differences in instruments, observing geometries, sampling strategies, etc. imbue different measurement characteristics in the various satellite CO2 data products. We focus on validating model and satellite observation attributes that impact flux estimates and CO2 assimilation, including accurate error estimates, correlated and random errors, overall biases, biases by season and latitude, the impact of coincidence criteria, validation of seasonal cycle phase and amplitude, yearly growth, and daily variability. We evaluate dry-air mole fraction (XCO2) for Greenhouse gases Observing SATellite (GOSAT) (Atmospheric CO2 Observations from Space, ACOS b3.5) and SCanning Imaging Absorption spectroMeter for Atmospheric CHartographY (SCIAMACHY) (Bremen Optimal Estimation DOAS, BESD v2.00.08) as well as the CarbonTracker (CT2013b) simulated CO2 mole fraction fields and the Monitoring Atmospheric Composition and Climate (MACC) CO2 inversion system (v13.1) and compare these to Total Carbon Column Observing Network (TCCON) observations (GGG2012/2014). We find standard deviations of 0.9, 0.9, 1.7, and 2.1 ppm vs. TCCON for CT2013b, MACC, GOSAT, and SCIAMACHY, respectively, with the single observation errors 1.9 and 0.9 times the predicted errors for GOSAT and SCIAMACHY, respectively. We quantify how satellite error drops with data averaging by interpreting according to error2 = a2 + b2/n (with n being the number of observations averaged, a the systematic (correlated) errors, and b the random (uncorrelated) errors). a and b are estimated by satellites, coincidence criteria, and hemisphere. Biases at individual stations have year-to-year variability of ˜ 0.3 ppm, with biases larger than the TCCON

  4. Consistent Evaluation of ACOS-GOSAT, BESD-SCIAMACHY, CarbonTracker, and MACC Through Comparisons to TCCON

    NASA Technical Reports Server (NTRS)

    Kulawik, Susan; Wunch, Debra; O’Dell, Christopher; Frankenberg, Christian; Reuter, Maximilian; Chevallier, Frederic; Oda, Tomohiro; Sherlock, Vanessa; Buchwitz, Michael; Osterman, Greg; Miller, Charles E.; Iraci, Laura T.; Wolf, Joyce

    2016-01-01

    Consistent validation of satellite CO2 estimates is a prerequisite for using multiple satellite CO2 measurements for joint flux inversion, and for establishing an accurate long-term atmospheric CO2 data record. Harmonizing satellite CO2 measurements is particularly important since the differences in instruments, observing geometries, sampling strategies, etc. imbue different measurement characteristics in the various satellite CO2 data products. We focus on validating model and satellite observation attributes that impact flux estimates and CO2 assimilation, including accurate error estimates, correlated and random errors, overall biases, biases by season and latitude, the impact of coincidence criteria, validation of seasonal cycle phase and amplitude, yearly growth, and daily variability. We evaluate dry-air mole fraction (X(sub CO2)) for Greenhouse gases Observing SATellite (GOSAT) (Atmospheric CO2 Observations from Space, ACOS b3.5) and SCanning Imaging Absorption spectroMeter for Atmospheric CHartographY (SCIAMACHY) (Bremen Optimal Estimation DOAS, BESD v2.00.08) as well as the CarbonTracker (CT2013b) simulated CO2 mole fraction fields and the Monitoring Atmospheric Composition and Climate (MACC) CO2 inversion system (v13.1) and compare these to Total Carbon Column Observing Network (TCCON) observations (GGG2012/2014). We find standard deviations of 0.9, 0.9, 1.7, and 2.1 parts per million vs. TCCON for CT2013b, MACC, GOSAT, and SCIAMACHY, respectively, with the single observation errors 1.9 and 0.9 times the predicted errors for GOSAT and SCIAMACHY, respectively. We quantify how satellite error drops with data averaging by interpreting according to (error(sup 2) equals a(sup 2) plus b(sup 2) divided by n (with n being the number of observations averaged, a the systematic (correlated) errors, and b the random (uncorrelated) errors). a and b are estimated by satellites, coincidence criteria, and hemisphere. Biases at individual stations have year

  5. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  6. Reactor System Transient Code.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  7. Thermal Reactor Safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  8. Use of artificial intelligence in severe accident diagnosis for PWRs

    SciTech Connect

    Wu, Zheng; Okrent, D.; Kastenberg, W.E.

    1995-12-31

    A combination approach of an expert system and neural networks is used to implement a prototype severe accident diagnostic system which would monitor the progression of the severe accident and provide necessary plant status information to assist the plant staff in accident management during the accident. The station blackout accident in a pressurized water reactor (PWR) is used as the study case. The current phase of research focus is on distinguishing different primary system failure modes and following the accident transient before and up to vessel breach.

  9. The diurnal variation in stratospheric ozone from the MACC reanalysis, the ERA-Interim reanalysis, WACCM and Earth observation data: characteristics and intercomparison

    NASA Astrophysics Data System (ADS)

    Schanz, A.; Hocke, K.; Kämpfer, N.; Chabrillat, S.; Inness, A.; Palm, M.; Notholt, J.; Boyd, I.; Parrish, A.; Kasai, Y.

    2014-12-01

    In this study we compare the diurnal variation in stratospheric ozone derived from free-running simulations of the Whole Atmosphere Community Climate Model (WACCM) and from reanalysis data of the atmospheric service MACC (Monitoring Atmospheric Composition and Climate) which both use a similar stratospheric chemistry module. We find good agreement between WACCM and the MACC reanalysis for the diurnal ozone variation in the high-latitude summer stratosphere based on photochemistry. In addition, we consult the ozone data product of the ERA-Interim reanalysis. The ERA-Interim reanalysis ozone system with its long-term ozone parametrization can not capture these diurnal variations in the upper stratosphere that are due to photochemistry. The good dynamics representations, however, reflects well dynamically induced ozone variations in the lower stratosphere. For the high-latitude winter stratosphere we describe a novel feature of diurnal variation in ozone where changes of up to 46.6% (3.3 ppmv) occur in monthly mean data. For this effect good agreement between the ERA-Interim reanalysis and the MACC reanalysis suggest quite similar diurnal advection processes of ozone. The free-running WACCM model seriously underestimates the role of diurnal advection processes at the polar vortex at the two tested resolutions. The intercomparison of the MACC reanalysis and the ERA-Interim reanalysis demonstrates how global reanalyses can benefit from a chemical representation held by a chemical transport model. The MACC reanalysis provides an unprecedented description of the dynamics and photochemistry of the diurnal variation of stratospheric ozone which is of high interest for ozone trend analysis and research on atmospheric tides. We confirm the diurnal variation in ozone at 5 hPa by observations of the Superconducting Submillimeter-Wave Limb-Emission Sounder (SMILES) experiment and selected sites of the Network for Detection of Atmospheric Composition Change (NDACC). The latter

  10. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  11. Detection of fuel release in a nuclear accident: a method for preconcentration and isolation of reactor-borne (239)Np using ion-specific extraction chromatography.

    PubMed

    Rosenberg, Brett L; Shozugawa, Katsumi; Steinhauser, Georg

    2015-09-01

    Although actinides are the most informative elements with respect to the nature of a nuclear accident, plutonium analysis is complicated by the background created by fallout from atmospheric nuclear explosions. Therefore, we propose (239)Np, a short-lived actinide that emits several γ rays, as a preferred proxy. The aim of this study was to screen ion specific extraction chromatography resins (RE-, TEVA-, UTEVA-, TRU-, and Actinide-Resin) for the highest possible recovery and separation of trace amounts of (239)Np from samples with large activities of fission products such as radiocesium, radioiodine, and, most importantly, radiotellurium, the latter of which causes spectral interference in gamma spectrometry through overlapping peaks with (239)Np. The investigated environmental media for these separations were aqueous solutions simulating rainwater and soil. Spiked samples containing (239)Np and the aforementioned volatile radionuclides were separated through extraction chromatographic columns to ascertain the most effective means of separating (239)Np from other fission products for detection by gamma spectroscopy. We propose a method for nuclear accident preparedness based on the use of Eichrom's RE-Resin. The proposed method was found most effective for isolating (239)Np from interfering radionuclides in both aqueous solution and soil using 8 M HNO3 as the loading solution and H2O as the eluent. The RE-Resin outperforms the more commonly used TEVA-Resin because the TEVA-Resin showed a higher affinity for interfering radiotellurium and radioiodine. PMID:26255661

  12. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2A: Accident model document, appendix

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The detailed abort sequence trees for the reference zirconium hydride (ZrH) reactor power module that have been generated for each phase of the reference Space Base program mission are presented. The trees are graphical representations of causal sequences. Each tree begins with the phase identification and the dichotomy between success and failure. The success branch shows the mission phase objective as being achieved. The failure branch is subdivided, as conditions require, into various primary initiating abort conditions.

  13. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    NASA Technical Reports Server (NTRS)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  14. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  15. A Near-real-time Data Transport System for Selected Stations in the Magnetometer Array for Cusp and Cleft Studies (MACCS)

    NASA Astrophysics Data System (ADS)

    Engebretson, M. J.; Valentic, T. A.; Stehle, R. H.; Hughes, W. J.

    2004-05-01

    The Magnetometer Array for Cusp and Cleft Studies (MACCS) is a two-dimensional array of eight fluxgate magnetometers that was established in 1992-1993 in the Eastern Canadian Arctic from 75° to over 80° MLAT to study electrodynamic interactions between the solar wind and Earth's magnetosphere and high-latitude ionosphere. A ninth site in Nain, Labrador, extends coverage down to 66° between existing Canadian and Greenland stations. Originally designed as part of NSF's GEM (Geospace Environment Modeling) Program, MACCS has contributed to the study of transients and waves at the magnetospheric boundary and in the near-cusp region as well as to large, cooperative, studies of ionospheric convection and substorm processes. Because of the limitations of existing telephone lines to each site, it has not been possible to economically access MACCS data promptly; instead, each month's collected data is recorded and mailed to the U.S. for processing and eventual posting on a publicly-accessible web site, http://space.augsburg.edu/space. As part of its recently renewed funding, NSF has supported the development of a near-real-time data transport system using the Iridium satellite network, which will be implemented at two MACCS sites in summer 2004. At the core of the new MACCS communications system is the Data Transport Network, software developed with NSF-ITR funding to automate the transfer of scientific data from remote field stations over unreliable, bandwidth-constrained network connections. The system utilizes a store-and-forward architecture based on sending data files as attachments to Usenet messages. This scheme not only isolates the instruments from network outages, but also provides a consistent framework for organizing and accessing multiple data feeds. Client programs are able to subscribe to data feeds to perform tasks such as system health monitoring, data processing, web page updates and e-mail alerts. The MACCS sites will employ the Data Transport Network

  16. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    SciTech Connect

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  17. Severe accident analysis using dynamic accident progression event trees

    NASA Astrophysics Data System (ADS)

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

  18. Brookhaven lecture series No. 227: The Chernobyl accident

    SciTech Connect

    Kouts, H.

    1986-09-24

    This lecture discusses the events leading to, during, and following the Chernobyl Reactor number 4 accident. A description of the light water cooled, graphite moderated reactor, the reactor site conditions leading to meltdown is presented. The emission of radioactive effluents and the biological radiation effects is also discussed. (FI)

  19. Nuclear Reactor Safety: a current awareness bulletin

    SciTech Connect

    Cunningham, D.C.

    1985-01-15

    Nuclear Reactor Safety announces on a semimonthly basis the current worldwide information available on all safety-related aspects of fission reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures.

  20. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  1. Nuclear accident dosimetry intercomparison studies.

    PubMed

    Sims, C S

    1989-09-01

    Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry. PMID:2777549

  2. 1994 Accident sequence precursor program results

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.

    1996-01-01

    The Accident Sequence Precursor (ASP) Program involves the systematic review and evaluation of operational events that have occurred at light-water reactors to identify and categorize precursors to potential severe core damage accident sequences. The results of the ASP Program are published in an annual report. The most recent report, which contains the analyses of the precursors for 1994, is NUREG/CR-4674, Vols. 21 and 22, Precursors to Potential Severe Core Damage Accidents: 1994, A Status Report, published in December 1995. This article provides an overview of the ASP review and evaluation process and a summary of the results for 1994. 12 refs., 2 figs., 4 tabs.

  3. 1995 Accident Sequence Precursor (ASP) Program results

    SciTech Connect

    Muhlheim, M.D.; Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; O`Reilly, P.D.; Dolan, B.W.; Minarick, J.W.

    1997-01-01

    The Accident Sequence Precursor (ASP) Program involves the systematic review and evaluation of operational events that have occurred at light-water reactors to identify and categorize precursors to potential severe core damage accident sequences. The results of the ASP Program are published in an annual report. The most recent report, which contains the precursors for 1995, is NUREG/CR-4674, Volume 23, Precursors to Potential Severe Core Damage Accidents: 1995, A Status Report, published in April 1997. This article provides an overview of the ASP review and evaluation process and a summary of the results for 1995.

  4. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    SciTech Connect

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-12-31

    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 ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause 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 parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation.

  5. [The radiation accident].

    PubMed

    Stögmann, W

    1988-08-26

    The reactor accident of Chernobyl in April 1986 has shown us all the dangers which are inherent ever in the peaceful use of atomic energy. The effects of exposure to ionizing radiation are dependent on biological effectiveness, on dose, on duration of exposure and on the age of the exposed person (the younger the graver). Acute ionizing radiation of the whole body leads to radiation disease or radiation syndrome of different stages of severity according to dosage. If the patient survives other consequences of ionizing radiation may arise: non-stochastic effects such as cataracts, keloid formation, fibrosis of the lungs and infertility) and stochastic effects (oncogenesis and mutagenesis). The sensitivity to ionizing radiation is especially high in childhood because of the high velocity of cell metabolism and cell growth, the large body-surface area and because their repair mechanism following radiation damage is not yet. PMID:3188527

  6. A 3-D evaluation of the MACC reanalysis dust product over the greater European region using CALIOP/CALIPSO satellite observations

    NASA Astrophysics Data System (ADS)

    Georgoulias, Aristeidis K.; Tsikerdekis, Athanasios; Amiridis, Vassilis; Marinou, Eleni; Benedetti, Angela; Zanis, Prodromos; Kourtidis, Konstantinos

    2016-04-01

    Significant amounts of dust are being transferred on an annual basis over the Mediterranean Basin and continental Europe from Northern Africa (Sahara Desert) and Middle East (Arabian Peninsula) as well as from other local sources. Dust affects a number of processes in the atmosphere modulating weather and climate also having an impact on human health and the economy. Therefore, the ability of simulating adequately the amount and optical properties of dust is essential. This work focuses on the evaluation of the MACC reanalysis dust product over the regions mentioned above. The evaluation procedure is based on pure dust satellite retrievals from CALIOP/CALIPSO that cover the period 2007-2012. The CALIOP/CALIPSO data utilized here come from an optimized retrieval scheme that was originally developed within the framework of the LIVAS (Lidar Climatology of Vertical Aerosol Structure for Space-Based LIDAR Simulation Studies) project. CALIOP/CALIPSO dust extinction coefficients and dust optical depth patterns at 532 nm are used for the validation of MACC natural aerosol extinction coefficients and dust optical depth patterns at 550 nm. Overall, it is shown in this work that space-based lidars may play a major role in the improvement of the MACC aerosol product. This research has been financed under the FP7 Programme MarcoPolo (Grand Number 606953, Theme SPA.2013.3.2-01).

  7. New petrophysical magnetic methods MACC and MAFM in permeability characterisation of petroleum reservoir rock cleaning, flooding modelling and determination of fines migration in formation damage

    NASA Astrophysics Data System (ADS)

    Ivakhnenko, O. P.

    2012-04-01

    Potential applications of magnetic techniques and methods in petroleum engineering and petrophysics (Ivakhnenko, 1999, 2006; Ivakhnenko & Potter, 2004) reveal their vast advantages for the petroleum reservoir characterisation and formation evaluation. In this work author proposes for the first time developed systematic methods of the Magnetic Analysis of Core Cleaning (MACC) and Magnetic Analysis of Fines Migration (MAFM) for characterisation of reservoir core cleaning and modelling estimations of fines migration for the petroleum reservoir formations. Using example of the one oil field we demonstrate results in application of these methods on the reservoir samples. Petroleum reservoir cores samples have been collected within reservoir using routine technique of reservoir sampling and preservation for PVT analysis. Immediately before the MACC and MAFM studies samples have been exposed to atmospheric air for a few days. The selected samples have been in detailed way characterised after fluid cleaning and core flooding by their mineralogical compositions and petrophysical parameters. Mineralogical composition has been estimated utilizing XRD techniques. The petrophysical parameters, such as permeability and porosity have been measured on the basis of total core analysis. The results demonstrate effectiveness and importance of the MACC and MAFM methods for the routine core analysis (RCAL) and the special core analysis (SCAL) in the reservoir characterisation, core flooding and formation damage analysis.

  8. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    SciTech Connect

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  9. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory

    SciTech Connect

    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.

  10. Severe accident testing of electrical penetration assemblies

    SciTech Connect

    Clauss, D.B. )

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  11. TMI-2 accident: core heat-up analysis

    SciTech Connect

    Ardron, K.H.; Cain, D.G.

    1981-01-01

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

  12. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  13. Summary of the SRS Severe Accident Analysis Program, 1987--1992

    SciTech Connect

    Long, T.A.; Hyder, M.L.; Britt, T.E.; Allison, D.K.; Chow, S.; Graves, R.D.; DeWald, A.B. Jr.; Monson, P.R. Jr.; Wooten, L.A.

    1992-11-01

    The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe accidents in SRS reactors that is adequate to support safety documentation for these reactors, including the Safety Analysis Report (SAR), the Probabilistic Risk Assessment (PRA), and other studies evaluating the safety of reactor operation; To provide tools and bases for the evaluation of existing or proposed safety related equipment in the SRS reactors; To provide bases for the development of accident management procedures for the SRS reactors; To develop and maintain on the site a sufficient body of knowledge, including documents, computer codes, and cognizant engineers and scientists, that can be used to authoritatively resolve questions or issues related to reactor accidents. The Severe Accident Analysis Program was instituted in 1987 and has already produced a substantial amount of information, and specialized calculational tools. Products of the Severe Accident Analysis Program (listed in Section 9 of this report) have been used in the development of the Safety Analysis Report (SAR) and the Probabilistic Risk Assessment (PRA), and in the development of technical specifications for the SRS reactors. A staff of about seven people is currently involved directly in the program and in providing input on severe accidents to other SRS activities.

  14. Fukushima Daiichi Accident and Its Radiological Impact on the Environment

    ERIC Educational Resources Information Center

    Bevelacqua, J. J.

    2012-01-01

    The Fukushima Daiichi nuclear accident is a topic of current media and public interest. It provides a means to motivate students to understand the fission process and the barriers that have been designed to prevent the release of fission products to the environment following a major nuclear power reactor accident. The Fukushima Daiichi accident…

  15. Applicability of health physics lessons learned from the Three Mile Island Unit 2 accident to the Fukushima Daiichi accident.

    PubMed

    Bevelacqua, J J

    2012-02-01

    The TMI-2 and Fukushima Daiichi accidents appear to be dissimilar because they involve different reactor types. However, the health physics related lessons learned from TMI-2 are applicable, and can enhance the Fukushima Daiichi recovery effort. PMID:22230016

  16. Assessment of CRBR core disruptive accident energetics

    SciTech Connect

    Theofanous, T.G.; Bell, C.R.

    1984-03-01

    The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly.

  17. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  18. Perspectives on reactor safety

    SciTech Connect

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  19. Review of models applicable to accident aerosols

    SciTech Connect

    Glissmeyer, J.A.

    1983-07-01

    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.

  20. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-01-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  1. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-11-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  2. Full-length fuel rod behavior under severe accident conditions

    SciTech Connect

    Lombardo, N J; Lanning, D D; Panisko, F E

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

  3. Review of light water reactor safety

    SciTech Connect

    Cheng, H.S.

    1980-12-01

    A review of the present status of light water reactor (LWR) safety is presented. The review starts with a brief discussion of the outstanding accident scenarios concerning LWRs. Where possible the areas of present technological uncertainties are stressed. To provide a better perspective of reactor safety, it then reviews the probabilistic assessment of the outstanding LWR accidents considered in the Reactor Safety Study (WASH-1400) and discusses the potential impact of the present technological uncertainties on WASH-1400.

  4. Summary of a workshop on severe accident management for BWRs

    SciTech Connect

    Kastenberg, W.E.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D.

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of strategies there may be several options available to the operator; and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrument behavior during an accident. During the period September 26--28, 1990, a workshop was held at the University of California, Los Angeles, to address these uncertainties for Boiling Water Reactors (BWRs). This report contains a summary of the workshop proceedings.

  5. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project.

  6. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

  7. Planetary Boundary Layer (PBL) Heights Derived From NASA Langley Airborne High Spectral Resolution Lidar (HSRL) Data Acquired During TexAQS/GoMACCS, CHAPS, and MILAGRO

    NASA Astrophysics Data System (ADS)

    Burton, S. P.; Ferrare, R. A.; Hostetler, C. A.; Hair, J. W.; Cook, A.; Harper, D.; Obland, M. D.; Rogers, R. R.

    2007-12-01

    The NASA Langley Research Center airborne High Spectral Resolution Lidar (HSRL) was deployed on the NASA Langley B-200 King Air aircraft in the Mexico City metropolitan area during the Mega-city Initiative: Local and Global Research Observations (MILAGRO) campaign in March 2006; in the Houston metropolitan area during the Texas Air Quality Study (TexAQS)/Gulf of Mexico Atmospheric Composition and Climate Study (GoMACCS) in August and September 2006; and in the Oklahoma City area during Cumulus Humilis Aerosol Processing Study (CHAPS) in June 2007. The HSRL instrument measures profiles of aerosol extinction, backscatter and depolarization. The height of the Planetary Boundary Layer was derived by identifying sharp gradients in the HSRL 532-nm aerosol backscatter signal profiles using an automated technique based on Brooks (2003) [I.M. Brooks, Finding Boundary Layer Top: Application of Wavelet Covariance Transform to Lidar Backscatter Profiles. Journal of Atmospheric and Oceanic Technology 20, 1092-1105, 2003]. The technique uses a Haar wavelet covariance transform with multiple wavelet dilation values to adapt to non-ideal conditions where there can be gradients in the background signals and the boundary layer can be ill defined. The technique also identifies the top and bottom of the transition (i.e. entrainment) zone. We have further modified the algorithm to find PBL heights using HSRL backscatter data acquired during GoMACCS and MILAGRO, where complex terrain and overlying aerosol layers further complicate identifying the boundary layer. In addition, PBL heights are derived from HSRL backscatter data acquired during the CHAPS campaign, in another urban environment where the terrain is not as complex. We will describe the algorithm modifications we have made and show boundary layer heights and transition zone thicknesses for HSRL measurements over the Oklahoma City, Houston, and Mexico City areas during CHAPS, TexAQS/GoMACCS, and MILAGRO.

  8. Seasonal and interannual variability of carbon monoxide based on MOZAIC observations, MACC reanalysis, and model simulations over an urban site in India

    NASA Astrophysics Data System (ADS)

    Sheel, Varun; Sahu, L. K.; Kajino, M.; Deushi, M.; Stein, O.; Nedelec, P.

    2014-07-01

    The spatial and temporal variations of carbon monoxide (CO) are analyzed over a tropical urban site, Hyderabad (17°27'N, 78°28'E) in central India. We have used vertical profiles from the Measurement of ozone and water vapor by Airbus in-service aircraft (MOZAIC) aircraft observations, Monitoring Atmospheric Composition and Climate (MACC) reanalysis, and two chemical transport model simulations (Model for Ozone And Related Tracers (MOZART) and MRI global Chemistry Climate Model (MRI-CCM2)) for the years 2006-2008. In the lower troposphere, the CO mixing ratio showed strong seasonality, with higher levels (>300 ppbv) during the winter and premonsoon seasons associated with a stable anticyclonic circulation, while lower CO values (up to 100 ppbv) were observed in the monsoon season. In the planetary boundary layer (PBL), the seasonal distribution of CO shows the impact of both local meteorology and emissions. While the PBL CO is predominantly influenced by strong winds, bringing regional background air from marine and biomass burning regions, under calm conditions CO levels are elevated by local emissions. On the other hand, in the free troposphere, seasonal variation reflects the impact of long-range transport associated with the Intertropical Convergence Zone and biomass burning. The interannual variations were mainly due to transition from El Niño to La Niña conditions. The overall modified normalized mean biases (normalization based on the observed and model mean values) with respect to the observed CO profiles were lower for the MACC reanalysis than the MOZART and MRI-CCM2 models. The CO in the PBL region was consistently underestimated by MACC reanalysis during all the seasons, while MOZART and MRI-CCM2 show both positive and negative biases depending on the season.

  9. World commercial aircraft accidents

    SciTech Connect

    Kimura, C.Y.

    1993-01-01

    This report is a compilation of all accidents world-wide involving aircraft in commercial service which resulted in the loss of the airframe or one or more fatality, or both. This information has been gathered in order to present a complete inventory of commercial aircraft accidents. Events involving military action, sabotage, terrorist bombings, hijackings, suicides, and industrial ground accidents are included within this list. Included are: accidents involving world commercial jet aircraft, world commercial turboprop aircraft, world commercial pistonprop aircraft with four or more engines and world commercial pistonprop aircraft with two or three engines from 1946 to 1992. Each accident is presented with information in the following categories: date of the accident, airline and its flight numbers, type of flight, type of aircraft, aircraft registration number, construction number/manufacturers serial number, aircraft damage, accident flight phase, accident location, number of fatalities, number of occupants, cause, remarks, or description (brief) of the accident, and finally references used. The sixth chapter presents a summary of the world commercial aircraft accidents by major aircraft class (e.g. jet, turboprop, and pistonprop) and by flight phase. The seventh chapter presents several special studies including a list of world commercial aircraft accidents for all aircraft types with 100 or more fatalities in order of decreasing number of fatalities, a list of collision accidents involving commercial aircrafts, and a list of world commercial aircraft accidents for all aircraft types involving military action, sabotage, terrorist bombings, and hijackings.

  10. Evaluation of the MACC operational forecast system - potential and challenges of global near-real-time modelling with respect to reactive gases in the troposphere

    NASA Astrophysics Data System (ADS)

    Wagner, A.; Blechschmidt, A.-M.; Bouarar, I.; Brunke, E.-G.; Clerbaux, C.; Cupeiro, M.; Cristofanelli, P.; Eskes, H.; Flemming, J.; Flentje, H.; George, M.; Gilge, S.; Hilboll, A.; Inness, A.; Kapsomenakis, J.; Richter, A.; Ries, L.; Spangl, W.; Stein, O.; Weller, R.; Zerefos, C.

    2015-03-01

    Monitoring Atmospheric Composition and Climate (MACC/MACCII) currently represents the European Union's Copernicus Atmosphere Monitoring Service (CAMS) (http://www.copernicus.eu), which will become fully operational in the course of 2015. The global near-real-time MACC model production run for aerosol and reactive gases provides daily analyses and 5 day forecasts of atmospheric composition fields. It is the only assimilation system world-wide that is operational to produce global analyses and forecasts of reactive gases and aerosol fields. We have investigated the ability of the MACC analysis system to simulate tropospheric concentrations of reactive gases (CO, O3, and NO2) covering the period between 2009 and 2012. A validation was performed based on CO and O3 surface observations from the Global Atmosphere Watch (GAW) network, O3 surface observations from the European Monitoring and Evaluation Programme (EMEP) and furthermore, NO2 tropospheric columns derived from the satellite sensors SCIAMACHY and GOME-2, and CO total columns derived from the satellite sensor MOPITT. The MACC system proved capable of reproducing reactive gas concentrations in consistent quality, however, with a seasonally dependent bias compared to surface and satellite observations: for northern hemispheric surface O3 mixing ratios, positive biases appear during the warm seasons and negative biases during the cold parts of the years, with monthly Modified Normalised Mean Biases (MNMBs) ranging between -30 and 30% at the surface. Model biases are likely to result from difficulties in the simulation of vertical mixing at night and deficiencies in the model's dry deposition parameterization. Observed tropospheric columns of NO2 and CO could be reproduced correctly during the warm seasons, but are mostly underestimated by the model during the cold seasons, when anthropogenic emissions are at a highest, especially over the US, Europe and Asia

  11. Evaluation of the MACC operational forecast system - potential and challenges of global near-real-time modelling with respect to reactive gases in the troposphere

    NASA Astrophysics Data System (ADS)

    Wagner, A.; Blechschmidt, A.-M.; Bouarar, I.; Brunke, E.-G.; Clerbaux, C.; Cupeiro, M.; Cristofanelli, P.; Eskes, H.; Flemming, J.; Flentje, H.; George, M.; Gilge, S.; Hilboll, A.; Inness, A.; Kapsomenakis, J.; Richter, A.; Ries, L.; Spangl, W.; Stein, O.; Weller, R.; Zerefos, C.

    2015-12-01

    The Monitoring Atmospheric Composition and Climate (MACC) project represents the European Union's Copernicus Atmosphere Monitoring Service (CAMS) (http://www.copernicus.eu/), which became fully operational during 2015. The global near-real-time MACC model production run for aerosol and reactive gases provides daily analyses and 5-day forecasts of atmospheric composition fields. It is the only assimilation system worldwide that is operational to produce global analyses and forecasts of reactive gases and aerosol fields. We have investigated the ability of the MACC analysis system to simulate tropospheric concentrations of reactive gases covering the period between 2009 and 2012. A validation was performed based on carbon monoxide (CO), nitrogen dioxide (NO2) and ozone (O3) surface observations from the Global Atmosphere Watch (GAW) network, the O3 surface observations from the European Monitoring and Evaluation Programme (EMEP) and, furthermore, NO2 tropospheric columns, as well as CO total columns, derived from satellite sensors. The MACC system proved capable of reproducing reactive gas concentrations with consistent quality; however, with a seasonally dependent bias compared to surface and satellite observations - for northern hemispheric surface O3 mixing ratios, positive biases appear during the warm seasons and negative biases during the cold parts of the year, with monthly modified normalised mean biases (MNMBs) ranging between -30 and 30 % at the surface. Model biases are likely to result from difficulties in the simulation of vertical mixing at night and deficiencies in the model's dry deposition parameterisation. Observed tropospheric columns of NO2 and CO could be reproduced correctly during the warm seasons, but are mostly underestimated by the model during the cold seasons, when anthropogenic emissions are at their highest level, especially over the US, Europe and Asia. Monthly MNMBs of the satellite data

  12. Inter-comparison between HERMESv2.0 and TNO-MACC-II emission data using the CALIOPE air quality system (Spain)

    NASA Astrophysics Data System (ADS)

    Guevara, Marc; Pay, María Teresa; Martínez, Francesc; Soret, Albert; Denier van der Gon, Hugo; Baldasano, José M.

    2014-12-01

    This work examines and compares the performance of two emission datasets on modelling air quality concentrations for Spain: (i) the High-Elective Resolution Modelling Emissions System (HERMESv2.0) and (ii) the TNO-MACC-II emission inventory. For this purpose, the air quality system CALIOPE-AQFS (WRF-ARW/CMAQ/BSC-DREAM8b) was run over Spain for February and June 2009 using the two emission datasets (4 km × 4 km and 1 h). Nitrogen dioxide (NO2), sulphur dioxide (SO2), Ozone (O3) and particular matter (PM10) modelled concentrations were compared with measurements at different type of air quality stations (i.e. rural background, urban, suburban industrial). A preliminary emission comparison showed significant discrepancies between the two datasets, highlighting an overestimation of industrial emissions in urban areas when using TNO-MACC-II. However, simulations showed similar performances of both emission datasets in terms of air quality. Modelled NO2 concentrations were similar between both datasets at the background stations, although TNO-MACC-II presented lower underestimations due to differences in industrial, other mobile sources and residential emissions. At Madrid urban stations NO2 was significantly underestimated in both cases despite the fact that HERMESv2.0 estimates traffic emissions using a more local information and detailed methodology. This NO2 underestimation problem was not found in Barcelona due to the influence of international shipping emissions located in the coastline. An inadequate characterization of some TNO-MACC-II's point sources led to high SO2 biases at industrial stations, especially in northwest Spain where large facilities are grouped. In general, surface O3 was overestimated regardless of the emission dataset used, depicting the problematic of CMAQ on overestimating low ozone at night. On the other hand, modelled PM10 concentrations were less underestimated in urban areas when applying HERMESv2.0 due to the inclusion of road dust

  13. The Verification of MACC-II Global Reactive Gas Forecasts with Global Atmosphere Watch Surface Observations and Ozonesonde Measurements from the NDACC, WOUDC, NILU and SHADOZ Databases

    NASA Astrophysics Data System (ADS)

    Wagner, A.

    2012-04-01

    The MACC-II (Monitoring Atmospheric Composition and Climate- phase II) project, funded under the 7th framework program of the European Union, provides a comprehensive monitoring and operational forecasting system for atmospheric constituents relevant to climate and air quality issues and surface solar radiation. The MACC-II forecast system is based on the global weather forecasting system operated by the European Centre for Medium-Range Weather Forecasts (ECMWF) coupled with the chemistry transport models MOZART (Model for OZone and Related chemical Tracers) and TM5. On a near real time (NRT) basis observational data and forecasts, as well as reanalyses are made available for greenhouse and reactive gases, UV radiation and aerosol optical density. The sub group "VAL" is focusing on the evaluation of reactive gases, thus, stratospheric and tropospheric ozone as well as its precursors and aerosols. The GAW (Global Atmosphere Watch) network provides the ground-based observational data for the evaluation of model simulation forecast and reanalysis of the reactive gases CO and O3 at surface levels on a global scale. Contributing stations in this validation process provide their data in rapid delivery mode (within 1 day to 1 month), thus enabling a fast evaluation process. Currently, there are 12 stations providing data in Near-Real-Time. The validation process is performed online and daily updates of the results are displayed on the MACC website (http://www.gmes-atmosphere.eu/d/services/gac/verif/grg/gaw/). For the validation of stratospheric and free tropospheric ozone forecasts, balloon sonde measurements from the data bases NDACC (Network for the Detection of Atmospheric Composition Change), WOUDC (World Ozone and Ultraviolet Radiation Data Centre), NILU (Norwegian Institute for Air Research) and SHADOZ (Southern Hemisphere Additional OZon Sondes) are used. Here, we will give an overview on the status and the results of this Near-Real-Time (NRT) validation of the

  14. Designing an Experimental "Accident"

    ERIC Educational Resources Information Center

    Picker, Lester

    1974-01-01

    Describes an experimental "accident" that resulted in much student learning, seeks help in the identification of nematodes, and suggests biology teachers introduce similar accidents into their teaching to stimulate student interest. (PEB)

  15. Accident tolerant fuels for LWRs: A perspective

    NASA Astrophysics Data System (ADS)

    Zinkle, S. J.; Terrani, K. A.; Gehin, J. C.; Ott, L. J.; Snead, L. L.

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  16. Dose estimates from the Chernobyl accident

    SciTech Connect

    Lange, R.; Dickerson, M.H.; Gudiksen, P.H.

    1987-11-01

    The Lawrence Livermore National Laboratory Atmospheric Release Advisory Capability (ARAC) responded to the Chernobyl nuclear reactor accident in the Soviet Union by utilizing long-range atmospheric dispersion modeling to estimate the amount of radioactivity released (source term) and the radiation dose distribution due to exposure to the radioactive cloud over Europe and the Northern Hemisphere. In later assessments, after the release of data on the accident by the Soviet Union, the ARAC team used their mesoscale to regional scale model to focus in on the radiation dose distribution within the Soviet Union and the vicinity of the Chernobyl plant. 22 refs., 5 figs., 5 tabs.

  17. SILAM and MACC reanalysis aerosol data used for simulating the aerosol direct radiative effect with the NWP model HARMONIE for summer 2010 wildfire case in Russia

    NASA Astrophysics Data System (ADS)

    Toll, V.; Reis, K.; Ots, R.; Kaasik, M.; Männik, A.; Prank, M.; Sofiev, M.

    2015-11-01

    Persistent high pressure conditions over the European part of Russia during summer 2010 were responsible for an extended period of hot and dry weather, creating favourable conditions for severe wildfires. The chemical transport model SILAM is used to simulate the dispersion of smoke aerosol for this case. Aerosol fields from SILAM are compared to the Monitoring Atmospheric Composition and Climate (MACC) reanalysis. Moreover, the model output is compared to in situ and remote sensing measurements, paying particular attention to the most intense fire period of August 7 to 9, when the plume reached the Baltic countries and Finland. The maximum observed aerosol optical depth was more than 4 at 550 nm during this time. The aerosol distributions from the SILAM run and the MACC reanalysis are subsequently used in meteorological simulations using the Hirlam Aladin Research for Mesoscale Operational Numerical Weather Prediction in Euromed (HARMONIE) model. The modelling results show a significant reduction of the daily average shortwave radiation fluxes at the surface (up to 125 W/m2) and daily average near-surface temperature (up to 4 °C) through the aerosol direct radiative effect. The simulated near-surface temperature and vertical temperature profile agree better with the observations, when the aerosol direct radiative effect is considered in the meteorological simulation. The boundary layer is more stably stratified, creating poorer dispersion conditions for the smoke.

  18. Cloud Nucleating Properties of Aerosols During TexAQS - GoMACCS 2006: Influence of Aerosol Sources, Composition, and Size

    NASA Astrophysics Data System (ADS)

    Quinn, P. K.; Bates, T. S.; Coffman, D. J.; Covert, D. S.; Onasch, T. B.; Alllan, J. D.; Worsnop, D.

    2006-12-01

    TexAQS - GoMACCS 2006 was conducted from July to September 2006 in the Gulf of Mexico and Houston Ship Channel to investigate sources and processing of gas and particulate phase species and to determine their impact on regional air quality and climate. As part of the experiment, the NOAA R.V. Ronald H. Brown transited from Charleston, S.C. to the study region. The ship was equipped with a full compliment of gas and aerosol instruments. To determine the cloud nucleating properties of aerosols, measurements were made of the aerosol number size distribution, aerosol chemical composition, and cloud condensation nuclei (CCN) concentration at five supersaturations. During the transit and over the course of the experiment, a wide range of aerosol sources and types was encountered. These included urban and industrial emissions from the S.E. U.S. as the ship left Charleston, a mixture of Saharan dust and marine aerosol during the transit around Florida and across the Gulf of Mexico, urban emissions from Houston, and emissions from the petrochemical industries, oil platforms, and marine vessels in the Gulf coast region. Highest activation ratios (ratio of CCN to total particle number concentration at 0.4 percent supersaturation) were measured in anthropogenic air masses when the aerosol was composed primarily of ammonium sulfate salts and in marine air masses with an aerosol composed of sulfate and sea salt. A strong gradient in activation ratio was measured as the ship moved from the Gulf of Mexico to the end of the Houston Ship Channel (values decreasing from about 0.8 to less than 0.1) and the aerosol changed from marine to industrial. The activation ratio under these different regimes in addition to downwind of marine vessels and oil platforms will be discussed in the context of the aerosol size distribution and chemical composition. The discussion of composition will include the organic mass fraction of the aerosol, the degree of oxidation of the organics, and the water

  19. GOME-2 total ozone columns from MetOp-A/MetOp-B and assimilation in the MACC system

    NASA Astrophysics Data System (ADS)

    Hao, N.; Koukouli, M. E.; Inness, A.; Valks, P.; Loyola, D. G.; Zimmer, W.; Balis, D. S.; Zyrichidou, I.; Van Roozendael, M.; Lerot, C.; Spurr, R. J. D.

    2014-03-01

    The two Global Ozone Monitoring Instrument (GOME-2) sensors operated in tandem are flying onboard EUMETSAT's MetOp-A and MetOp-B satellites, launched in October 2006 and September 2012 respectively. This paper presents the operational GOME-2/MetOp-A (GOME-2A) and GOME-2/MetOp-B (GOME-2B) total ozone products provided by the EUMETSAT Satellite Application Facility on Ozone and Atmospheric Chemistry Monitoring (O3M-SAF). These products are generated using the latest version of the GOME Data Processor (GDP version 4.7). The enhancements in GDP 4.7, including the application of Brion-Daumont-Malicet ozone absorption cross-sections, are presented here. On a global scale, GOME-2B has the same high accuracy as the corresponding GOME-2A products. There is an excellent agreement between the ozone total columns from the two sensors, with GOME-2B values slightly lower with a mean difference of only 0.55 ± 0.29%. First global validation results for 6 months of GOME-2B total ozone using ground-based measurements show that on average the GOME-2B total ozone data obtained with GDP 4.7 slightly overestimate Dobson observations by about 2.0 ± 1.0% and Brewer observations by about 1.0 ± 0.8%. It is concluded that the total ozone columns (TOCs) provided by GOME-2A and GOME-2B are consistent and may be used simultaneously without introducing trends or other systematic effects. GOME-2A total ozone data have been used operationally in the Copernicus atmospheric service project MACC-II (Monitoring Atmospheric Composition and Climate - Interim Implementation) near-real-time (NRT) system since October 2013. The magnitude of the bias correction needed for assimilating GOME-2A ozone is reduced (to about -6 DU in the global mean) when the GOME-2 ozone retrieval algorithm changed to GDP 4.7.

  20. GOME-2 total ozone columns from MetOp-A/MetOp-B and assimilation in the MACC system

    NASA Astrophysics Data System (ADS)

    Hao, N.; Koukouli, M. E.; Inness, A.; Valks, P.; Loyola, D. G.; Zimmer, W.; Balis, D. S.; Zyrichidou, I.; Van Roozendael, M.; Lerot, C.; Spurr, R. J. D.

    2014-09-01

    The two Global Ozone Monitoring Instrument (GOME-2) sensors operated in tandem are flying onboard EUMETSAT's (European Organisation for the Exploitation of Meteorological Satellites) MetOp-A and MetOp-B satellites, launched in October 2006 and September 2012 respectively. This paper presents the operational GOME-2/MetOp-A (GOME-2A) and GOME-2/MetOp-B (GOME-2B) total ozone products provided by the EUMETSAT Satellite Application Facility on Ozone and Atmospheric Chemistry Monitoring (O3M-SAF). These products are generated using the latest version of the GOME Data Processor (GDP version 4.7). The enhancements in GDP 4.7, including the application of Brion-Daumont-Malicet ozone absorption cross sections, are presented here. On a global scale, GOME-2B has the same high accuracy as the corresponding GOME-2A products. There is an excellent agreement between the ozone total columns from the two sensors, with GOME-2B values slightly lower with a mean difference of only 0.55±0.29%. First global validation results for 6 months of GOME-2B total ozone using ground-based measurements show that on average the GOME-2B total ozone data obtained with GDP 4.7 are slightly higher than, both, Dobson observations by about 2.0±1.0% and Brewer observations by about 1.0±0.8%. It is concluded that the total ozone columns (TOCs) provided by GOME-2A and GOME-2B are consistent and may be used simultaneously without introducing systematic effects, which has been illustrated for the Antarctic ozone hole on 18 October 2013. GOME-2A total ozone data have been used operationally in the Copernicus atmospheric service project MACC-II (Monitoring Atmospheric Composition and Climate - Interim Implementation) near-real-time (NRT) system since October 2013. The magnitude of the bias correction needed for assimilating GOME-2A ozone is reduced (to about -6 DU in the global mean) when the GOME-2 ozone retrieval algorithm changed to GDP 4.7.

  1. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

  2. Visualization of Traffic Accidents

    NASA Technical Reports Server (NTRS)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

  3. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  4. A Review of Criticality Accidents 2000 Revision

    SciTech Connect

    Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

    2000-05-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

  5. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect

    Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

    2014-10-01

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being

  6. A Methodology for the Neutronics Design of Space Nuclear Reactors

    SciTech Connect

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

    2004-02-04

    A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

  7. Laser accidents: Being Prepared

    SciTech Connect

    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.

  8. Soviet space nuclear reactor incidents - Perception versus reality

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1992-01-01

    Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

  9. The Chernobyl accident: Causes and consequences

    SciTech Connect

    Malinauskas, A.P.

    1987-01-01

    Two explosions, one immediately following the other, in Unit 4 of the Chernobyl nuclear power station in the Soviet Union signaled the worst disaster ever to befall the commercial nuclear power production industry. This accident, which occurred at 1:24 a.m. on April 26, 1986, resulted from an almost incredible series of operational errors associated, ironically, with an attempt to enhance the capability of the reactor to safely accommodate station blackout accidents (i.e., accidents arising from a loss of station electrical power). Disruption of the core, due to a prompt criticality excursion, resulted in the destruction of the core vault and reactor building and the sudden dispersal of about 3% of the fuel from the core region into the environment. Lesser but significant releases of radioactivity continued through May 6, 1986, before attempts to certain the radioactivity and cool the remnants of the core were successful. The amount and composition of material released in the course of the accident remain somewhat uncertain, and inconsistencies in the release estimates are evident. The Soviet estimates, in addition to the dispersal of about 3% of the fuel, include complete release of the noble gas core inventory, 20% of the fission product iodine inventory, 15% of the tellurium inventory, and 10 to 13% of the fission product cesium inventory. The iodine and cesium release estimates are not consistent with the noble gas values, and are as much as a factor of two less than some estimates made by experts outside the Soviet Union.

  10. Accident Tolerant Fuel Analysis

    SciTech Connect

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  11. Accident tolerant fuel analysis

    SciTech Connect

    Smith, Curtis; Chichester, Heather; Johns, Jesse; Teague, Melissa; Tonks, Michael Idaho National Laboratory; Youngblood, Robert

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  12. Severe accident natural circulation studies at the INEL

    SciTech Connect

    Bayless, P.D.; Brownson, D.A.; Dobbe, C.A.; Jones, K.R.; O`Brien, J.E.; Pafford, D.J.; Schlenker, L.D.; Tung, V.X.

    1995-02-01

    Severe accident natural circulation flows have been investigated at the Idaho National Engineering Laboratory to better understand these flows and their potential impacts on the progression of a pressurized water reactor severe accident. Parameters affecting natural circulation in the reactor vessel and hot legs were identified and ranked based on their perceived importance. Reviews of the scaling of the 1/7-scale experiments performed by Westinghouse were undertaken. RELAP5/MOD3 calculations of two of the experiments showed generally good agreement between the calculated and observed behavior. Analyses of hydrogen behavior in the reactor vessel showed that hydrogen stratification is not likely to occur, and that an initially stratified layer of hydrogen would quickly mix with a recirculating steam flow. An analysis of the upper plenum behavior in the Three Mile Island, Unit 2 reactor concluded that vapor temperatures could have been significantly higher than the temperatures seen by the control rod drive lead screws, supporting the premise that a strong natural circulation flow was likely present during the accident. SCDAP/RELAP5 calculations of a commercial pressurized water reactor severe accident without operator actions showed that the natural circulation flows enhance the likelihood of ex-vessel piping failures long before failure of the reactor vessel lower head.

  13. Accident mortality among children

    PubMed Central

    Swaroop, S.; Albrecht, R. M.; Grab, B.

    1956-01-01

    The authors present statistics on mortality from accidents, with special reference to those relating to the age-group 1-19 years. For a number of countries figures are given for the proportional mortality from accidents (the number of accident deaths expressed as a percentage of the number of deaths from all causes) and for the specific death-rates, per 100 000 population, from all causes of death, from selected causes, from all causes of accidents, and from various types of accident. From these figures it appears that, in most countries, accidents are becoming relatively increasingly prominent as a cause of death in childhood, primarily because of the conquest of other causes of death—such as infectious and parasitic diseases, which formerly took a heavy toll of children and adolescents—but also to some extent because the death-rate from motor-vehicle accidents is rising and cancelling out the reduction in the rate for other causes of accidental death. In the authors' opinion, further epidemiological investigations into accident causation are required for the purpose of devising quicker and more effective methods of accident prevention. PMID:13383361

  14. The role of NUREG-1150 in accident management

    SciTech Connect

    Camp, A.L.; Cramond, W.R.; Sype, T.T.

    1988-01-01

    NUREG-1150 is being prepared by the NRC and its contractors to estimate the risk from five commercial light water reactors: Surry, Sequoyah, Peach Bottom, Grand Gulf, and Zion. Level 3 probabilistic risk assessments (PRAs) are being prepared for each of these plants. These PRAs provide a framework for evaluating accident management alternatives from a risk standpoint. This paper describes the accident management benefits that NUREG-1150 is providing.

  15. The effects of springtime mid-latitude storms on trace gas composition determined from the MACC reanalysis

    NASA Astrophysics Data System (ADS)

    Knowland, K. E.; Doherty, R. M.; Hodges, K. I.

    2014-10-01

    The relationship between springtime air pollution transport of ozone (O3) and carbon monoxide (CO) and mid-latitude cyclones is explored for the first time using the Monitoring Atmospheric Composition and Climate (MACC) reanalysis for the period 2003-2012. In this study, the most intense spring storms (95th percentile) are selected for two regions, the North Pacific (NP) and the North Atlantic (NA). These storms (~60 storms over each region) often track over the major emission sources of East Asia and eastern North America. By compositing the storms, the distributions of O3 and CO within a "typical" intense storm are examined. We compare the storm-centered composite to background composites of "average conditions" created by sampling the reanalysis data of the previous year to the storm locations. Mid-latitude storms are found to redistribute concentrations of O3 and CO horizontally and vertically throughout the storm. This is clearly shown to occur through two main mechanisms: (1) vertical lifting of CO-rich and O3-poor air isentropically from near the surface to the mid- to upper-troposphere in the region of the warm conveyor belt; and (2) descent of O3-rich and CO-poor air isentropically in the vicinity of the dry intrusion, from the stratosphere toward the mid-troposphere. This can be seen in the composite storm's life cycle as the storm intensifies, with area-averaged O3 (CO) increasing (decreasing) between 200 and 500 hPa. At the time of maximum intensity, area-averaged O3 around the storm center at 300 hPa is enhanced by 50 and 36% for the NP and NA regions respectively, compared to the background, and by 11 and 7.6% at 500 hPa. In contrast, area-averaged CO at 300 hPa decreases by 12% for NP and 5.5% for NA, and at 500 hPa area-averaged CO decreases by 2.4% for NP while there is little change over the NA region at 500 hPa. From the mid-troposphere, O3-rich air is clearly seen to be transported toward the surface but the downward transport of CO-poor air is

  16. The effects of springtime mid-latitude storms on trace gas composition determined from the MACC reanalysis

    NASA Astrophysics Data System (ADS)

    Knowland, K. E.; Doherty, R. M.; Hodges, K. I.

    2015-03-01

    The relationship between springtime air pollution transport of ozone (O3) and carbon monoxide (CO) and mid-latitude cyclones is explored for the first time using the Monitoring Atmospheric Composition and Climate (MACC) reanalysis for the period 2003-2012. In this study, the most intense spring storms (95th percentile) are selected for two regions, the North Pacific (NP) and the North Atlantic (NA). These storms (∼60 storms over each region) often track over the major emission sources of East Asia and eastern North America. By compositing the storms, the distributions of O3 and CO within a "typical" intense storm are examined. We compare the storm-centered composite to background composites of "average conditions" created by sampling the reanalysis data of the previous year to the storm locations. Mid-latitude storms are found to redistribute concentrations of O3 and CO horizontally and vertically throughout the storm. This is clearly shown to occur through two main mechanisms: (1) vertical lifting of CO-rich and O3-poor air isentropically, from near the surface to the mid- to upper-troposphere in the region of the warm conveyor belt; and (2) descent of O3-rich and CO-poor air isentropically in the vicinity of the dry intrusion, from the stratosphere toward the mid-troposphere. This can be seen in the composite storm's life cycle as the storm intensifies, with area-averaged O3 (CO) increasing (decreasing) between 200 and 500 hPa. The influence of the storm dynamics compared to the background environment on the composition within an area around the storm center at the time of maximum intensity is as follows. Area-averaged O3 at 300 hPa is enhanced by 50 and 36% and by 11 and 7.6% at 500 hPa for the NP and NA regions, respectively. In contrast, area-averaged CO at 300 hPa decreases by 12% for NP and 5.5% for NA, and area-averaged CO at 500 hPa decreases by 2.4% for NP while there is little change over the NA region. From the mid-troposphere, O3-rich air is

  17. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  18. The polonium-210 problem in thermonuclear reactor

    SciTech Connect

    Shchipakhin, O.L.; Borisov, N.B.; Churkin, S.L.

    1993-12-31

    Polonium 210 forms in the lithium-lead eutectic blanket of a thermonuclear reactor. On the basis of obtained experimental data some estimates have been calculated on the ITER blanket accident consequences. The LOCA type accident represents the failure of eutectic circuit in the process of transfusion of liquid eutectic from blanket to the tritium reprocessing plant.

  19. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes.

  20. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. Fukushima Daiichi Accident and Its Radiological Impact on the Environment

    NASA Astrophysics Data System (ADS)

    Bevelacqua, J. J.

    2012-09-01

    The Fukushima Daiichi nuclear accident is a topic of current media and public interest. It provides a means to motivate students to understand the fission process and the barriers that have been designed to prevent the release of fission products to the environment following a major nuclear power reactor accident. The Fukushima Daiichi accident further encourages a discussion of the effect of fission products upon the environment, including the resulting contamination of air, water, soil, animals, fish, milk, and crops. Accident-generated radiation levels that caused the evacuation of people 20-30 km from the facility further serve to foster student interest and desire to understand the science associated with the Fukushima Daiichi accident.

  2. Airline accident response.

    PubMed

    Bettes, Thomas

    2002-01-01

    This article outlines government regulations affecting accident response and offers guidelines for airline contingency plans in the face of major air disasters, such as those encountered on September 11, 2001. The author also touches upon the role of the corporate medical department in accident investigation and victim identification. PMID:11872433

  3. Civil aircraft accident investigation.

    PubMed

    Haines, Daniel

    2013-01-01

    This talk reviews some historic aircraft accidents and some more recent. It reflects on the division of accident causes, considering mechanical failures and aircrew failures, and on aircrew training. Investigation results may lead to improved aircraft design, and to appropriate crew training. PMID:24057309

  4. Anatomy of an Accident.

    ERIC Educational Resources Information Center

    Mobley, Michael

    1984-01-01

    The findings of industrial safety engineers in the areas of accident causation and prevention are wholly applicable to adventure programs. Adventure education instructors can use safety engineering concepts to assess the risk in a particular activity, understand factors that cause accidents, and intervene to minimize injuries and damages if…

  5. Radiation accidents and nuclear energy: medical consequences and therapy.

    PubMed

    Champlin, R E; Kastenberg, W E; Gale, R P

    1988-11-01

    After the accidents at Chernobyl, the Soviet Union, and in Goiania, Brazil, there is increasing concern about the medical risks from radiation accidents. This overview summarizes the principles of nuclear energy, the biologic effects of accidental radiation exposure, the emergency response to nuclear accidents, and approaches to treating radiation injuries. Also discussed are the related issues of reactor safety, the disposal of radioactive waste, and the proliferation of nuclear weapons. With the increasing use of radioactive materials for power, weapons, and medical diagnostics, the medical community needs to understand the health consequences of radiation exposure. PMID:3056171

  6. Particulate organic acids and overall water-soluble aerosol composition measurements from the 2006 Gulf of Mexico Atmospheric Composition and Climate Study (GoMACCS)

    NASA Astrophysics Data System (ADS)

    Sorooshian, Armin; Ng, Nga L.; Chan, Arthur W. H.; Feingold, Graham; Flagan, Richard C.; Seinfeld, John H.

    2007-07-01

    The Center for Interdisciplinary Remotely-Piloted Aircraft Studies (CIRPAS) Twin Otter participated in the Gulf of Mexico Atmospheric Composition and Climate Study (GoMACCS) mission during August-September 2006. A particle-into-liquid sampler (PILS) coupled to ion chromatography was used to characterize the water-soluble ion composition of aerosol and cloud droplet residual particles (976 5-min PM1.0 samples in total). Sulfate and ammonium dominated the water-soluble mass (NH4+ + SO42- = 84 ± 14%), while organic acids contributed 3.4 ± 3.7%. The average NH4+:SO42- molar ratio was 1.77 ± 0.85. Particulate concentrations of organic acids increased with decreasing carbon number from C9 to C2. Organic acids were most abundant above cloud, presumably as a result of aqueous phase chemistry in cloud droplets, followed by subsequent droplet evaporation above cloud tops; the main product of this chemistry was oxalic acid. The evolution of organic acids with increasing altitude in cloud provides evidence for the multistep nature of oxalic acid production; predictions from a cloud parcel model are consistent with the observed oxalate:glyoxylate ratio as a function of altitude in GoMACCS cumuli. Suppressed organic acid formation was observed in clouds with relatively acidic droplets, as determined by high particulate nitrate concentrations (presumably high HNO3 levels too) and lower liquid water content, as compared to other cloud fields probed. In the Houston Ship Channel region, an area with significant volatile organic compound emissions, oxalate, acetate, formate, benzoate, and pyruvate, in decreasing order, were the most abundant organic acids. Photo-oxidation of m-xylene in laboratory chamber experiments leads to a particulate organic acid product distribution consistent with the Ship Channel area observations.

  7. Evaluation of Emissions and Photochemical Processing Within Air Quality Model Forecasts During the 2006 TexAQS/GoMACCS Field Study

    NASA Astrophysics Data System (ADS)

    McKeen, S. A.; Grell, G.; Peckham, S.; McQueen, J.; Lee, P.; McHenry, J.; Gong, W.; Bouchet, V.; Tang, Y.; Carmichael, G.; Wilczak, J.; Djalalova, I.; None, N.

    2007-12-01

    Several air-quality models provided real-time forecasts of ozone and PM2.5 aerosols during the TexAQS/GoMACCS field campaign. These forecast models include two versions of the NOAA/ESRL/GSD WRF/Chem model, a developmental version of the NWS/NCEP CMAQ/WRF model, the Canadian Meteorological Services CHRONOS and AURAMS models, the MM5 based MAQSIP model from Baron Advanced Meteorological Services Inc., and the University of Iowa STEM model. Statistical evaluations of each model with the U.S. EPA AIRNow ozone and PM2.5 network over Eastern Texas during the summer of 2006 point to persistent model biases in surface predictions of these two criteria pollutants. Uncertainties in emission inventories and photochemical mechanisms are likely sources of forecast error within each model. Detailed observations of dozens of gas-phase ozone precursors and aerosol components collected on board the NOAA-WP3 aircraft during TexAQS/GoMACCS are used to compare model and observed concentrations. Aircraft flight tracks were designed to characterize up-wind conditions and the evolving composition of urban plumes down-wind of Houston and Dallas, TX within the planetary boundary layer. The aircraft data for 10 flights during September of 2006 are used in three diagnostic evaluations of the various models: characterizing the background composition up-wind of the two urban areas, evaluating the photochemical processing leading to ozone and PM2.5 formation various distances down-wind of the urban sources, and using ratios of above-background concentrations to infer and compare emission ratios of key ozone and PM2.5 precursors.

  8. Savannah River Site K-Reactor Probabilistic Safety Assessment

    SciTech Connect

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O`Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety.

  9. N Reactor operational safety summary

    SciTech Connect

    Franz, G.R.; Quapp, W.J.; Ogden, D.M.

    1988-08-01

    This report is a safety summary of the N Reactor. Beginning with its conceptual design in the mid-1950`s, and throughout its 23 years of operation, continuous efforts have been made to ensure safe N Reactor operation and protection of the public health and safety. The N Reactor Updated Safety Analysis Report, completed in 1978(UNC1978), and its subsequent amendments document the safety bases of N Reactor. Following the April 1986 Chernobyl accident in the Soviet Union, a major effort to confirm N Reactor safety and further increase its safety margin was initiated. This effort, called the Safety Enhancement Program, reassessed the N Reactor using the latest accepted analysis techniques and commercial light-water reactor guidelines, where applicable. 122 refs., 38 figs., 10 tabs.

  10. Persistence of airline accidents.

    PubMed

    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. PMID:20618386

  11. Sleep related vehicle accidents.

    PubMed Central

    Horne, J. A.; Reyner, L. A.

    1995-01-01

    OBJECTIVES--To assess the incidence, time of day, and driver morbidity associated with vehicle accidents where the most likely cause was the driver falling asleep at the wheel. DESIGN--Two surveys were undertaken, in southwest England and the midlands, by using police databases or on the spot interviews. SUBJECTS--Drivers involved in 679 sleep related vehicle accidents. RESULTS--Of all vehicle accidents to which the police were summoned, sleep related vehicle accidents comprised 16% on major roads in southwest England, and over 20% on midland motorways. During the 24 hour period there were three major peaks: at around 0200, 0600, and 1600. About half these drivers were men under 30 years; few such accidents involved women. CONCLUSIONS--Sleep related vehicle accidents are largely dependent on the time of day and account for a considerable proportion of vehicle accidents, especially those on motorways and other monotonous roads. As there are no norms for the United Kingdom on road use by age and sex for time of day with which to compare these data, we cannot determine what the hourly exposure v risk factors are for these subgroups. The findings are in close agreement with those from other countries. PMID:7888930

  12. Safety Is No Accident.

    ERIC Educational Resources Information Center

    Christiansen, Monty L.

    1985-01-01

    Liability suits involving accidents in park and recreation areas are expensive and intangible costs are incalculable. Risk management practices related to park planning, personnel, and administrative practices are discussed. (MT)

  13. Accident resistant transport container

    DOEpatents

    Andersen, John A.; Cole, James K.

    1980-01-01

    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.

  14. Accident resistant transport container

    DOEpatents

    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.

  15. XENON-133 IN CALIFORNIA, NEVADA, AND UTAH FROM THE CHERNOBYL ACCIDENT (JOURNAL VERSION)

    EPA Science Inventory

    The accident at the Chernobyl nuclear reactor in the USSR introduced numerous radioactive nuclides into the atmosphere, including the noble gas xenon-133. EPA's Environmental Monitoring Systems Laboratory, Las Vegas, NV, detected xenon-133 from the Chernobyl accident in air sampl...

  16. FATAL ACCIDENT REPORTING SYSTEM (FARS)

    EPA Science Inventory

    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.

  17. BWR reactor vessel bottom head failure modes

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. The effect of the BWR procedural and structural differences upon the progression of a severe accident sequence during the period preceding movement of core debris into the reactor vessel lower plenum has been discussed previously. It is the purpose of this paper to briefly address the events occurring after debris relocation past the core plate and to describe the subsequent expected modes of bottom head pressure boundary failure. As an example, the calculated timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is also presented. 14 refs., 4 figs., 1 tab.

  18. Severe Accident Test Station Activity Report

    SciTech Connect

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  19. ANS severe accident program overview & planning document

    SciTech Connect

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  20. Children's reactions to the threat of nuclear plant accidents.

    PubMed

    Schwebel, M; Schwebel, B

    1981-04-01

    In the wake of Three Mile Island nuclear plant accident, questionnaire and interview responses of children in elementary and secondary schools revealed their perceptions of the dangers entailed in the continued use of nuclear reactors. Results are compared with a parallel study conducted close to 20 years ago, and implications for mental health are examined. PMID:7223871

  1. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    SciTech Connect

    Robb, Kevin R.

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, Fe

  2. Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

    SciTech Connect

    Gamble, K. A.; Hales, J. D.; Yu, J.; Zhang, Y.; Bai, X.; Andersson, D.; Patra, A.; Wen, W.; Tome, C.; Baskes, M.; Martinez, E.; Stanek, C. R.; Miao, Y.; Ye, B.; Hofman, G. L.; Yacout, A. M.; Liu, W.

    2015-09-01

    U3Si2 and iron-chromium-aluminum (Fe-Cr-Al) alloys are two of many proposed accident-tolerant fuel concepts for the fuel and cladding, respectively. The behavior of these materials under normal operating and accident reactor conditions is not well known. As part of the Department of Energy’s Accident Tolerant Fuel High Impact Problem program significant work has been conducted to investigate the U3Si2 and FeCrAl behavior under reactor conditions. This report presents the multiscale and multiphysics effort completed in fiscal year 2015. The report is split into four major categories including Density Functional Theory Developments, Molecular Dynamics Developments, Mesoscale Developments, and Engineering Scale Developments. The work shown here is a compilation of a collaborative effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory and Anatech Corp.

  3. [Psychogenesis of accidents].

    PubMed

    Giannattasio, E; Nencini, R; Nicolosi, N

    1988-01-01

    After having carried out a historical review of industrial psychology with specific attention to the evolution of the concept of causality in accidents, the Authors formulate their work hypothesis from that research which take into highest consideration the executives' attitudes in the genesis of the accidents. As dogmatism appears to be one of the most negative of executives' attitudes, the Authors administered Rockeach's Scale to 130 intermediate executives from 6 industries in Latium and observed the frequency index for accidents and the morbidity index (absenteeism) of the 2149 workhand. The Authors assumed that to high degree of dogmatism on the executives' side should correspond o a higher level of accidents and absenteeism among the staff. The data processing revealed that, due to the type of machinery employed, three of the industries examined should be considered as High Risk Industrie (HRI), while the remaining three could be considered as Low Risk Industries (LRI): in fact, due to the different working conditions, a significant lower number of accidents occurred in last the three. A statistically significant correlation between the executives' dogmatism and the number of accidents among their workhand in the HRI has been noticed, while this has not been observed in the LRI. This confirms, as had already been pointed out by Gemelli in 1944, that some "objective conditions" are requested so that the accident may actually take place. On the other hand the morbidity index has not shown any difference related to the different kind of industries (HRI, LRI): in both cases statistically significant correlations were obtained between the executives' dogmatism and the staff's absenteeism. absenteeism.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:3154344

  4. Characterization of a nuclear accident dosimeter

    SciTech Connect

    Burrows, R.A.

    1995-12-01

    The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12--16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia National Laboratories (SNL) personal nuclear accident dosimeter (PNAD). Two separate critical assemblies, SHEBA and Godiva, were used to generate seven separate neutron spectra for use in dose comparisons. SNL`s PNAD measured absorbed doses that were within +16 to +26% of the reference doses. In addition, a preliminary investigation was undertaken to determine the feasibility of using the data obtained from an irradiated PNAD to correct for body orientation. This portion of the experiment was performed with a TRIGA reactor at the Nuclear Science Center at Texas A and M University.

  5. Scientific aspects of the Tohoku earthquake and Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Koketsu, Kazuki

    2016-04-01

    We investigated the 2011 Tohoku earthquake, the accident of the Fukushima Daiichi nuclear power plant, and assessments conducted beforehand for earthquake and tsunami potential in the Pacific offshore region of the Tohoku District. The results of our investigation show that all the assessments failed to foresee the earthquake and its related tsunami, which was the main cause of the accident. Therefore, the disaster caused by the earthquake, and the accident were scientifically unforeseeable at the time. However, for a zone neighboring the reactors, a 2008 assessment showed tsunamis higher than the plant height. As a lesson learned from the accident, companies operating nuclear power plants should be prepared using even such assessment results for neighboring zones.

  6. [The Fukushima nuclear accident: consequences for Japan and for us].

    PubMed

    Grosche, B

    2013-04-01

    The Fukushima accident was the consequence of a preceding 2-fold natural catastrophe: the earth quake of 11 March 2011 and the subsequent tsunami. Due to favourable winds and to evacuation measures the radiation exposure to the general population in Japan as a whole and with some exceptions in the region outside the evacuation zone, too, was low. In this article the attempt is made to give an estimate of health consequences to the public. This is based upon WHO's dose estimates, knowledge of the consequences of the Chernobyl accident, of the atmospheric nuclear bomb testing in Kazakhstan and on the risk of childhood leukaemia after low dose radiation exposure. For Germany, there was no radiation threat due to the accident. Nonetheless, the events in Japan made clear that the rules and standards that were developed for the case of a reactor accident need to be revised. PMID:23576143

  7. Russian-American venture designs new reactor

    SciTech Connect

    Newman, P.

    1994-01-03

    Russian and American nuclear energy experts have completed a joint design study of a small, low-cost and demonstrably accident-proof reactor that they say could revolutionize the way conventional reactors are designed, marketed and operated. The joint design is helium-cooled and graphite-moderated and has a power density of 3 MWt/cubic meter, which is significantly less than the standard American reactor. A prototype of this design should be operating in Chelyabinsk by June 1996.

  8. Injuries are not accidents

    PubMed Central

    Gutiérrez, María Isabel

    2014-01-01

    Injuries are the result of an acute exposure to exhort of energy or a consequence of a deficiency in a vital element that exceeds physiological thresholds resulting threatens life. They are classified as intentional or unintentional. Injuries are considered a global health issue because they cause more than 5 million deaths per year worldwide and they are an important contributor to the burden of disease, especially affecting people of low socioeconomic status in low- and middle-income countries. A common misconception exists where injuries are thought to be the same as accidents; however, accidents are largely used as chance events, without taken in consideration that all these are preventable. This review discusses injuries and accidents in the context of road traffic and emphasizes injuries as preventable events. An understanding of the essence of injuries enables the standardization of terminology in public use and facilitates the development of a culture of prevention among all of us. PMID:25386040

  9. Accidents in Childhood

    PubMed Central

    Keddy, J. Arthur

    1964-01-01

    The causes of injury to 17,141 children brought to the emergency department of a large pediatric hospital in one year were studied. The leading causes of injury were: falls, 5682; cuts or piercings, 1902; poisonings, 1597; and transportation accidents, 1368. Included in these are 587 falls on or down stairs, 401 cuts due to glass, 630 poisonings from household or workshop substances, 510 poisonings from salicylate tablets, and 449 accidents involving bicycles or tricycles. Other findings included 333 injuries to fingers or hands in doors, usually car doors; 122 instances of pulled arms; 384 ingestions and 53 inhalations of foreign bodies; 60 alleged sexual assaults, 58 chemical burns, 127 wringer injuries, and four attempted suicides. A rewarding opportunity in accident prevention exists for hospitals that undertake to compile and distribute pertinent source data. PMID:14201260

  10. NRC policy on future reactor designs

    SciTech Connect

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  11. The Fukushima radiation accident: consequences for radiation accident medical management.

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

    The March 2011 radiation accident in Fukushima, Japan, is a textbook example of a radiation accident of global significance. In view of the global dimensions of the accident, it is important to consider the lessons learned. In this context, emphasis must be placed on consequences for planning appropriate medical management for radiation accidents including, for example, estimates of necessary human and material resources. The specific characteristics of the radiation accident in Fukushima are thematically divided into five groups: the exceptional environmental influences on the Fukushima radiation accident, particular circumstances of the accident, differences in risk perception, changed psychosocial factors in the age of the Internet and globalization, and the ignorance of the effects of ionizing radiation both among the general public and health care professionals. Conclusions like the need for reviewing international communication, interfacing, and interface definitions will be drawn from the Fukushima radiation accident. PMID:22951483

  12. Small marine craft emission factors observed from the NOAA R/V Ronald H. Brown during TexAQS/GoMACCS 2006

    NASA Astrophysics Data System (ADS)

    Lerner, B. M.; Lack, D.; Murphy, P. C.; Williams, E. J.

    2007-12-01

    During the TexAQS/GoMACCS 2006 field campaign, the NOAA R/V Ronald H. Brown often encountered small marine recreational craft and small fishing vessels while sailing close to the Texas coast, especially in Galveston Bay. Measurement of a suite of trace gases at high time resolution (1 Hz) allowed us to calculate emission factors (EFs), relative to carbon dioxide, for nitrogen oxides (NOx), sulfur dioxide (SO2) and carbon monoxide (CO) for distinct exhaust plumes from these sources. Photoacoustic aerosol absorption spectroscopy (PAS) measurements made concurrently allowed for the first quantification of mass EFs for light-absorbing particles from fishing craft. As previously observed along the New England coast, gasoline-powered recreational vessels showed significantly higher NOx/CO2 and lower CO/CO2 EFs than current emissions inventories predict, although in agreement with the most recent published literature of laboratory studies. These findings imply lower volatile organic compound emissions from these vessels, although this was not directly measured.

  13. Physics in Accident Investigations.

    ERIC Educational Resources Information Center

    Brake, Mary L.

    1981-01-01

    Describes physics formulas which can be used by law enforcement officials to determine the possible velocity of vehicles involved in traffic accidents. These include, among others, the slide to stop-level road, slide to stop-sloping roadway, and slide to stop-two different surfaces formulas. (JN)

  14. The Chernobyl accident ten years later

    SciTech Connect

    Squires, D.J.

    1996-12-31

    On April 26, 1986 at 1:23 AM a fire and explosion occurred at the fourth unit of the Chernobyl Nuclear Power Plant Complex, located in the Ukraine, that resulted in the destruction of the reactor core and most of the building in which it was housed. Several environmental impacts resulting from the accident will be discussed in this paper, which will include the effects on plant and wild life, radioactive waste generated and stored or disposed of, effects of evacuations relating to residents within the subsequently established 10km and 30km control zones, impacts of the emergency containment structure (sarcophagus), and potential effects on world opinion and future development of nuclear power. As an immediate result of the fire, 31 people died (2 from the fire & smoke, and 29 from excessive radiation); 237 cases of acute radiation sickness occurred; the total fatalities based upon induced chronic diseases as a result of the accident is unknown: more than 100,000 people were evacuated from within the subsequently established 30 km control zone; in excess of 50 million curies of radionuclides that included finely dispersed nuclear fuel, fragments of graphite, concrete and other building materials were released from the reactor into the environment; an estimated one million cubic meters of radioactive waste were generated (LLW, ILW, HLW); more than 5000 tons of materials (sand, boron, dolomite, cement, and lead) were used to put the fire out by helicopter; shutdown of the adjacent power plants were performed; and other environmental impacts occurred. The Chernobyl Nuclear Power Plant Unit No 4 is an RBMK-1000. It initiated operations in 1983, it was a 1000 MWe with a power output of 3200 MW(th), the reactor core contained 190 MT of fuel, with 1659 assemblies (plus 211 control rods), the average burnup rate was 10.3 MWd/kg, and the reactor operated on a continuous basis with maintenance and fuel reload performed during operations.

  15. PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise

    SciTech Connect

    Hill, Robin L.; Conrady, Matthew M.

    2011-10-28

    This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

  16. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K.

    2012-07-01

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  17. Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

    SciTech Connect

    Wichner, R.P.; Weber, C.F.; Hodge, S.A.; Beahm, E.C.; Wright, A.L.

    1984-01-01

    This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor.

  18. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    SciTech Connect

    Robb, Kevin R

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  19. PARET-ANL. Program for the Analysis of Reactor Transients

    SciTech Connect

    Woodruff, W.L.

    1984-11-01

    PERET-ANL is designed for use in predicting the course and consequences of nondestructive reactivity accidents in research and test reactor cores. It can be used for both steady-state and transient analysis.

  20. Study Gives Good Odds on Nuclear Reactor Safety

    ERIC Educational Resources Information Center

    Russell, Cristine

    1974-01-01

    Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

  1. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  2. Applying STAMP in Accident Analysis

    NASA Technical Reports Server (NTRS)

    Leveson, Nancy; Daouk, Mirna; Dulac, Nicolas; Marais, Karen

    2003-01-01

    Accident models play a critical role in accident investigation and analysis. Most traditional models are based on an underlying chain of events. These models, however, have serious limitations when used for complex, socio-technical systems. Previously, Leveson proposed a new accident model (STAMP) based on system theory. In STAMP, the basic concept is not an event but a constraint. This paper shows how STAMP can be applied to accident analysis using three different views or models of the accident process and proposes a notation for describing this process.

  3. [Equestrian accidents in children].

    PubMed

    Giebel, G; Braun, K; Mittelmeier, W

    1993-11-01

    In a retrospective study we reviewed 262 horse riding related injuries in children younger than 16 which were treated between 1975 and 1989 at the Section of Traumatology in the Department of Surgery, University Hospital Homburg/Saar. In 155 of these accidents, detailed information was gained via a questionnaire. The typical patient profile was that of young female equestrians with little experience and little weekly riding practice, without practicing falling-exercises and warming up often using different horses. At the time of the accident only 59% were wearing a head protection. Most accidents happened in the summer months in the afternoon during leisure riding on a large familiar horse in the riding hall. Apart from the typical accidents like falling of the horse (64.9%) and falling with the horse (5.7%) accidents in handling the horse were of special significance: Kick by horse's hoof (11.8%), being stepped by horse (3.8%), horsebite (7.3%) and injuries of horse's bridle had their own pattern of injuries. Injuries of the distal parts of the upper extremity are preeminent in falling of the horse, whilst in falling with the horse head injuries and shoulder injuries are preeminent. Remarkably often injuries of kick by horse's hoof were causing sometimes even dangerous head injuries (41.6%). Overall in horse riding related injuries in childhood superficial soft tissue injuries (48.6%) and fractures (30.6%) were predominant. Fractures of the clavicle which are well known as a riding injury proved to be typical for a fall with the horse, whilst a fractured vertebra was only seen once amongst the 262 children treated. The severity of the injuries was lower than expected: In 85.1% of all the injuries only one body region was injured, 90.1% could be assigned to an injury severity score (ISS) of 1-3. Ponyriders had less severe injuries than riders of large horses. One fatal accident happened in handling a horse, in these situations preventive measures are often

  4. 49 CFR 835.11 - Obtaining Board accident reports, factual accident reports, and supporting information.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Obtaining Board accident reports, factual accident... Board accident reports, factual accident reports, and supporting information. It is the responsibility... obtain Board accident reports, factual accident reports, and accompanying accident docket files....

  5. 49 CFR 835.11 - Obtaining Board accident reports, factual accident reports, and supporting information.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Obtaining Board accident reports, factual accident... Board accident reports, factual accident reports, and supporting information. It is the responsibility... obtain Board accident reports, factual accident reports, and accompanying accident docket files....

  6. Accident Flying Squad

    PubMed Central

    Snook, Roger

    1972-01-01

    This paper describes the organization, evaluation, and costing of an independently financed and operated accident flying squad. 132 accidents involving 302 casualties were attended, six deaths were prevented, medical treatment contributed to the survival of a further four, and the condition or comfort of many other casualties was improved. The calls in which survival was influenced were evenly distributed throughout the three-and-a-half-year survey and seven of the 10 so aided were over 16 and under 30 years of age, all 10 being in the working age group. The time taken to provide the service was not excessive and the expense when compared with the overall saving was very small. The scheme was seen to be equally suitable for basing on hospital or general practice or both, and working as an integrated team with the ambulance service. The use of specialized transport was found to be unnecessary. Other benefits of the scheme included use of the experience of attending accidents to ensure relevant and realistic training for emergency service personnel, and an appreciation of the effect of ambulance design on the patient. ImagesFIG. 1FIG. 4 PMID:5069642

  7. Perspectives on reactor safety. Revision 1

    SciTech Connect

    Haskin, F.E.; Camp, A.L.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  8. Safety aspect concerning radiolytic gas generation in reactors.

    PubMed

    Ramshesh, V

    2001-01-01

    In water cooled and water moderated reactors (H2O in boiling water reactors/pressurised water reactors, D2O in pressurised heavy water reactors) during normal operation, radiolysis is a source of production of hydrogen/deuterium and oxygen. During the progress of a nuclear accident, while there are other important sources of hydrogen/deuterium, the oxygen availability can occur only through radiolysis or direct contact with air. In air saturated with water vapour at room temperature and pressure when H2/D2 concentration exceeds 4 vol % (a conservative estimate), a combustible mixture with oxygen can be formed. It is proposed to examine the basic principles of water radiolysis as far as they pertain to generation of H2/D2 and O2 and try to apply these concepts to reactors both under operating conditions and in accident situations. It is concluded that the possibility of an accident taking place through radiolysis is highly unlikely. PMID:11382138

  9. Nuclear reactor vessel fuel thermal insulating barrier

    SciTech Connect

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  10. Space Reactor Launch Safety-An Acceptably Low Risk

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham; Wright, Steven

    2008-01-01

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10-5 latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  11. Space Reactor Launch Safety--An Acceptably Low Risk

    SciTech Connect

    Weitzberg, Abraham; Wright, Steven

    2008-01-21

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10{sup -5} latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  12. Safety evaluation of MHTGR licensing basis accident scenarios

    SciTech Connect

    Kroeger, P.G.

    1989-04-01

    The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential accident scenarios that might lead to fuel failures due to excessive core temperatures and/or to vessel damage, due to excessive vessel temperatures. The design basis accident scenario leading to the highest vessel temperatures is the depressurized core heatup scenario without any forced cooling and with decay heat rejection to the passive Reactor Cavity Cooling System (RCCS). This scenario was evaluated, including numerous parametric variations of input parameters, like material properties and decay heat. It was found that significant safety margins exist, but that high confidence levels in the core effective thermal conductivity, the reactor vessel and RCCS thermal emissivities and the decay heat function are required to maintain this safety margin. Severe accident extensions of this depressurized core heatup scenario included the cases of complete RCCS failure, cases of massive air ingress, core heatup without scram and cases of degraded RCCS performance due to absorbing gases in the reactor cavity. Except for no-scram scenarios extending beyond 100 hr, the fuel never reached the limiting temperature of 1600/degree/C, below which measurable fuel failures are not expected. In some of the scenarios, excessive vessel and concrete temperatures could lead to investment losses but are not expected to lead to any source term beyond that from the circulating inventory. 19 refs., 56 figs., 11 tabs.

  13. Savannah River Site production reactor technical specifications. K Production Reactor

    SciTech Connect

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  14. Radionuclide release calculations for selected severe accident scenarios

    SciTech Connect

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. )

    1990-08-01

    This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs.

  15. Possible consequences of severe accidents at the Lubiatowo site, Poland

    NASA Astrophysics Data System (ADS)

    Seibert, Petra; Philipp, Anne; Hofman, Radek; Gufler, Klaus; Sholly, Steven

    2014-05-01

    The construction of a nuclear power plant is under consideration in Poland. One of the sites under discussion is near Lubiatowo, located on the cost of the Baltic Sea northwest of Gdansk. An assessment of possible environmental consequences is carried out for 88 real meteorological cases with the Lagrangian particle dispersion model FLEXPART. Based on literature research, three reactor designs (ABWR, EPR, AP 1000) were identified as being under discussion in Poland. For each of the designs, a set of accident scenarios was evaluated and two source terms per reactor design were selected for analysis. One of the selected source terms was a relatively large release while the second one was a severe accident with an intact containment. Considered endpoints of the calculations are ground contamination with Cs-137 and time-integrated concentrations of I-131 in air as well as committed doses. They are evaluated on a grid of ca. 3 km mesh size covering eastern Central Europe.

  16. Simulation of Accident Sequences Including Emergency Operating Procedures

    SciTech Connect

    Queral, Cesar; Exposito, Antonio; Hortal, Javier

    2004-07-01

    Operator actions play an important role in accident sequences. However, design analysis (Safety Analysis Report, SAR) seldom includes consideration of operator actions, although they are required by compulsory Emergency Operating Procedures (EOP) to perform some checks and actions from the very beginning of the accident. The basic aim of the project is to develop a procedure validation system which consists of the combination of three elements: a plant transient simulation code TRETA (a C based modular program) developed by the CSN, a computerized procedure system COPMA-III (Java technology based program) developed by the OECD-Halden Reactor Project and adapted for simulation with the contribution of our group and a software interface that provides the communication between COPMA-III and TRETA. The new combined system is going to be applied in a pilot study in order to analyze sequences initiated by secondary side breaks in a Pressurized Water Reactors (PWR) plant. (authors)

  17. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  18. Test Data for USEPR Severe Accident Code Validation

    SciTech Connect

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  19. Consequences and countermeasures in a nuclear power accident: Chernobyl experience.

    PubMed

    Kirichenko, Vladimir A; Kirichenko, Alexander V; Werts, Day E

    2012-09-01

    Despite the tragic accidents in Fukushima and Chernobyl, the nuclear power industry will continue to contribute to the production of electric energy worldwide until there are efficient and sustainable alternative sources of energy. The Chernobyl nuclear accident, which occurred 26 years ago in the former Soviet Union, released an immense amount of radioactivity over vast territories of Belarus, Ukraine, and the Russian Federation, extending into northern Europe, and became the most severe accident in the history of the nuclear industry. This disaster was a result of numerous factors including inadequate nuclear power plant design, human errors, and violation of safety measures. The lessons learned from nuclear accidents will continue to strengthen the safety design of new reactor installations, but with more than 400 active nuclear power stations worldwide and 104 reactors in the Unites States, it is essential to reassess fundamental issues related to the Chernobyl experience as it continues to evolve. This article summarizes early and late events of the incident, the impact on thyroid health, and attempts to reduce agricultural radioactive contamination. PMID:22853775

  20. OVERVIEW OF MODULAR HTGR SAFETY CHARACTERIZATION AND POSTULATED ACCIDENT BEHAVIOR LICENSING STRATEGY

    SciTech Connect

    Ball, Sydney J

    2014-06-01

    This report provides an update on modular high-temperature gas-cooled reactor (HTGR) accident analyses and risk assessments. One objective of this report is to improve the characterization of the safety case to better meet current regulatory practice, which is commonly geared to address features of today s light water reactors (LWRs). The approach makes use of surrogates for accident prevention and mitigation to make comparisons with LWRs. The safety related design features of modular HTGRs are described, along with the means for rigorously characterizing accident selection and progression methodologies. Approaches commonly used in the United States and elsewhere are described, along with detailed descriptions and comments on design basis (and beyond) postulated accident sequences.

  1. Markov Model of Severe Accident Progression and Management

    SciTech Connect

    Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

    2012-06-25

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  2. Chernobyl accident and the problem of {sup 241}Am

    SciTech Connect

    Pazukhin, E.M.; Drozd, I.P.; Tokarevskii, V.V.

    1995-07-01

    The accumulation and decay of {sup 241}Am formed in the reactor of the fourth block of the Chernobyl atomic energy station at the time of the accident of April 26, 1986 are analyzed. Possible pathways for {sup 241}Am uptake in man are examined. The maximum equivalent dose after 70 years is estimated for a critical population living at the edge of the exclusion zone.

  3. Safe new reactor for radionuclide production

    SciTech Connect

    Gray, P.L.

    1995-02-15

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible.

  4. Reactor vessel lower head integrity

    SciTech Connect

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  5. KERENA safety concept in the context of the Fukushima accident

    SciTech Connect

    Zacharias, T.; Novotny, C.; Bielor, E.

    2012-07-01

    Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basic physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)

  6. Rear-end accident victims. Importance of understanding the accident.

    PubMed Central

    Sehmer, J. M.

    1993-01-01

    Family physicians regularly treat victims of rear-end vehicle accidents. This article describes how taking a detailed history of the accident and understanding the significance of the physical events is helpful in understanding and anticipating patients' morbidity and clinical course. Eight questions to ask patients are suggested to help physicians understand the severity of injury. PMID:8495140

  7. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  8. Oxygenated fraction and mass of organic aerosol from direct emission and atmospheric processing measured on the R/V Ronald Brown during TEXAQS/GoMACCS 2006

    NASA Astrophysics Data System (ADS)

    Russell, L. M.; Takahama, S.; Liu, S.; Hawkins, L. N.; Covert, D. S.; Quinn, P. K.; Bates, T. S.

    2009-04-01

    Submicron particles collected on Teflon filters aboard the R/V Ronald Brown during the Texas Air Quality Study and Gulf of Mexico Atmospheric Composition and Climate Study (TexAQS/GoMACCS) 2006 in and around the port of Houston, Texas, were measured by Fourier transform infrared (FTIR) and X-ray fluorescence for organic functional groups and elemental composition. Organic mass (OM) concentrations (1-25 μg m-3) for ambient particle samples measured by FTIR showed good agreement with measurements made with an aerosol mass spectrometer. The fractions of organic mass identified as alkane and carboxylic acid groups were 47% and 32%, respectively. Three different types of air masses were identified on the basis of the air mass origin and the radon concentration, with significantly higher carboxylic acid group mass fractions in air masses from the north (35%) than the south (29%) or Gulf of Mexico (26%). Positive matrix factorization analysis attributed carboxylic acid fractions of 30-35% to factors with mild or strong correlations (r > 0.5) to elemental signatures of oil combustion and 9-24% to wood smoke, indicating that part of the carboxylic acid fraction of OM was formed by the same sources that controlled the metal emissions, namely the oil and wood combustion activities. The implication is that a substantial part of the measured carboxylic acid contribution was formed independently of traditionally "secondary" processes, which would be affected by atmospheric (both photochemical and meteorological) conditions and other emission sources. The carboxylic acid group fractions in the Gulf of Mexico and south air masses (GAM and SAM, respectively) were largely oil combustion emissions from ships as well as background marine sources, with only limited recent land influences (based on radon concentrations). Alcohol groups accounted for 14% of OM (mostly associated with oil combustion emissions and background sources), and amine groups accounted for 4% of OM in all air

  9. Safety analysis results for cryostat ingress accidents in ITER

    SciTech Connect

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits. 6 refs., 2 figs., 1 tab.

  10. Safety analysis results for cryostat ingress accidents in ITER

    SciTech Connect

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1996-12-31

    Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  11. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    NASA Astrophysics Data System (ADS)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  12. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    SciTech Connect

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.

  13. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR & BWR accident sequences

    SciTech Connect

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-08-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences.

  14. Report to the American Physical Society of the Study Group on Radionuclide Release From Severe Accidents at Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Shaw, George

    The release of radioiodine during the Three Mile Island (TMI) accident was more than an order of magnitude smaller than what had been predicted from analyses of hypothetical nuclear accidents. The Reactor Safety Study of 1975 (RSS), which carried out the analyses, is a fundamental factor in formulating regulations concerned with such accidents. This American Physical Society (APS) study group report is a result of the obvious need to reevaluate the RSS analysis of the “source term,” that is, the amount of various radionuclides that are predicted to be emitted under various reactor failure scenarios.The report includes an introductory background to the history of nuclear reactor accidents and accident studies and to the health aspects of radionuclide releases. It then describes nuclear reactors and reactor failure modes, including reasonably detailed descriptions of particular modes thought to be especially critical. The most extensive discussion concerns the chemical and physical processes important in the generation, transport, and release of radionuclides. The large computer codes used to model these processes are considered and evaluated. The results of some of the computer runs are examined in the light of a simplified but informative model to evaluate those features of an accident that are most likely to affect the source term. A review of the research programs currently underway precedes both the study group conclusions about the need to revise the source terms from those in the RSS and recommendations for further studies that are necessary to better evaluate the source term.

  15. German aircraft accident statistics, 1930

    NASA Technical Reports Server (NTRS)

    Weitzmann, Ludwig

    1932-01-01

    The investigation of all serious accidents, involving technical defects in the airplane or engine, is undertaken by the D.V.L. in conjunction with the imperial traffic minister and other interested parties. All accidents not clearly explained in the reports are subsequently cleared up.

  16. Weather types and traffic accidents.

    PubMed

    Klaić, Z B

    2001-06-01

    Traffic accident data for the Zagreb area for the 1981-1982 period were analyzed to investigate possible relationships between the daily number of accidents and the weather conditions that occurred for the 5 consecutive days, starting two days before the particular day. In the statistical analysis of low accident days weather type classification developed by Poje was used. For the high accident days a detailed analyses of surface and radiosonde data were performed in order to identify possible front passages. A test for independence by contingency table confirmed that conditional probability of the day with small number of accidents is the highest, provided that one day after it "N" or "NW" weather types occur, while it is the smallest for "N1" and "Bc" types. For the remaining 4 days of the examined periods dependence was not statistically confirmed. However, northern ("N", "NE" and "NW") and anticyclonic ("Vc", "V4", "V3", "V2" and "mv") weather types predominated during 5-days intervals related to the days with small number of accidents. On the contrary, the weather types with cyclonic characteristics ("N1", "N2", "N3", "Bc", "Dol1" and "Dol"), that are generally accompanied by fronts, were the rarest. For 85% days with large number of accidents, which had not been caused by objective circumstances (such as poor visibility, damaged or slippery road etc.), at least one front passage was recorded during the 3-days period, starting one day before the day with large number of accidents. PMID:11787547

  17. First Responders and Criticality Accidents

    SciTech Connect

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  18. Fukushima nuclear power plant accident was preventable

    NASA Astrophysics Data System (ADS)

    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

  19. MELCOR analyses of severe accident scenarios in Oconee, a B&W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-03-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  20. MELCOR analyses of severe accident scenarios in Oconee, a B W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S. ); Foulds, R. )

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  1. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  2. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  3. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  4. Industrial accidents triggered by lightning.

    PubMed

    Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

    2010-12-15

    Natural disasters can cause major accidents in chemical facilities where they can lead to the release of hazardous materials which in turn can result in fires, explosions or toxic dispersion. Lightning strikes are the most frequent cause of major accidents triggered by natural events. In order to contribute towards the development of a quantitative approach for assessing lightning risk at industrial facilities, lightning-triggered accident case histories were retrieved from the major industrial accident databases and analysed to extract information on types of vulnerable equipment, failure dynamics and damage states, as well as on the final consequences of the event. The most vulnerable category of equipment is storage tanks. Lightning damage is incurred by immediate ignition, electrical and electronic systems failure or structural damage with subsequent release. Toxic releases and tank fires tend to be the most common scenarios associated with lightning strikes. Oil, diesel and gasoline are the substances most frequently released during lightning-triggered Natech accidents. PMID:20817399

  5. Probabilistic methods for accident-progression analysis

    SciTech Connect

    Jamali, K. M.

    1981-01-01

    Probabilistic methods that can be used as basis for deterministic calculations of transients or accidents in nuclear power plants are described. They include obtaining initiator-dependent sequences on the component level and related analyses, propagation of primary event uncertainties in the ranking of sequences, and detailed treatment of dependent failures. The results are shown for protected transients in the short term forced circulation phase of decay heat removal in the Clinch River Breeder Reactor. Higher values of unavailabilities are obtained than previous works as a result of more detailed common cause/mode failure modeling. The unavailability of decay heat removal by forced circulation for the loss of off-site power and loss of main feedwater system initiators is estimated at 4 x 10/sup -3//yr and 9 x 10/sup -3//yr, respectively. 15 refs., 1 fig., 2 tabs.

  6. Accident analysis of the windowless target system

    SciTech Connect

    Bianchi, F.; Ferri, R.

    2006-07-01

    Transmutation systems are able to reduce the radio-toxicity and amount of High-Level Wastes (HLW), which are the main concerns related to the peaceful use of nuclear energy, and therefore they should make nuclear energy more easily acceptable by population. A transmutation system consists of a sub-critical fast reactor, an accelerator and a Target System, where the spallation reactions needed to sustain the chain reaction take place. Three options were proposed for the Target System within the European project PDS-XADS (Preliminary Design Studies on an Experimental Accelerator Driven System): window, windowless and solid. This paper describes the constraints taken into account in the design of the windowless Target System for the large Lead-Bismuth-Eutectic cooled XADS and deals with the results of the calculations performed to assess the behaviour of the target during some accident sequences related to pump trips. (authors)

  7. Experimental assessment of computer codes used for safety analysis of integral reactors

    SciTech Connect

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B.

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  8. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    SciTech Connect

    Not Available

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  9. Overview of the radiological accidents in the world, updated December 1989.

    PubMed

    Nénot, J C

    1990-06-01

    Radiological accidents can be divided into two categories, depending on whether the accident involves large groups of the population with relatively low doses or a few individuals with high doses resulting in acute health effects. The accidents involving large groups are related to the dispersion of radioactive materials in the environment; although they may have different causes, the source is always very important. Most of the accidents which have occurred originated in civilian installations; two reactor accidents can be considered without any human consequences: the accidents in the UK (Windscale) in 1957 and in the USA (TMI) in 1979. The Chernobyl accident (USSR) in 1986 resulted in extensive contamination of the environment, with non-negligible doses to the population around the plant and large collective doses in the northern hemisphere; in addition, the Chernobyl accident caused the deaths of 31 workers and firemen who intervened to bring the installation back under control. Violations of the most elementary safety rules for the operation of medical sources were at the origin of two severe environmental contaminations with human consequences: in Mexico (1983-4) and Brazil (1987), with sources of 60Co and 137Cs, respectively. The accidents concerning only a few individuals are not always known with the same documented accuracy. Between the 1940s and 1960s six critical accidents caused eight deaths; since then only one has occurred, in Argentina in 1983. The fatal radiation accidents are due to high-energy radiation sources, such as 60Co, 137Cs, and 192Ir. The total number of deaths which has been registered is 28. The accidents related to internal exposure are not exceptional, but result very rarely in health consequences. PMID:1971835

  10. Columbia Accident Investigation Board

    NASA Technical Reports Server (NTRS)

    2003-01-01

    The Columbia Accident Investigation Board gathers for a second day for its third public hearing, held in Cape Canaveral, Florida. The CAIB was set up to examine STS-107 and analyze exploratory tests. Navy Admiral Harold W. 'Hal' Gehman Jr. was designated as the Chairman of the Board. From left to right in this photo sit Board Members Steven B. Wallace, Scott Hubbard, Dr. John Logsdon, Rear Admiral Stephen Turcotte, Hal Gehman, General Duane Deal, Dr. Douglas Osheroff, and Maj. General Kenneth W. Hess. Not shown are Maj. General John Barry, Dr. James N. Hallock, Roger Tetrault, Dr. Sheila Widnall, and Dr. Sally Ride. For more information on STS-107, please see GRIN Columbia General Explanation

  11. Transport aircraft accident dynamics

    NASA Technical Reports Server (NTRS)

    Cominsky, A.

    1982-01-01

    A study was carried out of 112 impact survivable jet transport aircraft accidents (world wide) of 27,700 kg (60,000 lb.) aircraft and up extending over the last 20 years. This study centered on the effect of impact and the follow-on events on aircraft structures and was confined to the approach, landing and takeoff segments of the flight. The significant characteristics, frequency of occurrence and the effect on the occupants of the above data base were studied and categorized with a view to establishing typical impact scenarios for use as a basis of verifying the effectiveness of potential safety concepts. Studies were also carried out of related subjects such as: (1) assessment of advanced materials; (2) human tolerance to impact; (3) merit functions for safety concepts; and (4) impact analysis and test methods.

  12. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  13. Chernobyl lessons learned review of N Reactor

    SciTech Connect

    Weber, E.T.; McNeece, J.P.; Omberg, R.P.; Stepnewski, D.D.; Lutz, R.J.; Henry, R.E.; Bonser, K.D.; Miller, N.R.

    1987-10-01

    A broad-base review of the N Reactor plant, design characteristics, administrative controls and responses unique to upset conditions has been completed. The review was keyed to Nuclear Regulatory Commission (NRC)-defined issues associated with the Chernobyl accident. Physical features of N Reactor that preclude an accident like Chernobyl include: lack of autocatalytic reactivity insertion (i.e., negative coolant void and power coefficents) and two separate, fast-acting scram systems. Administrative controls in place at N Reactor would effectively protect against the operator errors and safety violations that set up the Chernobyl accident. Several items were identified where further near-term action is appropriate to ensure effectiveness of existing safety features: Resolve a question concerning the exact point at which Emergency Core Cooling System (ECCS) activation by manual actions should be implemented or deferred if automatic ECCS trip fails. Ensure appropriate revision of the Emergency Response Guides and full communication of the correct procedure to all Operations, Safety and cognizant Technology staff. Train reactor operators in the currently recognized significance of the Graphite and Shield Cooling System (GSCS) in severe accident situations and cover this appropriately in the Emergency Response Guides. Complete reviews which establish an independent verification that pressure tube rupture will not propagate to other tubes. 15 refs., 3 tabs.

  14. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  15. Analysis of the SL-1 Accident Using RELAPS5-3D

    SciTech Connect

    Francisco, A.D. and Tomlinson, E. T.

    2007-11-08

    On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with a discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).

  16. Physics of reactor safety. Quarterly report, October-December 1980. Volume IV

    SciTech Connect

    Not Available

    1981-02-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  17. AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

    SciTech Connect

    Knight, T.D.

    1996-07-01

    The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA.

  18. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  19. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    SciTech Connect

    Wenxi Tian; Suizheng Qiu; Guanghui Su; Dounan Jia; Xingmin Liu - China Institute of Atomic Energy

    2006-07-01

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  20. Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

    SciTech Connect

    Rempe, Joy L.; Knudson, Darrell L.

    2015-02-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

  1. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  2. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  3. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  4. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  5. [Hidden statistics of traffic accidents].

    PubMed

    Nordentoft, E L; Larsen, C F; Jørgensen, H R

    1989-10-23

    Only 19% of the 3,071 injured persons who were treated in the casualty department of Odense Hospital following traffic accidents in 1987 could be found again in the police registers of traffic accidents from the same region. All of the registrations from the police registers from the central region could be found again in the casualty department. In 1971, the corresponding coverage was 36%. The degree of coverage is particularly low for single bicycle accidents, other bicycle accidents, other single accidents and the hours immediately after midnight. Considerable disagreement exists concerning registration of the use of safety belts and crash helmets. In Odense, the municipal road authorities utilize the localization of the accidents reported by the casualty department. The decrease in the degree of coverage is due mainly to an increasing proportion of bicycle accidents. Where casualties require admission to hospital, the coverage is approximately 75%. This has remained unchanged throughout the years and it is therefore suggested that this proportion should be employed as indicator of the effect of the majority of prophylactic measures. In addition, proposals are made for simplification of the police registration forms. PMID:2588362

  6. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  7. [Prevention of bicycle accidents].

    PubMed

    Zwipp, H; Barthel, P; Bönninger, J; Bürkle, H; Hagemeister, C; Hannawald, L; Huhn, R; Kühn, M; Liers, H; Maier, R; Otte, D; Prokop, G; Seeck, A; Sturm, J; Unger, T

    2015-04-01

    For a very precise analysis of all injured bicyclists in Germany it would be important to have definitions for "severely injured", "seriously injured" and "critically injured". By this, e.g., two-thirds of surgically treated bicyclists who are not registered by the police could become available for a general analysis. Elderly bicyclists (> 60 years) are a minority (10 %) but represent a majority (50 %) of all fatalities. They profit most by wearing a helmet and would be less injured by using special bicycle bags, switching on their hearing aids and following all traffic rules. E-bikes are used more and more (145 % more in 2012 vs. 2011) with 600,000 at the end of 2011 and are increasingly involved in accidents but still have a lack of legislation. So even for pedelecs 45 with 500 W and a possible speed of 45 km/h there is still no legislative demand for the use of a protecting helmet. 96 % of all injured cyclists in Germany had more than 0.5 ‰ alcohol in their blood, 86 % more than 1.1 ‰ and 59 % more than 1.7 ‰. Fatalities are seen in 24.2 % of cases without any collision partner. Therefore the ADFC calls for a limit of 1.1 ‰. Some virtual studies conclude that integrated sensors in bicycle helmets which would interact with sensors in cars could prevent collisions or reduce the severity of injury by stopping the cars automatically. Integrated sensors in cars with opening angles of 180° enable about 93 % of all bicyclists to be detected leading to a high rate of injury avoidance and/or mitigation. Hanging lamps reduce with 35 % significantly bicycle accidents for children, traffic education for children and special trainings for elderly bicyclists are also recommended as prevention tools. As long as helmet use for bicyclists in Germany rates only 9 % on average and legislative orders for using a helmet will not be in force in the near future, coming up campaigns seem to be necessary to be promoted by the Deutscher

  8. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  9. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  10. Thyroid consequences of the Chernobyl nuclear accident.

    PubMed

    Pacini, F; Vorontsova, T; Molinaro, E; Shavrova, E; Agate, L; Kuchinskaya, E; Elisei, R; Demidchik, E P; Pinchera, A

    1999-12-01

    It is well recognized that the use of external irradiation of the head and neck to treat patients with various non-thyroid disorders increases their risk of developing papillary thyroid carcinoma years after radiation exposure. An increased risk of thyroid cancer has also been reported in survivors of the atomic bombs in Japan, as well as in Marshall Island residents exposed to radiation during the testing of hydrogen bombs. More recently, exposure to radioactive fallout as a result of the Chernobyl nuclear reactor accident has clearly caused an enormous increase in the incidence of childhood thyroid carcinoma in Belarus, Ukraine, and, to a lesser extent, in the Russian Federation, starting in 1990. When clinical and epidemiological features of thyroid carcinomas diagnosed in Belarus after the Chernobyl accident are compared with those of naturally occurring thyroid carcinomas in patients of the same age group in Italy and France, it becomes apparent that the post-Chernobyl thyroid carcinomas were much less influenced by gender, virtually always papillary (solid and follicular variants), more aggressive at presentation and more frequently associated with thyroid autoimmunity. Gene mutations involving the RET proto-oncogene, and less frequently TRK, have been shown to be causative events specific for papillary cancer. RET activation was found in nearly 70% of the patients who developed papillary thyroid carcinomas following the Chernobyl accident. In addition to thyroid cancer, radiation-induced thyroid diseases include benign thyroid nodules, hypothyroidism and autoimmune thyroiditis, with or without thyroid insufficiency, as observed in populations after environmental exposure to radioisotopes of iodine and in the survivors of atomic bomb explosions. On this basis, the authors evaluated thyroid autoimmune phenomena in normal children exposed to radiation after the Chernobyl accident. The results demonstrated an increased prevalence of circulating thyroid

  11. Underreporting of maritime accidents to vessel accident databases.

    PubMed

    Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

    2011-11-01

    Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. PMID:21819835

  12. Nuclear Reactor Engineering Analysis Laboratory

    SciTech Connect

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  13. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    SciTech Connect

    Nourbakhsh, H.P.

    1993-11-01

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

  14. Markov Model of Accident Progression at Fukushima Daiichi

    SciTech Connect

    Cuadra A.; Bari R.; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G.; Mubayi, V.; Pratt, T.; Yue, M.

    2012-11-11

    On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

  15. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  16. Fuel behavior comparison for a research reactor

    NASA Astrophysics Data System (ADS)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  17. Columbia Accident Probe Widens

    NASA Technical Reports Server (NTRS)

    Covault, Craig

    2003-01-01

    The Columbia Accident Investigation Board has identified about a dozen shuttle program safety concerns it will address in its final report, in addition to foam shedding from the Lockheed Martin external tank-believed by many board members to be the direct cause for the loss of Columbia and her crew. As new evidence narrows the location of Columbia's left-wing breach to a lower corner of reinforced carbon-carbon (RCC) Panel 8 and its adjoining T-seal, the board is broadening its penetration of other shuttle safety issues. As the board works in Houston, United Space Alliance technicians here at Kennedy last week sent the first six of 22 RCC panels from the orbiter Atlantis left wing to Vought Aircraft Industries Inc. in Dallas for extensive testing to assess their integrity. The move is a key step toward both returning the shuttle to flight with Atlantis and obtaining more data on RCC panels subjected to fewer flights, and less exposure to the weather, than the older panels used on Columbia.

  18. Aircraft accidents : method of analysis

    NASA Technical Reports Server (NTRS)

    1929-01-01

    This report on a method of analysis of aircraft accidents has been prepared by a special committee on the nomenclature, subdivision, and classification of aircraft accidents organized by the National Advisory Committee for Aeronautics in response to a request dated February 18, 1928, from the Air Coordination Committee consisting of the Assistant Secretaries for Aeronautics in the Departments of War, Navy, and Commerce. The work was undertaken in recognition of the difficulty of drawing correct conclusions from efforts to analyze and compare reports of aircraft accidents prepared by different organizations using different classifications and definitions. The air coordination committee's request was made "in order that practices used may henceforth conform to a standard and be universally comparable." the purpose of the special committee therefore was to prepare a basis for the classification and comparison of aircraft accidents, both civil and military. (author)

  19. 49 CFR 801.32 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Accident reports. 801.32 Section 801.32... PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.32 Accident reports. (a) The NTSB....S. civil transportation accidents, in accordance with 49 U.S.C. 1131(e). (b) These reports may...

  20. 29 CFR 1960.29 - Accident investigation.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 29 Labor 9 2011-07-01 2011-07-01 false Accident investigation. 1960.29 Section 1960.29 Labor... MATTERS Inspection and Abatement § 1960.29 Accident investigation. (a) While all accidents should be investigated, including accidents involving property damage only, the extent of such investigation shall...

  1. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  2. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 4 2011-10-01 2011-10-01 false Accident reports. 230.22 Section 230.22... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam... persons, the railroad on whose line the accident occurred shall immediately make a telephone report of...

  3. 49 CFR 845.40 - Accident report.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Accident report. 845.40 Section 845.40... RULES OF PRACTICE IN TRANSPORTATION; ACCIDENT/INCIDENT HEARINGS AND REPORTS Board Reports § 845.40 Accident report. (a) The Board will issue a detailed narrative accident report in connection with...

  4. 29 CFR 1960.29 - Accident investigation.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 9 2010-07-01 2010-07-01 false Accident investigation. 1960.29 Section 1960.29 Labor... MATTERS Inspection and Abatement § 1960.29 Accident investigation. (a) While all accidents should be investigated, including accidents involving property damage only, the extent of such investigation shall...

  5. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Accident reports. 230.22 Section 230.22... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam... persons, the railroad on whose line the accident occurred shall immediately make a telephone report of...

  6. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 shall as soon...

  7. 49 CFR 845.40 - Accident report.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Accident report. 845.40 Section 845.40... RULES OF PRACTICE IN TRANSPORTATION; ACCIDENT/INCIDENT HEARINGS AND REPORTS Board Reports § 845.40 Accident report. (a) The Board will issue a detailed narrative accident report in connection with...

  8. 49 CFR 801.32 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Accident reports. 801.32 Section 801.32... PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.32 Accident reports. (a) The NTSB....S. civil transportation accidents, in accordance with 49 U.S.C. 1131(e). (b) These reports may...

  9. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect

    Wang, S.-J.; Chiang, K.-S.; Chiang, S.-C.

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  10. Advanced accident sequence precursor analysis level 2 models

    SciTech Connect

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L.

    1996-03-01

    The U.S. Nuclear Regulatory Commission Accident Sequence Precursor program pursues the ultimate objective of performing risk significant evaluations on operational events (precursors) occurring in commercial nuclear power plants. To achieve this objective, the Office of Nuclear Regulatory Research is supporting the development of simple probabilistic risk assessment models for all commercial nuclear power plants (NPP) in the U.S. Presently, only simple Level 1 plant models have been developed which estimate core damage frequencies. In order to provide a true risk perspective, the consequences associated with postulated core damage accidents also need to be considered. With the objective of performing risk evaluations in an integrated and consistent manner, a linked event tree approach which propagates the front end results to back end was developed. This approach utilizes simple plant models that analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude and timing of a radioactive release to the environment, and calculate the consequences for a given release. Detailed models and results from previous studies, such as the NUREG-1150 study, are used to quantify these simple models. These simple models are then linked to the existing Level 1 models, and are evaluated using the SAPHIRE code. To demonstrate the approach, prototypic models have been developed for a boiling water reactor, Peach Bottom, and a pressurized water reactor, Zion.

  11. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  12. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  13. Spine Immobilizer for Accident Victims

    NASA Technical Reports Server (NTRS)

    Vykukal, H. C.; Lampson, K.

    1983-01-01

    Proposed conformal bladder filled with tiny spheres called "microballoons," enables spine of accident victim to be rapidly immobilized and restrained and permit victim to be safely removed from accident scene in extremely short time after help arrives. Microballoons expand to form rigid mass when pressure within bladder is less than ambient. Bladder strapped to victim is also strapped to rescue chair. Void between bladder and chair is filled with cloth wedges.

  14. Preparation, conduct, and experimental results of the AVR loss-of-coolant accident simulation test

    SciTech Connect

    Kruger, K.; Bergerfurth, A.; Burger, S.; Pohl, P.; Wimmers, M. ); Cleveland, J.C. )

    1991-02-01

    A loss-of-coolant accident (LOCA) is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small high-temperature gas-cooled reactor (HTGR) design, LOCA simulation tests have been conducted at the Arbeitsgemeinschaft Versuchsreaktor (AVR), the German pebble-bed-high-temperature reactor plant. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions without emergency cooling. This paper presents the planning and licensing activities including pretest predictions performed for the LOCA test are described, and the conduct of the test and experimental results. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The test demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into optimized modular HTGR designs. The main LOCA test lasted for 5 days. After the test began, core temperatures increased for {approx}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded the simulated decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components.

  15. The Accident at Fukushima: What Happened?

    SciTech Connect

    Fujie, Takao

    2012-07-01

    At 2:46 PM, on the coast of the Pacific Ocean in eastern Japan, people were spending an ordinary afternoon. The earthquake had a magnitude of 9.0, the fourth largest ever recorded in the world. Avery large number of aftershocks were felt after the initial earthquake. More than 100 of them had a magnitude of over 6.0. There were very few injured or dead at this point. The large earthquake caused by this enormous crustal deformation spawned a rare and enormous tsunami that crashed down 30-40 minutes later. It easily cleared the high levees, washing away cars and houses and swallowing buildings of up to three stories in height. The largest tsunami reading taken from all regions was 40 meters in height. This tsunami reached the West Coast of the United States and the Pacific coast of South America, with wave heights of over two meters. It was due to this tsunami that the disaster became one of a not imaginable scale, which saw the number of dead or missing reach about 20,000 persons. The enormous tsunami headed for 15 nuclear power plants on the Pacific coast, but 11 power plants withstood the tsunami and attained cold shutdown. The flood height of the tsunami that struck each power station ranged to a maximum of 15 meters. The Fukushima Daiichi Nuclear Power Plant Units experienced the largest and the cores of three reactors suffered meltdown. As a result, more than 160,000 residents were forced to evacuate, and are still living in temporary accommodation. The main focus of this presentation is on what happened at the Fukushima Daiichi, and how station personnel responded to the accident, with considerable international support. A year after the Fukushima Daiichi accident, Japan is in the process of leveraging the lessons learned from the accident to further improve the safety of nuclear power facilities and regain the trust of society. In this connection, not only international organizations, including IAEA, and WANO, but also governmental organizations and nuclear

  16. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    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.

  17. Analytical Study on Fire and Explosion Accidents Assumed in HTGR Hydrogen Production System

    SciTech Connect

    Inaba, Yoshitomo; Nishihara, Tetsuo; Nitta, Yoshikazu

    2004-04-15

    One of the most important safety design issues for a hydrogen production system coupling with a high-temperature gas-cooled reactor (HTGR) is to ensure reactor safety against fire and explosion accidents because a large amount of combustible fluid is dealt with in the system. The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the high-temperature engineering test reactor (HTTR). In the plan, we developed the P2A code system to analyze event sequences and consequences in detail on the fire and explosion accidents assumed in the HTGR or HTTR hydrogen production system. This paper describes the three accident scenarios assumed in the system, the structure of P2A, the analysis procedure with P2A, and the results of the numerical analyses based on the accident scenarios. It is shown that P2A is a useful tool for the accident analysis in the system.

  18. Program for the Analysis of Reactor Transients

    Energy Science and Technology Software Center (ESTSC)

    2002-01-29

    This program is designed for use in predicting the course of and consequence of nondestructive accidents in research and test reactor cores. It is intended primarily for the analysis of plate type research and test reactors and has been subjected to extensive comparisons with the SPERT I and SPERT II experiments. These comparisons were quite favorable for a wide range of transients up to and including melting of the clad. Favorable comparisons have also beenmore » made for TRIGA reactor pulses in pin geometry. The PARET/ANL code has been used by the RERTR (Reduced Enrichment Research and Test Reactor) Program for the safety evaluation of many of the candidate reactors for reduced enrichment.« less

  19. Program for the Analysis of Reactor Transients

    SciTech Connect

    Woodruff, W. L.; Smith, R. S.

    2002-01-29

    This program is designed for use in predicting the course of and consequence of nondestructive accidents in research and test reactor cores. It is intended primarily for the analysis of plate type research and test reactors and has been subjected to extensive comparisons with the SPERT I and SPERT II experiments. These comparisons were quite favorable for a wide range of transients up to and including melting of the clad. Favorable comparisons have also been made for TRIGA reactor pulses in pin geometry. The PARET/ANL code has been used by the RERTR (Reduced Enrichment Research and Test Reactor) Program for the safety evaluation of many of the candidate reactors for reduced enrichment.

  20. Incorporation of phenomenological uncertainties in probabilistic safety analysis - application to LMFBR core disruptive accident energetics

    SciTech Connect

    Najafi, B; Theofanous, T G; Rumble, E T; Atefi, B

    1984-08-01

    This report describes a method for quantifying frequency and consequence uncertainty distribution associated with core disruptive accidents (CDAs). The method was developed to estimate the frequency and magnitude of energy impacting the reactor vessel head of the Clinch River Breeder Plant (CRBRP) given the occurrence of hypothetical CDAs. The methodology is illustrated using the CRBR example.