Environmental consequences of postulated plutonium releases from Westinghouse PFDL, Cheswick, Pennsylvania, as a result of severe natural phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

McPherson, R.B.; Watson, E.C.

1979-06-01

Potential environmental consequences in terms of radiation dose to people are presented for postulated accidents due to earthquakes, tornadoes, high straight-line winds, and floods. Maximum plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays except Earthquake No. 4 and the 260-mph tornado. The most likely maximum residual plutonium contamination estimated to be deposited offsite following Earthquake No. 4, and themore » 200-mph and 260-mph tornadoes are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 ..mu..Ci/m/sup 2/. The deposition values following the other severe natural phenomena are below the EPA proposed guideline.« less

Analysis and testing of Koornstra-type induced exposure models

DOT National Transportation Integrated Search

1985-10-01

Induced exposure models postulate a structure for accident data which permits the : estimation of two factors: exposure and proneness. Since information on exposure : is needed in order to assess the accident risk of different driver, vehicle, and : ...

78 FR 60321 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-01

... sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that... in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR... changes to the plant design basis or postulated accidents resulting from potential tube degradation. The...

DOE Office of Scientific and Technical Information (OSTI.GOV)

PIEPHO, M.G.

Four bounding accidents postulated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing a hydrogen explosion, and a fire breaching filter vessel and enclosure. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

78 FR 75379 - Entergy Operations, Inc.; Combined License Application for Grand Gulf Unit 3; Exemption From the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-11

... accidents are increased. Therefore, there is no undue risk to public health and safety. Plant construction... at this time. Based on the nature of the requested exemption as described above, no new accident... a license decision is made, the probability of postulated accidents is not increased. Additionally...

To Err is Human Case Reports of Two Military Aircraft Accidents

PubMed Central

Dikshit, Mohan B

2010-01-01

It has been postulated that pilot error or in-flight incapacitation may be the main contributory factors to 70–80% of aircraft accidents. Two fatal aircraft accidents are presented in which either of the above possibilities may have played a role. The first case report describes an erroneous decision by a fighter pilot to use a seat position adjustment of the ejection seat leading to fatal injuries when he had to eject from his aircraft. Injuries to the body of the pilot, and observations on the state of his flying clothing and the ejection seat were used to postulate the mechanism of fatal injury and establish the cause of the accident. The second case report describes the sequence of events which culminated in the incapacitation of a fighter pilot while executing a routine manouevre. This resulted in a fatal air crash. Possible contributions of environmental factors which may have resulted in failure of his physiological mechanisms are discussed. PMID:21509093

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozluk, M.J.; Vijay, D.K.

Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption thatmore » resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.« less

Radioactive release during nuclear accidents in Chernobyl and Fukushima

NASA Astrophysics Data System (ADS)

Nur Ain Sulaiman, Siti; Mohamed, Faizal; Rahim, Ahmad Nabil Ab

2018-01-01

Nuclear accidents that occurred in Chernobyl and Fukushima have initiated many research interests to understand the cause and mechanism of radioactive release within reactor compound and to the environment. Common types of radionuclide release are the fission products from the irradiated fuel rod itself. In case of nuclear accident, the focus of monitoring will be mostly on the release of noble gases, I-131 and Cs-137. As these are the only accidents have been rated within International Nuclear Events Scale (INES) Level 7, the radioactive release to the environment was one of the critical insights to be monitored. It was estimated that the release of radioactive material to the atmosphere due to Fukushima accident was approximately 10% of the Chernobyl accident. By referring to the previous reports using computational code systems to model the release rate, the release activity of I-131 and Cs-137 in Chernobyl was significantly higher compare to Fukushima. The simulation code also showed that Chernobyl had higher release rate of both radionuclides on the day of accident. Other factors affecting the radioactive release for Fukushima and Chernobyl accidents such as the current reactor technology and safety measures are also compared for discussion.

Unexpected source of Fukushima-derived radiocesium to the coastal ocean of Japan

PubMed Central

Sanial, Virginie; Buesseler, Ken O.; Charette, Matthew A.; Nagao, Seiya

2017-01-01

There are 440 operational nuclear reactors in the world, with approximately one-half situated along the coastline. This includes the Fukushima Dai-ichi Nuclear Power Plant (FDNPP), which experienced multiple reactor meltdowns in March 2011 followed by the release of radioactivity to the marine environment. While surface inputs to the ocean via atmospheric deposition and rivers are usually well monitored after a nuclear accident, no study has focused on subterranean pathways. During our study period, we found the highest cesium-137 (137Cs) levels (up to 23,000 Bq⋅m−3) outside of the FDNPP site not in the ocean, rivers, or potable groundwater, but in groundwater beneath sand beaches over tens of kilometers away from the FDNPP. Here, we present evidence of a previously unknown, ongoing source of Fukushima-derived 137Cs to the coastal ocean. We postulate that these beach sands were contaminated in 2011 through wave- and tide-driven exchange and sorption of highly radioactive Cs from seawater. Subsequent desorption of 137Cs and fluid exchange from the beach sands was quantified using naturally occurring radium isotopes. This estimated ocean 137Cs source (0.6 TBq⋅y−1) is of similar magnitude as the ongoing releases of 137Cs from the FDNPP site for 2013–2016, as well as the input of Fukushima-derived dissolved 137Cs via rivers. Although this ongoing source is not at present a public health issue for Japan, the release of Cs of this type and scale needs to be considered in nuclear power plant monitoring and scenarios involving future accidents. PMID:28973919

75 FR 19665 - Final Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-15

... processes, including welding, heat treating, and nondestructive testing, are controlled and accomplished by... that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the...

Haz-Map Glossary

MedlinePlus

... lung. Radiation Accident Large-scale accidents from atomic bomb testing fallout released iodine-131 and strontium-90. ... lung. Radiation Accident Large-scale accidents from atomic bomb testing fallout released iodine-131 and strontium-90. ...

OXIDATION OF INCONEL 718 IN AIR AT TEMPERATURES FROM 973K TO 1620K.

DOE Office of Scientific and Technical Information (OSTI.GOV)

GREENE,G.A.; FINFROCK,C.C.

2000-10-01

As part of the APT project, it was necessary to quantify the release of tungsten from the APT spallation target during postulated accident conditions in order to develop accident source terms for accident consequence characterization. Experiments with tungsten rods at high temperatures in a flowing steam environment characteristic of postulated accidents revealed that considerable vaporization of the tungsten occurred as a result of reactions with the steam and that the aerosols which formed were readily transported away from the tungsten surfaces, thus exposing fresh tungsten to react with more steam. The resulting tungsten release fractions and source terms were undesirablemore » and it was decided to clad the tungsten target with Inconel 718 in order to protect it from contact with steam during an accident and mitigate the accident source term and the consequences. As part of the material selection criteria, experiments were conducted with Inconel 718 at high temperatures to evaluate the rate of oxidation of the proposed clad material over as wide a temperature range as possible, as well as to determine the high-temperature failure limit of the material. Samples of Inconel 718 were inserted into a preheated furnace at temperatures ranging from 973 K to 1620 K and oxidized in air for varying periods of time. After oxidizing in air at a constant temperature for the prescribed time and then being allowed to cool, the samples would be reweighed to determine their weight gain due to the uptake of oxygen. From these weight gain measurements, it was possible to identify three regimes of oxidation for Inconel 718: a low-temperature regime in which the samples became passivated after the initial oxidation, an intermediate-temperature regime in which the rate of oxidation was limited by diffusion and exhibited a constant parabolic rate dependence, and a high-temperature regime in which material deformation and damage accompanied an accelerated oxidation rate above the parabolic regime. At temperatures below 1173 K, the rate of oxidation of the Inconel 718 surface was found to decrease markedly with time; the parabolic oxidation rate coefficient was not a constant but decreased with time. This was taken to indicate that the oxide film on the surface was having a passivating effect on oxygen transport through the oxide to the underlying metal. For temperatures in the range 1173 K to 1573 K, the time-dependent rate of oxidation as determined once again by weight-gain measurements was found to display the classical parabolic rate behavior, indicating that the rate of transport of reactants through the oxide was controlled by diffusion through the growing oxide layer. Parabolic rate coefficients were determined by least-squares analysis of time-dependent mass-gain data at 1173 K, 1273 K, 1373 K, 1473 K and 1573 K. At temperatures above 1540 K, post test examination of the oxidized samples revealed that the Inconel 718 began to lose strength and to deform. At 1540 K, samples which were suspended from their ends during testing began to demonstrate axial curvature as they lost strength and bowed under their own weight. As the temperatures of the tests were increased, rivulets were seen to appear on the surfaces of the test specimens; damage became severe at 1560 K. Although melting was never observed in any of these tests even up to. 1620 K, it was concluded from these data that the Inconel 718 clad should not be expected to protect the underlying tungsten at temperatures above 1540 K.« less

Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoover, M.D.; Farrell, R.F.; Newton, G.J.

1995-12-01

The recent 1995 WIPP Safety Analysis Report (SAR) Update provided detailed analyses of potential radiation doses to members of the public at the site boundary during postulated accident scenarios at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). The SAR Update addressed the complete spectrum of potential accidents associated with handling and emplacing transuranic waste at WIPP, including damage to waste drums from fires, punctures, drops, and other disruptions. The report focused on the adequacy of the multiple layers of safety practice ({open_quotes}defense-in-depth{close_quotes}) at WIPP, which are designed to (1) reduce the likelihood of accidents and (2) limitmore » the consequences of those accidents. The safeguards which contribute to defense-in-depth at WIPP include a substantial array of inherent design features, engineered controls, and administrative procedures. The SAR Update confirmed that the defense-in-depth at WIPP is adequate to assure the protection of the public and environment. As a supplement to the 1995 SAR Update, we have conducted additional analyses to confirm that these controls will also provide adequate protection to workers at the WIPP. The approaches and results of the worker dose assessment are summarized here. In conformance with the guidance of DOE Standard 3009-94, we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposures under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR Update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, members of the public, and the environment.« less

10 CFR 963.13 - Preclosure suitability evaluation method.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to evaluate whether the geologic repository is likely to comply with the applicable radiation... prevent or mitigate the effects of postulated Category 1 and 2 event sequences. The preclosure safety... prevent accidents. ...

75 FR 42170 - Final Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-20

..., techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the... (EDEX) radiation exposures. These methods provide a conservative estimate of the EDEX and may be used to...

The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Guentay, S.; Andreani, M.

2012-07-01

During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). Themore » tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)« less

75 FR 9627 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-03

... staff uses in evaluating specific problems or postulated accidents, and data that the staff needs in its... qualification testing of a prototype unit under the most adverse design conditions. II. Further Information In...

Development and Validation of Accident Models for FeCrAl Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to present the work completed in regards to material model development for FeCrAl cladding and highlight the results of applying these models to Loss of Coolant Accidents (LOCA) and Station Blackouts (SBO). With the limited experimental data available (essentially only the data used to create the models) true validation is not possible. In the absence of another alternative, qualitative comparisons during postulated accident scenarios between FeCrAl and Zircaloy-4 cladded rods have been completed demonstrating the superior performance of FeCrAl.

Three dimensional contact/impact methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

Preliminary assessment of the health and environmental impacts of continuing to store M55 rockets at Lexington-Blue Grass Depot Activity, Anniston Army Depot, Umatilla Depot Activity, Pine Bluff Arsenal, and Tooele Army Depot

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyette, J.A.; Breck, J.E.; Coleman, P.R.

1986-03-01

The purpose is to provide an assessment of the potential health and environmental impacts of continuing to store M55 rockets filled with nerve agent GB or VX at their current storage locations at Anniston Army Depot in Alabama, Lexington-Blue Grass Depot Activity in Kentucky, Pine Bluff Arsenal in Arkansas, Tooele Army Depot in Utah, and Umatilla Depot Activity in Oregon. The assessment considers the possible impacts of (1) normal storage (with no release to the environment) and (2) two postulated accidents on the air quality, ground and surface water, aquatic ecology, terrestrial ecology, human health, and cultural and socioeconomic resourcesmore » in and around the various storage depots. The analysis considers three basic scenarios during storage: (1) normal operations; (2) a minor spill of agent (the contents of one rocket released to the biosphere); and (3) a maximum credible event or MCE. The MCE is an igloo fire resulting in the aerosolization of a small (in the case of GB) or an extremely small (in the case of VX) percentage of the igloo's nerve agent contents to the biosphere. The extremely low probabilities of such accidents, which are reported elsewhere, are noted. Our assessments of the impacts of a minor spill and of an MCE consider two sets of meteorological conditions: conservative most likely and worst-case. In addition, we assume that an agent plume would travel toward the area of highest population density. 21 figs., 47 tabs.« less

Major Railroad Accidents Involving Hazardous Materials Release, Composite Summaries 1969-1978

DOT National Transportation Integrated Search

1980-07-31

This report presents composite summaries describing 75 major railroad accidents in which hazardous materials were released. The selected accidents occurred during the years 1969-1978. The data contained in the individual summaries were derived from v...

Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures.

PubMed

Dickson, E D; Hamby, D M

2014-03-01

The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building's radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology.

Hot Cell Installation and Demonstration of the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« less

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

76 FR 66333 - Carolina Power & Light Company, H.B. Robinson Steam Electric Plant, Unit No. 2; Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-26

..., and hydrogen generation after a postulated loss-of-coolant accident. Therefore, both of these... quality. There are no impacts to historical and cultural resources. In addition, there are also no known...

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. Copyright © 2010 Elsevier B.V. All rights reserved.

Dose Calculation For Accidental Release Of Radioactive Cloud Passing Over Jeddah

NASA Astrophysics Data System (ADS)

Alharbi, N. D.; Mayhoub, A. B.

2011-12-01

For the evaluation of doses after the reactor accident, in particular for the inhalation dose, a thorough knowledge of the concentration of the various radionuclide in air during the passage of the plume is required. In this paper we present an application of the Gaussian Plume Model (GPM) to calculate the atmospheric dispersion and airborne radionuclide concentration resulting from radioactive cloud over the city of Jeddah (KSA). The radioactive cloud is assumed to be emitted from a reactor of 10 MW power in postulated accidental release. Committed effective doses (CEDs) to the public at different distance from the source to the receptor are calculated. The calculations were based on meteorological condition and data of the Jeddah site. These data are: pasquill atmospheric stability is the class B and the wind speed is 2.4m/s at 10m height in the N direction. The residence time of some radionuclides considered in this study were calculated. The results indicate that, the values of doses first increase with distance, reach a maximum value and then gradually decrease. The total dose received by human is estimated by using the estimated values of residence time of each radioactive pollutant at different distances.

Release of plutonium isotopes into the environment from the Fukushima Daiichi Nuclear Power Plant accident: what is known and what needs to be known.

PubMed

Zheng, Jian; Tagami, Keiko; Uchida, Shigeo

2013-09-03

The Fukushima Daiichi Nuclear Power Plant (FDNPP) accident has caused serious contamination in the environment. The release of Pu isotopes renewed considerable public concern because they present a large risk for internal radiation exposure. In this Critical Review, we summarize and analyze published studies related to the release of Pu from the FDNPP accident based on environmental sample analyses and the ORIGEN model simulations. Our analysis emphasizes the environmental distribution of released Pu isotopes, information on Pu isotopic composition for source identification of Pu releases in the FDNPP-damaged reactors or spent fuel pools, and estimation of the amounts of Pu isotopes released from the FDNPP accident. Our analysis indicates that a trace amount of Pu isotopes (∼2 × 10(-5)% of core inventory) was released into the environment from the damaged reactors but not from the spent fuel pools located in the reactor buildings. Regarding the possible Pu contamination in the marine environment, limited studies suggest that no extra Pu input from the FDNPP accident could be detected in the western North Pacific 30 km off the Fukushima coast. Finally, we identified knowledge gaps remained on the release of Pu into the environment and recommended issues for future studies.

76 FR 28102 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-13

..., Probabilistic Risk Assessment Branch, Division of Risk Analysis, Office of Nuclear Regulatory Research, U.S... approaches and methods (whether quantitative or qualitative, deterministic or probabilistic), data, and... uses in evaluating specific problems or postulated accidents, and data that the staff needs in its...

75 FR 61530 - Issuance of Regulatory Guides

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-05

... Materials Code Case Acceptability, ASME Section III,'' and RG 1.147, Rev. 16, ``Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1.'' FOR FURTHER INFORMATION CONTACT: Wallace E. Norris... specific problems or postulated accidents, and data the staff needs in its review of applications for...

76 FR 25725 - Final Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-05

....18, Revision 2, ``Information Relevant to Ensuring that Occupational Radiation Exposures at Medical..., techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the..., ``Information Relevant to Ensuring that Occupational Radiation Exposures at Medical Institutions will be As Low...

Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Weber, C.F.; Hodge, S.A.

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 ontomore » the drywell floor.« less

Westinghouse Small Modular Reactor passive safety system response to postulated events

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. C.; Wright, R. F.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. Themore » integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)« less

Progress in Solving the Elusive Ag Transport Mechanism in TRISO Coated Particles: What is new?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella Van Rooyen

2014-10-01

The TRISO particle for HTRs has been developed to an advanced state where the coating withstands internal gas pressures and retains fission products during irradiation and under postulated accidents. However, one exception is Ag that has been found to be released from high quality TRISO coated particles when irradiated and can also during high temperature accident heating tests. Although out- of- pile laboratory tests have never hither to been able to demonstrate a diffusion process of Ag in SiC, effective diffusion coefficients have been derived to successfully reproduce measured Ag-110m releases from irradiated HTR fuel elements, compacts and TRISO particlesmore » It was found that silver transport through SiC does not proceed via bulk volume diffusion. Presently grain boundary diffusion that may be irradiation enhanced either by neutron bombardment or by the presence of fission products such as Pd, are being investigated. Recent studies of irradiated AGR-1 TRISO fuel using scanning transmission electron microscopy (STEM), transmission kukuchi diffraction (TKD) patterns and high resolution transmission electron microscopy (HRTEM) have been used to further the understanding of Ag transport through TRISO particles. No silver was observed in SiC grains, but Ag was identified at triple-points and grain boundaries of the SiC layer in the TRISO particle. Cadmium was also found in some of the very same triple junctions, but this could be related to silver behavior as Ag-110m decays to Cd-110. Palladium was identified as the main constituent of micron-sized precipitates present at the SiC grain boundaries and in most SiC grain boundaries and the potential role of Pd in the transport of Ag will be discussed.« less

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

75 FR 5630 - Draft Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-03

... specific parts of the NRC's regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff needs in its review of applications for permits and licenses... in license applications and for monitoring and reporting effluent data by licensees. The guidance is...

75 FR 27599 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

..., Revision 2, ``Preparation of an Environmental Report To Support a Rulemaking Petition Seeking an Exemption... staff uses in evaluating specific problems or postulated accidents, and data that the staff needs in its... preparation of environmental reports (ERs), that are submitted to support a rulemaking petition for an...

75 FR 52999 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-30

... or postulated accidents, and data that the staff needs in its review of applications for permits and... Regulations (10 CFR) Part 20, ``Standards for Protection Against Radiation,'' the associated specific... (i.e. radiation doses in addition to and accounted for separately from the doses received under the...

Analysis of multiple tank car releases in train accidents.

PubMed

Liu, Xiang; Liu, Chang; Hong, Yili

2017-10-01

There are annually over two million carloads of hazardous materials transported by rail in the United States. The American railroads use large blocks of tank cars to transport petroleum crude oil and other flammable liquids from production to consumption sites. Being different from roadway transport of hazardous materials, a train accident can potentially result in the derailment and release of multiple tank cars, which may result in significant consequences. The prior literature predominantly assumes that the occurrence of multiple tank car releases in a train accident is a series of independent Bernoulli processes, and thus uses the binomial distribution to estimate the total number of tank car releases given the number of tank cars derailing or damaged. This paper shows that the traditional binomial model can incorrectly estimate multiple tank car release probability by magnitudes in certain circumstances, thereby significantly affecting railroad safety and risk analysis. To bridge this knowledge gap, this paper proposes a novel, alternative Correlated Binomial (CB) model that accounts for the possible correlations of multiple tank car releases in the same train. We test three distinct correlation structures in the CB model, and find that they all outperform the conventional binomial model based on empirical tank car accident data. The analysis shows that considering tank car release correlations would result in a significantly improved fit of the empirical data than otherwise. Consequently, it is prudent to consider alternative modeling techniques when analyzing the probability of multiple tank car releases in railroad accidents. Copyright © 2017 Elsevier Ltd. All rights reserved.

Predictions of structural integrity of steam generator tubes under normal operating, accident, an severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Majumdar, S.

1997-02-01

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation was confirmed by further tests at high temperatures, as well as by finite-element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation was confirmedmore » by finite-element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate-sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure was developed and validated by tests under various temperature and pressure loadings that can occur during postulated severe accidents.« less

Isotopic evidence of plutonium release into the environment from the Fukushima DNPP accident

PubMed Central

Zheng, Jian; Tagami, Keiko; Watanabe, Yoshito; Uchida, Shigeo; Aono, Tatsuo; Ishii, Nobuyoshi; Yoshida, Satoshi; Kubota, Yoshihisa; Fuma, Shoichi; Ihara, Sadao

2012-01-01

The Fukushima Daiichi nuclear power plant (DNPP) accident caused massive releases of radioactivity into the environment. The released highly volatile fission products, such as 129mTe, 131I, 134Cs, 136Cs and 137Cs were found to be widely distributed in Fukushima and its adjacent prefectures in eastern Japan. However, the release of non-volatile actinides, in particular, Pu isotopes remains uncertain almost one year after the accident. Here we report the isotopic evidence for the release of Pu into the atmosphere and deposition on the ground in northwest and south of the Fukushima DNPP in the 20–30 km zones. The high activity ratio of 241Pu/239+240Pu (> 100) from the Fukushima DNPP accident highlights the need for long-term 241Pu dose assessment, and the ingrowth of 241Am. The results are important for the estimation of reactor damage and have significant implication in the strategy of decontamination. PMID:22403743

Toxic release consequence analysis tool (TORCAT) for inherently safer design plant.

PubMed

Shariff, Azmi Mohd; Zaini, Dzulkarnain

2010-10-15

Many major accidents due to toxic release in the past have caused many fatalities such as the tragedy of MIC release in Bhopal, India (1984). One of the approaches is to use inherently safer design technique that utilizes inherent safety principle to eliminate or minimize accidents rather than to control the hazard. This technique is best implemented in preliminary design stage where the consequence of toxic release can be evaluated and necessary design improvements can be implemented to eliminate or minimize the accidents to as low as reasonably practicable (ALARP) without resorting to costly protective system. However, currently there is no commercial tool available that has such capability. This paper reports on the preliminary findings on the development of a prototype tool for consequence analysis and design improvement via inherent safety principle by utilizing an integrated process design simulator with toxic release consequence analysis model. The consequence analysis based on the worst-case scenarios during process flowsheeting stage were conducted as case studies. The preliminary finding shows that toxic release consequences analysis tool (TORCAT) has capability to eliminate or minimize the potential toxic release accidents by adopting the inherent safety principle early in preliminary design stage. 2010 Elsevier B.V. All rights reserved.

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Hazmat transport: a methodological framework for the risk analysis of marshalling yards.

PubMed

Cozzani, Valerio; Bonvicini, Sarah; Spadoni, Gigliola; Zanelli, Severino

2007-08-17

A methodological framework was outlined for the comprehensive risk assessment of marshalling yards in the context of quantified area risk analysis. Three accident typologies were considered for yards: (i) "in-transit-accident-induced" releases; (ii) "shunting-accident-induced" spills; and (iii) "non-accident-induced" leaks. A specific methodology was developed for the assessment of expected release frequencies and equivalent release diameters, based on the application of HazOp and Fault Tree techniques to reference schemes defined for the more common types of railcar vessels used for "hazmat" transportation. The approach was applied to the assessment of an extended case-study. The results evidenced that "non-accident-induced" leaks in marshalling yards represent an important contribution to the overall risk associated to these zones. Furthermore, the results confirmed the considerable role of these fixed installations to the overall risk associated to "hazmat" transportation.

75 FR 54918 - Draft Regulatory Guide, DG-1247, “Design-Basis Hurricane and Hurricane Missiles for Nuclear Power...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-09

... that the staff uses in evaluating specific problems or postulated accidents, and data that the staff... turbine missiles. NUREG/CR 7004 is the technical basis for regulatory guidance on design-basis hurricane... hurricane wind speeds for new nuclear power plants. [[Page 54919

Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1

PubMed Central

Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

2015-01-01

In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by 137Cesium (137Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as 132Te-132I, 131I, 134Cs and 137Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h−1 per initial 137Cs deposition of 1000 kBq m−2, whereas it was 100 μGy h−1 around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m−2 for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums (134Cs + 137Cs) around Chernobyl and Fukushima-1, respectively. PMID:26568603

40 CFR 68.42 - Five-year accident history.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 16 2014-07-01 2014-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases from...

40 CFR 68.42 - Five-year accident history.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 15 2011-07-01 2011-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases from...

40 CFR 68.42 - Five-year accident history.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 16 2013-07-01 2013-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases from...

40 CFR 68.42 - Five-year accident history.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 16 2012-07-01 2012-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases from...

40 CFR 68.42 - Five-year accident history.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases from...

PROBABILISTIC SAFETY ASSESSMENT OF OPERATIONAL ACCIDENTS AT THE WASTE ISOLATION PILOT PLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rucker, D.F.

2000-09-01

This report presents a probabilistic safety assessment of radioactive doses as consequences from accident scenarios to complement the deterministic assessment presented in the Waste Isolation Pilot Plant (WIPP) Safety Analysis Report (SAR). The International Council of Radiation Protection (ICRP) recommends both assessments be conducted to ensure that ''an adequate level of safety has been achieved and that no major contributors to risk are overlooked'' (ICRP 1993). To that end, the probabilistic assessment for the WIPP accident scenarios addresses the wide range of assumptions, e.g. the range of values representing the radioactive source of an accident, that could possibly have beenmore » overlooked by the SAR. Routine releases of radionuclides from the WIPP repository to the environment during the waste emplacement operations are expected to be essentially zero. In contrast, potential accidental releases from postulated accident scenarios during waste handling and emplacement could be substantial, which necessitates the need for radiological air monitoring and confinement barriers (DOE 1999). The WIPP Safety Analysis Report (SAR) calculated doses from accidental releases to the on-site (at 100 m from the source) and off-site (at the Exclusive Use Boundary and Site Boundary) public by a deterministic approach. This approach, as demonstrated in the SAR, uses single-point values of key parameters to assess the 50-year, whole-body committed effective dose equivalent (CEDE). The basic assumptions used in the SAR to formulate the CEDE are retained for this report's probabilistic assessment. However, for the probabilistic assessment, single-point parameter values were replaced with probability density functions (PDF) and were sampled over an expected range. Monte Carlo simulations were run, in which 10,000 iterations were performed by randomly selecting one value for each parameter and calculating the dose. Statistical information was then derived from the 10,000 iteration batch, which included 5%, 50%, and 95% dose likelihood, and the sensitivity of each assumption to the calculated doses. As one would intuitively expect, the doses from the probabilistic assessment for most scenarios were found to be much less than the deterministic assessment. The lower dose of the probabilistic assessment can be attributed to a ''smearing'' of values from the high and low end of the PDF spectrum of the various input parameters. The analysis also found a potential weakness in the deterministic analysis used in the SAR, a detail on drum loading was not taken into consideration. Waste emplacement operations thus far have handled drums from each shipment as a single unit, i.e. drums from each shipment are kept together. Shipments typically come from a single waste stream, and therefore the curie loading of each drum can be considered nearly identical to that of its neighbor. Calculations show that if there are large numbers of drums used in the accident scenario assessment, e.g. 28 drums in the waste hoist failure scenario (CH5), then the probabilistic dose assessment calculations will diverge from the deterministically determined doses. As it is currently calculated, the deterministic dose assessment assumes one drum loaded to the maximum allowable (80 PE-Ci), and the remaining are 10% of the maximum. The effective average of drum curie content is therefore less in the deterministic assessment than the probabilistic assessment for a large number of drums. EEG recommends that the WIPP SAR calculations be revisited and updated to include a probabilistic safety assessment.« less

Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

NASA Astrophysics Data System (ADS)

Barber, Duncan Henry

During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A benchmark calculation demonstrates the improvement in agreement of the total inventory of those chemical elements included in the RMC fuel model to an ORIGEN-S calculation. ORIGEN-S is the Oak Ridge isotope generation and depletion computer program. The Gibbs energy minimizer requires a chemical database containing coefficients from which the Gibbs energy of pure compounds, gas and liquid mixtures, and solid solutions can be calculated. The RMC model of irradiated uranium dioxide fuel has been converted into the required format. The Gibbs energy minimizer has been incorporated into a new model of fission-product vaporization from the fuel surface. Calculated release fractions using the new code have been compared to results calculated with SOURCE IST 2.0P11 and to results of tests used in the validation of SOURCE 2.0. The new code shows improvements in agreement with experimental releases for a number of nuclides. Of particular significance is the better agreement between experimental and calculated release fractions for 140La. The improved agreement reflects the inclusion in the RMC model of the solubility of lanthanum (III) oxide (La2O3) in the fuel matrix. Calculated lanthanide release fractions from earlier computer programs were a challenge to environmental qualification analysis of equipment for some accident scenarios. The new prototype computer program would alleviate this concern. Keywords: Nuclear Engineering; Material Science; Thermodynamics; Radioactive Material, Gibbs Energy Minimization, Actinide Generation and Depletion, FissionProduct Generation and Depletion.

Probabilistic Approach to Conditional Probability of Release of Hazardous Materials from Railroad Tank Cars during Accidents

DOT National Transportation Integrated Search

2009-10-13

This paper describes a probabilistic approach to estimate the conditional probability of release of hazardous materials from railroad tank cars during train accidents. Monte Carlo methods are used in developing a probabilistic model to simulate head ...

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R J; Gottula, R C; Holcomb, E E

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

Site restoration: Estimation of attributable costs from plutonium-dispersal accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chanin, D.I.; Murfin, W.B.

1996-05-01

A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occurmore » in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U 3Si 2 Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, K. A.; Hales, J. D.; Miao, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into leadmore » test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

Loss of control air at Browns Ferry Unit One: accident sequence analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.; Hodge, S.A.

1986-04-01

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure injection systems. Accident sequence calculations are presented for Loss of Control Air sequences with assumed failure upon demand of the Reactor Core Isolation Cooling (RCIC) and the High Pressure Coolant Injection (HPCI) at Unit 1. Sequences with and without operator action are considered.more » Results show that the operators can prevent core uncovery if they take action to utilize the Control Rod Drive Hydraulic System as a backup high pressure injection system.« less

ATWS at Browns Ferry Unit One - accident sequence analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.; Hodge, S.A.

1984-07-01

This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence themore » quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.« less

49 CFR 831.13 - Flow and dissemination of accident or incident information.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 7 2010-10-01 2010-10-01 false Flow and dissemination of accident or incident...) NATIONAL TRANSPORTATION SAFETY BOARD ACCIDENT/INCIDENT INVESTIGATION PROCEDURES § 831.13 Flow and dissemination of accident or incident information. (a) Release of information during the field investigation...

77 FR 69509 - Combining Modal Responses and Spatial Components in Seismic Response Analysis

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-19

... available documents online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the.... ML122020A044. NRC's Public Document Room: You may examine and purchase copies of public documents at the NRC's... postulated accidents, and data that the staff needs in its review of applications for permits and licenses...

77 FR 11598 - Thermal Overload Protection for Electric Motors on Motor-Operated Valves

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-27

...-available documents online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the... may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint... or postulated accidents, and data that the staff needs in its review of applications for permits and...

Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1.

PubMed

Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

2015-12-01

In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by (137)Cesium ((137)Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as (132)Te-(132)I, (131)I, (134)Cs and (137)Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h(-1) per initial (137)Cs deposition of 1000 kBq m(-2), whereas it was 100 μGy h(-1) around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m(-2) for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums ((134)Cs + (137)Cs) around Chernobyl and Fukushima-1, respectively. © The Author 2015. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Ecological and toxicological aspects of the partial meltdown of the Chernobyl nuclear power plant reactor

USGS Publications Warehouse

Eisler, Ronald; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

1995-01-01

the partial meltdown of the 1000-MW reactor at Chernobyl, Ukraine, on April 26, 1986, released large amounts of radiocesium and other radionuclides into the environment, causing widespread radioactive contamination of Europe and the former Soviet Union.1-7 At least 3,000,000 trillion becquerels (TBq) were released from the fuel during the accident (Table 24.1), dwarfing, by orders of magnitude, radiation released from other highly publicized reactor accidents at Windscale (U.K.) and three-Mile Island (U.S.)3,8 The Chernobyl accident happened while a test was being conducted during a normal scheduled shutdown and is attributed mainly to human error.3

Estimations of direct release rate of 137Cs and 90Sr to the ocean from the Fukushima Dai-ichi Nuclear Power Plant for five-and-a-half years

NASA Astrophysics Data System (ADS)

Tsumune, Daisuke; Aoyama, Michio; Tsubono, Takaki; Misumi, Kazuhiro; Tateda, Yutaka

2017-04-01

A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant (1F NPP) following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition. Additional release pathways by river input and runoff from 1F NPP site with precipitation and were also effective for coastal zone in the specific periods before starting direct release on March 26 2011. Direct release from 1F NPP site is dominant one year after the accident. We estimated the direct release rate of 137Cs and 90Sr for more than five-and-a-half years after the accident by the Regional Ocean Model System (ROMS). Direct release rate of 137Cs were estimated for five-and-a-half years after the accident by comparing simulated results and measured activities adjacent to the 1F NPP site(at 5,6 discharge and south discharge). Directly release rate of 137Cs was estimated to be the order of magnitude of 1014 Bq/day and decreased exponentially with time to be the order of magnitude of 109 Bq/day by the end of September 2016. Estimated direct release rate have exponentially reduced with constant rate since November 2011. Apparent half-life of direct release rate was estimated to be 346 days. The estimated total amounts of directly released 137Cs was 3.7±0.7 PBq for five and a half years. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable. Simulated 137Cs activity affected off coast in the Fukushima prefecture. We used the measured 137Cs activities by the Tokyo Electric Power Company (TEPCO) for the estimation of direct release. The sea water samples were corrected from the coast. The averaged 137Cs activities from November 2013 to June 2016 were 391 and 383 Bq/m3 at 5,6 discharge and south discharge, respectively. The averaged 137Cs activities measured by the Nuclear Regulation Agency (NRA) is about five times smaller than the one by the TEPCO because the NRA corrected seawater samples at 300-500m offshore by ship. Horizontal resolution of the model was 1km x 1km, therefore it is important to consider the difference of activities in the sub-grid scale for the detailed estimations of direct release. 90Sr/137Cs activity ratio measured adjacent to the 1F NPP is variable with time. The 90Sr/137Cs activity ratio was 0.62 due to the global fallout before the accident. The 90Sr/137Cs activity ratio decreased to 0.01 after the accident before April 2011. And the ratio increased to 1 by September 2013. And then the ratio decreased to 0.1-1. After October 2015, the ratio decreased to 0.1-0.2. Directly release rate of 90Sr was estimated to be the order of magnitude of 1012 Bq/day and decreased to the order of magnitude of 108 Bq/day by the end of September 2016. The estimated total amounts of directly released 90Sr was 35 ± 7 TBq.

Using Microsoft Excel to compute the 5% overall site X/Q value and the 95th percentile of the distribution of doses to the nearest maximally exposed offsite individual (MEOI).

PubMed

Vickers, Linda D

2010-05-01

This paper describes the method using Microsoft Excel (Microsoft Corporation One Microsoft Way Redmond, WA 98052-6399) to compute the 5% overall site X/Q value and the 95th percentile of the distribution of doses to the nearest maximally exposed offsite individual (MEOI) in accordance with guidance from DOE-STD-3009-1994 and U.S. NRC Regulatory Guide 1.145-1982. The accurate determination of the 5% overall site X/Q value is the most important factor in the computation of the 95th percentile of the distribution of doses to the nearest MEOI. This method should be used to validate software codes that compute the X/Q. The 95th percentile of the distribution of doses to the nearest MEOI must be compared to the U.S. DOE Evaluation Guide of 25 rem to determine the relative severity of hazard to the public from a postulated, unmitigated design basis accident that involves an offsite release of radioactive material.

Final safety analysis report for the Galileo mission: Volume 3 (Book 2), Nuclear risk analysis document: Appendices: Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1989-01-25

It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequence analyses can be used in several ways. One way is to identify the potential range of consequences which might have to be dealt with ifmore » there were to be an accident with a release of fuel, so as to assure that, given such an accident, the health and safety of the public will be reasonably protected. Another use of the information, in conjunction with accident and release probabilities, is to estimate the risks associated with the mission. That is, most space launches occur without incident. Given an accident, the most probable result relative to the RTGs is complete containment of the radioactive material. Only a small fraction of accidents might result in a release of fuel and subsequent radiological consequences. The combination of probability with consequence is risk, which can be compared to other human and societal risks to assure that no undue risks are implied by undertaking the mission. Book 2 contains eight appendices.« less

Safety analysis report for the Galileo Mission. Volume 3, book 2: Nuclear risk analysis document. Appendices, revision 1

NASA Astrophysics Data System (ADS)

1989-01-01

It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequence analyses can be used in several ways. One way is to identify the potential range of consequences which might have to be dealt with if there were to be an accident with a release of fuel, so as to assure that, given such an accident, the health and safety of the public will be reasonably protected. Another use of the information, in conjunction with accident and release probabilities, is to estimate the risks associated with the mission. That is, most space launches occur without incident. Given an accident, the most probable result relative to the RTGs is complete containment of the radioactive material. Only a small fraction of accidents might result in a release of fuel and subsequent radiological consequences. The combination of probability with consequence is risk, which can be compared to other human and societal risks to assure that no undue risks are implied by undertaking the mission. Book 2 contains eight appendices.

Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

DOE PAGES

Di Maio, Francesco; Zio, Enrico; Smith, Curtis; ...

2015-07-06

The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs andmore » activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

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

PubMed

Steinhauser, Georg; Brandl, Alexander; Johnson, Thomas E

2014-02-01

The environmental impacts of the nuclear accidents of Chernobyl and Fukushima are compared. In almost every respect, the consequences of the Chernobyl accident clearly exceeded those of the Fukushima accident. In both accidents, most of the radioactivity released was due to volatile radionuclides (noble gases, iodine, cesium, tellurium). However, the amount of refractory elements (including actinides) emitted in the course of the Chernobyl accident was approximately four orders of magnitude higher than during the Fukushima accident. For Chernobyl, a total release of 5,300 PBq (excluding noble gases) has been established as the most cited source term. For Fukushima, we estimated a total source term of 520 (340-800) PBq. In the course of the Fukushima accident, the majority of the radionuclides (more than 80%) was transported offshore and deposited in the Pacific Ocean. Monitoring campaigns after both accidents reveal that the environmental impact of the Chernobyl accident was much greater than of the Fukushima accident. Both the highly contaminated areas and the evacuated areas are smaller around Fukushima and the projected health effects in Japan are significantly lower than after the Chernobyl accident. This is mainly due to the fact that food safety campaigns and evacuations worked quickly and efficiently after the Fukushima accident. In contrast to Chernobyl, no fatalities due to acute radiation effects occurred in Fukushima. © 2013.

Analysis of Radionuclide Releases from the Fukushima Dai-ichi Nuclear Power Plant Accident Part II

NASA Astrophysics Data System (ADS)

Achim, Pascal; Monfort, Marguerite; Le Petit, Gilbert; Gross, Philippe; Douysset, Guilhem; Taffary, Thomas; Blanchard, Xavier; Moulin, Christophe

2014-03-01

The present part of the publication (Part II) deals with long range dispersion of radionuclides emitted into the atmosphere during the Fukushima Dai-ichi accident that occurred after the March 11, 2011 tsunami. The first part (Part I) is dedicated to the accident features relying on radionuclide detections performed by monitoring stations of the Comprehensive Nuclear Test Ban Treaty Organization network. In this study, the emissions of the three fission products Cs-137, I-131 and Xe-133 are investigated. Regarding Xe-133, the total release is estimated to be of the order of 6 × 1018 Bq emitted during the explosions of units 1, 2 and 3. The total source term estimated gives a fraction of core inventory of about 8 × 1018 Bq at the time of reactors shutdown. This result suggests that at least 80 % of the core inventory has been released into the atmosphere and indicates a broad meltdown of reactor cores. Total atmospheric releases of Cs-137 and I-131 aerosols are estimated to be 1016 and 1017 Bq, respectively. By neglecting gas/particulate conversion phenomena, the total release of I-131 (gas + aerosol) could be estimated to be 4 × 1017 Bq. Atmospheric transport simulations suggest that the main air emissions have occurred during the events of March 14, 2011 (UTC) and that no major release occurred after March 23. The radioactivity emitted into the atmosphere could represent 10 % of the Chernobyl accident releases for I-131 and Cs-137.

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 traditionalmore » 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.« less

Estimating probable flaw distributions in PWR steam generator tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorman, J.A.; Turner, A.P.L.

1997-02-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regardingmore » uncertainties and assumptions in the data and analyses.« less

Learning lessons from Natech accidents - the eNATECH accident database

NASA Astrophysics Data System (ADS)

Krausmann, Elisabeth; Girgin, Serkan

2016-04-01

When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of equipment vulnerability models linking the natural-hazard severity to the observed damage almost impossible. As a consequence, the European Commission has set up the eNATECH database for the systematic collection of Natech accident data and near misses. The database exhibits the more sophisticated accident representation required to capture the characteristics of Natech events and is publicly accessible at http://enatech.jrc.ec.europa.eu. This presentation outlines the general lessons-learning process, introduces the eNATECH database and its specific structure, and discusses natural-hazard specific lessons learned and features common to Natech accidents triggered by different natural hazards.

Assessment of the risk due to release of carbon fiber in civil aircraft accidents, phase 2

NASA Technical Reports Server (NTRS)

Pocinki, L.; Cornell, M. E.; Kaplan, L.

1980-01-01

The risk associated with the potential use of carbon fiber composite material in commercial jet aircraft is investigated. A simulation model developed to generate risk profiles for several airports is described. The risk profiles show the probability that the cost due to accidents in any year exceeds a given amount. The computer model simulates aircraft accidents with fire, release of fibers, their downwind transport and infiltration of buildings, equipment failures, and resulting ecomomic impact. The individual airport results were combined to yield the national risk profile.

Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abrecht, David G.; Schwantes, Jon M.

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔG rxn°(T C))/(RT C)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG° rxn(T C). These models allowedmore » an estimate of the upper bound for the reactor temperatures of T C between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.« less

Monte Carlo simulation of single accident airport risk profile

NASA Technical Reports Server (NTRS)

1979-01-01

A computer simulation model was developed for estimating the potential economic impacts of a carbon fiber release upon facilities within an 80 kilometer radius of a major airport. The model simulated the possible range of release conditions and the resulting dispersion of the carbon fibers. Each iteration of the model generated a specific release scenario, which would cause a specific amount of dollar loss to the surrounding community. By repeated iterations, a risk profile was generated, showing the probability distribution of losses from one accident. Using accident probability estimates, the risks profile for annual losses was derived. The mechanics are described of the simulation model, the required input data, and the risk profiles generated for the 26 large hub airports.

78 FR 39781 - Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-02

... Spent Fuel Pool Study). The purpose of this study was to examine if faster removal of older, colder... NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the search, select ``ADAMS Public... of postulated spent fuel pool accidents. The purpose of this study is to examine if faster removal of...

Radiocarbon Releases from the 2011 Fukushima Nuclear Accident

PubMed Central

Xu, Sheng; Cook, Gordon T.; Cresswell, Alan J.; Dunbar, Elaine; Freeman, Stewart P. H. T.; Hou, Xiaolin; Jacobsson, Piotr; Kinch, Helen R.; Naysmith, Philip; Sanderson, David C. W.; Tripney, Brian G.

2016-01-01

Radiocarbon activities were measured in annual tree rings for the years 2009 to 2015 from Japanese cedar trees (Cryptomeria japonica) collected at six sites ranging from 2.5–38 km northwest and north of the Fukushima Dai-ichi nuclear power plant. The 14C specific activity varied from 280.4 Bq kg−1 C in 2010 to 226.0 Bq kg−1 C in 2015. The elevated 14C activities in the 2009 and 2010 rings confirmed 14C discharges during routine reactor operations, whereas those activities that were indistinguishable from background in 2012–2015 coincided with the permanent shutdown of the reactors after the accident in 2011. High-resolution 14C analysis of the 2011 ring indicated 14C releases during the Fukushima accident. The resulted 14C activity decreased with increasing distance from the plant. The maximum 14C activity released during the period of the accident was measured 42.4 Bq kg−1 C above the natural ambient 14C background. Our findings indicate that, unlike other Fukushima-derived radionuclides, the 14C released during the accident is indistinguishable from ambient background beyond the local environment (~30 km from the plant). Furthermore, the resulting dose to the local population from the excess 14C activities is negligible compared to the dose from natural/nuclear weapons sources. PMID:27841312

Radiocarbon Releases from the 2011 Fukushima Nuclear Accident

NASA Astrophysics Data System (ADS)

Xu, Sheng; Cook, Gordon T.; Cresswell, Alan J.; Dunbar, Elaine; Freeman, Stewart P. H. T.; Hou, Xiaolin; Jacobsson, Piotr; Kinch, Helen R.; Naysmith, Philip; Sanderson, David C. W.; Tripney, Brian G.

2016-11-01

Radiocarbon activities were measured in annual tree rings for the years 2009 to 2015 from Japanese cedar trees (Cryptomeria japonica) collected at six sites ranging from 2.5-38 km northwest and north of the Fukushima Dai-ichi nuclear power plant. The 14C specific activity varied from 280.4 Bq kg-1 C in 2010 to 226.0 Bq kg-1 C in 2015. The elevated 14C activities in the 2009 and 2010 rings confirmed 14C discharges during routine reactor operations, whereas those activities that were indistinguishable from background in 2012-2015 coincided with the permanent shutdown of the reactors after the accident in 2011. High-resolution 14C analysis of the 2011 ring indicated 14C releases during the Fukushima accident. The resulted 14C activity decreased with increasing distance from the plant. The maximum 14C activity released during the period of the accident was measured 42.4 Bq kg-1 C above the natural ambient 14C background. Our findings indicate that, unlike other Fukushima-derived radionuclides, the 14C released during the accident is indistinguishable from ambient background beyond the local environment (~30 km from the plant). Furthermore, the resulting dose to the local population from the excess 14C activities is negligible compared to the dose from natural/nuclear weapons sources.

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.

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan C.; Gauntt, Randall O.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of themore » of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.; Mattie, Patrick D.

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 studymore » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek J.; Diamond D.; Cuadra, A.

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 modelmore » 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.« less

Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

PubMed

Abrecht, David G; Schwantes, Jon M

2015-03-03

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

Nuclear fuel in a reactor accident.

PubMed

Burns, Peter C; Ewing, Rodney C; Navrotsky, Alexandra

2012-03-09

Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

Nuclear reactor safety research since three mile island.

PubMed

Mynatt, F R

1982-04-09

The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

Young Driver Attitude Scale: The Development and Field-Testing of an Instrument To Measure Young Driver Risk-Taking Attitudes.

ERIC Educational Resources Information Center

Malfetti, James L.; And Others

It has been postulated that young people are open to driving accidents because of age, lack of experience and risk-taking attitudes. The Young Driver Attitude Scale (YDAS) was developed to measure attitudes likely to influence driving practices. Over 200 items relating to the risk-taking attitudes of young people were shown to a small group of…

Individual dose due to radioactivity accidental release from fusion reactor.

PubMed

Nie, Baojie; Ni, Muyi; Wei, Shiping

2017-04-05

As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1kg HTO and 1000kg dust release) and 34-52years for case 2 (1kg HTO and 10kg-100kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation. Copyright © 2016 Elsevier B.V. All rights reserved.

FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Zawadzki, S.A.

The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified.

40 CFR 68.165 - Offsite consequence analysis.

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.165 Offsite consequence... released in pounds; (7) Release rate; (8) Release duration; (9) Wind speed and atmospheric stability class...

40 CFR 68.165 - Offsite consequence analysis.

Code of Federal Regulations, 2012 CFR

2012-07-01

... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.165 Offsite consequence... released in pounds; (7) Release rate; (8) Release duration; (9) Wind speed and atmospheric stability class...

40 CFR 68.165 - Offsite consequence analysis.

Code of Federal Regulations, 2014 CFR

2014-07-01

... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.165 Offsite consequence... released in pounds; (7) Release rate; (8) Release duration; (9) Wind speed and atmospheric stability class...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2013 CFR

2013-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2014 CFR

2014-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2012 CFR

2012-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

The safety review and approval process for space nuclear power sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, G.L.

1991-01-01

Over the past 30 yr. the U.S. Government has evolved a process for the safety review and launch approval of nuclear power sources (NPSs) proposed for launch into space. This process, which involves a number of governmental agencies, ensures that the various postulated accident scenarios are considered, that the responses of the NPSs to the accident environments are assessed, and that appropriate elements of the Federal Government are involved in the launch approval. This process has worked very well in the successful launches of 37 radioisotope thermoelectric generators and 1 reactor by the United States since 1961. Particular attention willmore » be focused on the recent launch of the Galileo spacecraft. 19 refs., 12 figs., 4 tabs.« less

(135)Cs/(137)Cs isotopic ratio as a new tracer of radiocesium released from the Fukushima nuclear accident.

PubMed

Zheng, Jian; Tagami, Keiko; Bu, Wenting; Uchida, Shigeo; Watanabe, Yoshito; Kubota, Yoshihisa; Fuma, Shoichi; Ihara, Sadao

2014-05-20

Since the Fukushima Daiichi nuclear power plant (FDNPP) accident in 2011, intensive studies of the distribution of released fission products, in particular (134)Cs and (137)Cs, in the environment have been conducted. However, the release sources, that is, the damaged reactors or the spent fuel pools, have not been identified, which resulted in great variation in the estimated amounts of (137)Cs released. Here, we investigated heavily contaminated environmental samples (litter, lichen, and soil) collected from Fukushima forests for the long-lived (135)Cs (half-life of 2 × 10(6) years), which is usually difficult to measure using decay-counting techniques. Using a newly developed triple-quadrupole inductively coupled plasma tandem mass spectrometry method, we analyzed the (135)Cs/(137)Cs isotopic ratio of the FDNPP-released radiocesium in environmental samples. We demonstrated that radiocesium was mainly released from the Unit 2 reactor. Considering the fact that the widely used tracer for the released Fukushima accident-sourced radiocesium in the environment, the (134)Cs/(137)Cs activity ratio, will become unavailable in the near future because of the short half-life of (134)Cs (2.06 years), the (135)Cs/(137)Cs isotopic ratio can be considered as a new tracer for source identification and long-term estimation of the mobility of released radiocesium in the environment.

Nuclear power and probabilistic safety assessment (PSA): past through future applications

NASA Astrophysics Data System (ADS)

Stamatelatos, M. G.; Moieni, P.; Everline, C. J.

1995-03-01

Nuclear power reactor safety in the United States is about to enter a new era -- an era of risk- based management and risk-based regulation. First, there was the age of `prescribed safety assessment,' during which a series of design-basis accidents in eight categories of severity, or classes, were postulated and analyzed. Toward the end of that era, it was recognized that `Class 9,' or `beyond design basis,' accidents would need special attention because of the potentially severe health and financial consequences of these accidents. The accident at Three Mile Island showed that sequences of low-consequence, high-frequency events and human errors can be much more risk dominant than the Class 9 accidents. A different form of safety assessment, PSA, emerged and began to gain ground against the deterministic safety establishment. Eventually, this led to the current regulatory requirements for individual plant examinations (IPEs). The IPEs can serve as a basis for risk-based regulation and management, a concept that may ultimately transform the U.S. regulatory process from its traditional deterministic foundations to a process predicated upon PSA. Beyond the possibility of a regulatory environment predicated upon PSA lies the possibility of using PSA as the foundation for managing daily nuclear power plant operations.

EMERALD REV.1. PWR Accident Activity Release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1975-10-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Nuclear Powerplant Safety: Source Terms. Nuclear Energy.

ERIC Educational Resources Information Center

Department of Energy, Washington, DC. Nuclear Energy Office.

There has been increased public interest in the potential effects of nuclear powerplant accidents since the Soviet reactor accident at Chernobyl. People have begun to look for more information about the amount of radioactivity that might be released into the environment as a result of such an accident. When this issue is discussed by people…

Hybrid and electric low-noise cars cause an increase in traffic accidents involving vulnerable road users in urban areas.

PubMed

Brand, Stephan; Petri, Maximilian; Haas, Philipp; Krettek, Christian; Haasper, Carl

2013-01-01

Due to resource scarcity, the number of low-noise and electric cars is expected to increase rapidly. The frequent use of these cars will lead to a significant reduction of traffic related noise and pollution. On the other hand, due to the adaption and conditioning of vulnerable road users the number of traffic accidents involving pedestrians and bicyclists is postulated to increase as well. Children, older people with reduced eyesight and the blind are especially reliant on a combination of acoustic and visual warning signals with approaching or accelerating vehicles. This is even more evident in urban areas where the engine sound is the dominating sound up to 30 kph (kilometres per hour). Above this, tyre-road interaction is the main cause of traffic noise. With the missing typical engine sound a new sound design is necessary to prevent traffic accidents in urban areas. Drivers should not be able to switch the sound generator off.

Effects of the Chernobyl and Fukushima nuclear accidents on atmospheric electricity parameters recorded at Polish observation stations

NASA Astrophysics Data System (ADS)

Kubicki, Marek; Baranski, Piotr; Odzimek, Anna; Michnowski, Stanislaw; Myslek-Laurikainen, Bogna

2013-04-01

We analyse the atmospheric electricity parameters, measured at Polish geophysical stations in Swider, Poland, and Hornsund, Spitsbergen, in connection with the radioactive incident in Fukushima, Japan, beginning on 11 March 2011, following the 9.0 earthquake and tsunami. We compare our results with the situation during and after the Chernobyl disaster on April 26, 1986, when the radioactive fallout detected at Swider increased in the last week of April 1986, from 4.111 to 238.7 Bq/m2 and up to 967.0 Bq/m2 in the second week of May 1986 - what was more than 235 times greater than the values measured prior to that accident. Besides the electric field especially the electric conductivity is very sensitive to the radioactive contamination of the air. Thus we postulate that these two measurements should be run at geophysical stations over the world and used as a relatively simple and low-cost tool for continuous monitoring of possible hazard caused by nuclear power plant accidents.

Station blackout at Browns Ferry Unit One: iodine and noble-gas distribution and release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Weber, C.F.; Lorenz, R.A.

1982-08-01

This is the second volume of a report describing the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to a postulated Station Blackout, defined as a loss of offsite power combined with failure of all onsite emergency diesel-generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion and the completely powerless state leads to core uncovery, meltdown, reactor vessel failure, and failure of the primary containment by overtemperature-induced degradation of the electrical penetration assembly seals. The sequence of events is described in Volume 1; the material in this volumemore » deals with the analysis of fission product noble gas and iodine transport during the accident. Factors which affect the fission product movements through the series of containment design barriers are reviewed. For a reactive material such as iodine, proper assessment of the rate of movement requires determination of the chemical changes along the pathway which alter the physical properties such as vapor pressure and solubility and thereby affect the transport rate. A methodology for accomplishing this is demonstrated in this report.« less

LARC-1: a Los Alamos release calculation program for fission product transport in HTGRs during the LOFC accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carruthers, L.M.; Lee, C.E.

1976-10-01

The theoretical and numerical data base development of the LARC-1 code is described. Four analytical models of fission product release from an HTGR core during the loss of forced circulation accident are developed. Effects of diffusion, adsorption and evaporation of the metallics and precursors are neglected in this first LARC model. Comparison of the analytic models indicates that the constant release-renormalized model is adequate to describe the processes involved. The numerical data base for release constants, temperature modeling, fission product release rates, coated fuel particle failure fraction and aged coated fuel particle failure fractions is discussed. Analytic fits and graphicmore » displays for these data are given for the Ft. St. Vrain and GASSAR models.« less

Transportation accident scenarios for commercial spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilmot, E L

1981-02-01

A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

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…

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

Flow reversal power limit for the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Uncertainties about the afterheat removal capability during the flow reversal has limited the reactor operating power to 30 MW. An experimental and analytical program to address these uncertainties is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safemore » operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW.« less

Chemical Safety Alert: Catastrophic Failure of Storage Tanks

EPA Pesticide Factsheets

Aboveground, atmospheric storage tanks can fail when flammable vapors in the tank explode and break either the shell-to-bottom or side seam, resulting in hazardous release accidents. Proper maintenance practices can help prevent accidents.

RMP Guidance for Chemical Distributors - Chapter 3: Five-Year Accident History

EPA Pesticide Factsheets

A five year accident history must be completed for each covered process, and all accidental release events meeting specified criteria must be reported in the Risk Management Plan (RMP) for that process.

Evaluation of Equipment Vulnerability and Potential Shock Hazards. [carbon fibers

NASA Technical Reports Server (NTRS)

Taback, I.

1980-01-01

The vulnerability of electric equipment to carbon fibers released from aircraft accidents is investigated and the parameters affecting vulnerability are discussed. The shock hazard for a hypothetical set of accidents is computed.

Enforcement Alert: EPA Enforcement Efforts Focus on Prevention of Chemical Accidents

EPA Pesticide Factsheets

This Alert is intended to inform the industry that companies must take responsibility to prevent accidental releases of dangerous chemicals like anhydrous ammonia through compliance with CAA’s Chemical Accident Prevention Program.

EMERALD REVISION 1; PWR accident activity release. [IBM360,370; FORTRAN IV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fowler, T.B.; Tobias, M.L.; Fox, J.N.

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.IBM360,370; FORTRAN IV; OS/360,370 (IBM360,370); 520K bytes of memory are required..« less

Spatial and temporal distributions of (134)Cs and (137)Cs derived from the TEPCO Fukushima Daiichi Nuclear Power Plant accident in the North Pacific Ocean by using optimal interpolation analysis.

PubMed

Inomata, Y; Aoyama, M; Tsubono, T; Tsumune, D; Hirose, K

2016-01-01

Optimal interpolation (OI) analysis was used to investigate the oceanic distributions of (134)Cs and (137)Cs released from the Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Plant (FNPP1) accident. From the end of March to early April 2011, extremely high activities were observed in the coastal surface seawater near the FNPP1. The high activities spread to a region near 165°E in the western North Pacific Ocean, with a latitudinal center of 40°N. Atmospheric deposition also caused high activities in the region between 180° and 130°W in the North Pacific Ocean. The inventory of FNPP1-released (134)Cs in the North Pacific Ocean was estimated to be 15.3 ± 2.6 PBq. About half of this activity (8.4 ± 2.6 PBq) was found in the coastal region near the FNPP1. After 6 April 2011, when major direct releases ceased, the FNPP1-released (134)Cs in the coastal region decreased exponentially with an apparent half-time of about 4.2 ± 0.5 days and declined to about 2 ± 0.4 PBq by the middle of May 2011. Taking into account that the (134)Cs/(137)Cs activity ratio was about 1 just after release and was extremely uniform during the first month after the accident, the amount of (137)Cs released by the FNPP1 accident increased the North Pacific inventory of (137)Cs due to bomb testing during the 1950s and early 1960s by 20%.

Hydrogen considerations in light-water power reactons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keilholtz, G.W.

1976-02-01

A critical review of the literature now available on hydrogen considerations in light-water power reactors (LWRs) and a bibliography of that literature are presented. The subject matter includes mechanisms for the generation of hydrogen-oxygen mixtures, a description of the fundamental properties of such mixtures, and their spontaneous ignition in both static and dynamic systems. The limits for hydrogen flammability and flame propagation are examined in terms of the effects of pressure, temperature, and additives; the emphasis is on the effects of steam and water vapor. The containment systems for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are compared, and methodsmore » to control hydrogen and oxygen under the conditions of both normal operation and postulated accidents are reviewed. It is concluded that hydrogen can be controlled so that serious complications from the production of hydrogen will not occur. The bibliography contains abstracts from the computerized files of the Nuclear Safety Information Center. Key-word, author, and permuted-title indexes are provided. The bibliography includes responses to questions asked by the U. S. Nuclear Regulatory Commission (NRC) which relate to hydrogen, as well as information on normal operations and postulated accidents including generation of hydrogen from core sprays. Other topics included in the ten sections of the bibliography are metal-water reactions, containment atmosphere, radiolytic gas, and recombiners.« less

After Fukushima: managing the consequences of a radiological release.

PubMed

Fitzgerald, Joe; Wollner, Samuel B; Adalja, Amesh A; Morhard, Ryan; Cicero, Anita; Inglesby, Thomas V

2012-06-01

Even amidst the devastation following the earthquake and tsunami in Japan that killed more than 20,000 people, it was the accident at the Fukushima Daiichi nuclear power plant that led the country's prime minister, Naoto Kan, to fear for "the very existence of the Japanese nation." While accidents that result in mass radiological releases have been rare throughout the operating histories of existing nuclear power plants, the growing number of plants worldwide increases the likelihood that such releases will occur again in the future. Nuclear power is an important source of energy in the U.S. and will be for the foreseeable future. Accidents far smaller in scale than the one in Fukushima could have major societal consequences. Given the extensive, ongoing Nuclear Regulatory Commission (NRC) and industry assessment of nuclear power plant safety and preparedness issues, the Center for Biosecurity of UPMC focused on offsite policies and plans intended to reduce radiation exposure to the public in the aftermath of an accident. This report provides an assessment of Japan's efforts at nuclear consequence management; identifies concerns with current U.S. policies and practices for "outside the fence" management of such an event in the U.S.; and makes recommendations for steps that can be taken to strengthen U.S. government, industry, and community response to large-scale accidents at nuclear power plants.

Management of Ultimate Risk of Nuclear Power Plants by Source Terms - Lessons Learned from the Chernobyl Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Genn Saji

2006-07-01

The term 'ultimate risk' is used here to describe the probabilities and radiological consequences that should be incorporated in siting, containment design and accident management of nuclear power plants for hypothetical accidents. It is closely related with the source terms specified in siting criteria which assures an adequate separation of radioactive inventories of the plants from the public, in the event of a hypothetical and severe accident situation. The author would like to point out that current source terms which are based on the information from the Windscale accident (1957) through TID-14844 are very outdated and do not incorporate lessonsmore » learned from either the Three Miles Island (TMI, 1979) nor Chernobyl accident (1986), two of the most severe accidents ever experienced. As a result of the observations of benign radionuclides released at TMI, the technical community in the US felt that a more realistic evaluation of severe reactor accident source terms was necessary. In this background, the 'source term research project' was organized in 1984 to respond to these challenges. Unfortunately, soon after the time of the final report from this project was released, the Chernobyl accident occurred. Due to the enormous consequences induced by then accident, the one time optimistic perspectives in establishing a more realistic source term were completely shattered. The Chernobyl accident, with its human death toll and dispersion of a large part of the fission fragments inventories into the environment, created a significant degradation in the public's acceptance of nuclear energy throughout the world. In spite of this, nuclear communities have been prudent in responding to the public's anxiety towards the ultimate safety of nuclear plants, since there still remained many unknown points revolving around the mechanism of the Chernobyl accident. In order to resolve some of these mysteries, the author has performed a scoping study of the dispersion and deposition mechanisms of fuel particles and fission fragments during the initial phase of the Chernobyl accident. Through this study, it is now possible to generally reconstruct the radiological consequences by using a dispersion calculation technique, combined with the meteorological data at the time of the accident and land contamination densities of {sup 137}Cs measured and reported around the Chernobyl area. Although it is challenging to incorporate lessons learned from the Chernobyl accident into the source term issues, the author has already developed an example of safety goals by incorporating the radiological consequences of the accident. The example provides safety goals by specifying source term releases in a graded approach in combination with probabilities, i.e. risks. The author believes that the future source term specification should be directly linked with safety goals. (author)« less

Multiple external hazards compound level 3 PSA methods research of nuclear power plant

NASA Astrophysics Data System (ADS)

Wang, Handing; Liang, Xiaoyu; Zhang, Xiaoming; Yang, Jianfeng; Liu, Weidong; Lei, Dina

2017-01-01

2011 Fukushima nuclear power plant severe accident was caused by both earthquake and tsunami, which results in large amount of radioactive nuclides release. That accident has caused the radioactive contamination on the surrounding environment. Although this accident probability is extremely small, once such an accident happens that is likely to release a lot of radioactive materials into the environment, and cause radiation contamination. Therefore, studying accidents consequences is important and essential to improve nuclear power plant design and management. Level 3 PSA methods of nuclear power plant can be used to analyze radiological consequences, and quantify risk to the public health effects around nuclear power plants. Based on multiple external hazards compound level 3 PSA methods studies of nuclear power plant, and the description of the multiple external hazards compound level 3 PSA technology roadmap and important technical elements, as well as taking a coastal nuclear power plant as the reference site, we analyzed the impact of off-site consequences of nuclear power plant severe accidents caused by multiple external hazards. At last we discussed the impact of off-site consequences probabilistic risk studies and its applications under multiple external hazards compound conditions, and explained feasibility and reasonableness of emergency plans implementation.

RMP Guidance for Warehouses - Chapter 3: Five-Year Accident History

EPA Pesticide Factsheets

A five year accident history must be completed for each covered process, and all accidental releases meeting specified criteria must be reported in the Risk Management Plan (RMP). Must include details about each event and its impacts.

Survey of hydrogen monitoring devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lai, W.

Presented are results of a survey of commercially available monitoring devices suitable for hydrogen detection in the secondary containment vessel of a nuclear power plant during the post postulated accident period. Available detectors were grouped into the following five classes: combustion, solid state, electrochemical, thermal conductivity, and absorption. The performance of most available sensors is likely to deteriorate when exposed to the postulated conditions which include moisture, which could be at high temperature, and radioactive noncondensibles. Of the commercial devices, those using metallic filament thermal conductivity detectors seem least susceptible to performance change. Absorption detectors are best suited for thismore » monitoring task but the only available device is designed for pipeline corrosion assessment. Initiation of experimental study to assess apparent deficiencies of commercial detectors is recommended. Also recommended is an analytical/experimental effort to determine the optimum detector array for monitoring in the secondary containment vessels.« less

Risk Assessment for Emergency Planning Related to Nuclear Weapons Accidents

DTIC Science & Technology

1985-09-25

accidents is higher than many of the other accidents, several reviewers thought it appropriate to evaluate these accidents in more detail. In particular, it...consequences is presented in Table 2-5. The plutonium releases in the 1F-3F gram range are due to plutonium burning in Type A weapons, while the higher ...explosion given a tanker casualty varies from 0.05 (in harbor) to 0.16 (at pier) to 0.19 (at sea). Although these numbers are somewhat higher than the

Joint release rate estimation and measurement-by-measurement model correction for atmospheric radionuclide emission in nuclear accidents: An application to wind tunnel experiments.

PubMed

Li, Xinpeng; Li, Hong; Liu, Yun; Xiong, Wei; Fang, Sheng

2018-03-05

The release rate of atmospheric radionuclide emissions is a critical factor in the emergency response to nuclear accidents. However, there are unavoidable biases in radionuclide transport models, leading to inaccurate estimates. In this study, a method that simultaneously corrects these biases and estimates the release rate is developed. Our approach provides a more complete measurement-by-measurement correction of the biases with a coefficient matrix that considers both deterministic and stochastic deviations. This matrix and the release rate are jointly solved by the alternating minimization algorithm. The proposed method is generic because it does not rely on specific features of transport models or scenarios. It is validated against wind tunnel experiments that simulate accidental releases in a heterogonous and densely built nuclear power plant site. The sensitivities to the position, number, and quality of measurements and extendibility of the method are also investigated. The results demonstrate that this method effectively corrects the model biases, and therefore outperforms Tikhonov's method in both release rate estimation and model prediction. The proposed approach is robust to uncertainties and extendible with various center estimators, thus providing a flexible framework for robust source inversion in real accidents, even if large uncertainties exist in multiple factors. Copyright © 2017 Elsevier B.V. All rights reserved.

Analysis of railroad tank car releases using a generalized binomial model.

PubMed

Liu, Xiang; Hong, Yili

2015-11-01

The United States is experiencing an unprecedented boom in shale oil production, leading to a dramatic growth in petroleum crude oil traffic by rail. In 2014, U.S. railroads carried over 500,000 tank carloads of petroleum crude oil, up from 9500 in 2008 (a 5300% increase). In light of continual growth in crude oil by rail, there is an urgent national need to manage this emerging risk. This need has been underscored in the wake of several recent crude oil release incidents. In contrast to highway transport, which usually involves a tank trailer, a crude oil train can carry a large number of tank cars, having the potential for a large, multiple-tank-car release incident. Previous studies exclusively assumed that railroad tank car releases in the same train accident are mutually independent, thereby estimating the number of tank cars releasing given the total number of tank cars derailed based on a binomial model. This paper specifically accounts for dependent tank car releases within a train accident. We estimate the number of tank cars releasing given the number of tank cars derailed based on a generalized binomial model. The generalized binomial model provides a significantly better description for the empirical tank car accident data through our numerical case study. This research aims to provide a new methodology and new insights regarding the further development of risk management strategies for improving railroad crude oil transportation safety. Copyright © 2015 Elsevier Ltd. All rights reserved.

On-line measurement of gaseous iodine species during a PWR severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haykal, I.; Doizi, D.; Perrin, A.

A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still notmore » clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)« less

Assessment of an explosive LPG release accident: a case study.

PubMed

Bubbico, Roberto; Marchini, Mauro

2008-07-15

In the present paper, an accident occurred during a liquefied petroleum gas (LPG) tank filling activity has been taken into consideration. During the transfer of LPG from the source road tank car to the receiving fixed storage vessel, an accidental release of LPG gave rise to different final consequences ranging from a pool fire, to a fireball and to the catastrophic rupture of the tank with successive explosion of its contents. The sequence of events has been investigated by using some of the consequence calculation models most commonly adopted in risk analysis and accident investigation. On one hand, this allows to better understand the link between the various events of the accident. On the other hand, a comparison between the results of the calculations and the damages actually observed after the accident, allows to check the accuracy of the prediction models and to critically assess their validity. In particular, it was shown that the largest uncertainty is associated with the calculation of the energy involved in the physical expansion of the fluid (both liquid and vapor) after the catastrophic rupture of the tank.

Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rates and transport in the reactor vessel during LWR core-melt accidents were studied. The Nuclear Safety Pilot Plant is dedicated to developing an expanded data basemore » on the behavior of aerosols generated during a severe accident.« less

Probability analysis of multiple-tank-car release incidents in railway hazardous materials transportation.

PubMed

Liu, Xiang; Saat, Mohd Rapik; Barkan, Christopher P L

2014-07-15

Railroads play a key role in the transportation of hazardous materials in North America. Rail transport differs from highway transport in several aspects, an important one being that rail transport involves trains in which many railcars carrying hazardous materials travel together. By contrast to truck accidents, it is possible that a train accident may involve multiple hazardous materials cars derailing and releasing contents with consequently greater potential impact on human health, property and the environment. In this paper, a probabilistic model is developed to estimate the probability distribution of the number of tank cars releasing contents in a train derailment. Principal operational characteristics considered include train length, derailment speed, accident cause, position of the first car derailed, number and placement of tank cars in a train and tank car safety design. The effect of train speed, tank car safety design and tank car positions in a train were evaluated regarding the number of cars that release their contents in a derailment. This research provides insights regarding the circumstances affecting multiple-tank-car release incidents and potential strategies to reduce their occurrences. The model can be incorporated into a larger risk management framework to enable better local, regional and national safety management of hazardous materials transportation by rail. Copyright © 2014 Elsevier B.V. All rights reserved.

Hybrid composites that retain graphite fibers on burning

NASA Technical Reports Server (NTRS)

House, E. E.

1980-01-01

A laboratory scale program was conducted to determine fiber release tendencies of graphite reinforced/resinous matrix composites currently used or projected for use in civil aircraft. In the event of an aircraft crash and burn situation, there is concern that graphite fibers will be released from the composites once the resin matrix is thermally decomposed. Hybridizing concepts aimed at preventing fiber release on burning were postulated and their effectiveness evaluated under fire, impact, and air flow during an aircraft crash.

An assessment on the environmental contamination caused by the Fukushima accident.

PubMed

Song, Jin Ho

2018-01-15

The radiological releases from the damaged fuel to the atmosphere and into the cooling water in the Fukushima Daiich Nuclear Power Plant (FDNPP) accident are investigated. Atmospheric releases to the land and ocean mostly occurred during the first week after the accident whereas continuous release from the damaged fuel into the cooling water resulted in an accumulation of contaminated water in the plant during last six years. An evaluation of measurement data and analytical model for the release of radionuclides indicated that atmospheric releases were mainly governed by the volatility of the radionuclides. Using the measurement data on the contaminated water, the mechanism for the release of long-lived radionuclides into the cooling water was analyzed. It was found that the radioactivity concentrations of 90 Sr in the contaminated water in the Primary Containment Vessel (PCV) of unit 2 and unit 3 were consistently higher than that of 137 Cs and the radioactivity concentration of 90 Sr in the turbine building of unit 1 in year 2015 was higher than that in year 2011. It was also observed that the radioactivity concentration of long-lived radionuclides in the contaminated water in the FDNPP is still high even in year 2015. The activity ratio of 238 Pu/ 239+240 Pu for the contaminated water was in the range of 1.7-5.4, which was significantly different from the ratios from the soil samples representing the atmospheric releases of FDNPP. It is concluded that the release mechanisms into the atmosphere and cooling water are clearly different and there has been significant amount of long-lived radionuclides released into the contaminated water. Copyright © 2017 Elsevier Ltd. All rights reserved.

Individual characteristics in occupational accidents due to imbalance: a case-control study of the employees of a railway company.

PubMed

Gauchard, G C; Chau, N; Touron, C; Benamghar, L; Dehaene, D; Perrin, PhP; Mur, J-M

2003-05-01

Falls are frequent occupational accidents, and are responsible for a significant amount of lost working time and, more importantly, for a high mortality. The factors involved in falling mechanisms can be of external or individual origin, the latter being less well identified. To assess the relations between certain individual characteristics and occupational accidents due to imbalance. A total of 427 male employees, who had been victims of at least one occupational accident with sick leave due to imbalance (cases) and 427 controls were recruited among the employees of a large French railway company. A standardised questionnaire on life conditions and professional factors, and a description of the accidents was filled in by an occupational physician for each subject. Some job categories were more affected by a specific release mechanism of work related falls. Certain individual characteristics such as smoking, alcohol consumption, inactivity, sleep disorders, and request for a job change were correlated with the occurrence of occupational accidents. Sick leaves of eight days or over were more frequent in older and overweight injured workers. Some lesions were linked with the specific fall released mechanisms. Individual characteristics can increase the risk of occupational accidents, especially falling. This study identified subjects most at risk on whom prevention related to working conditions and falls could be focused.

Preliminary calculations related to the accident at Three Mile Island

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirchner, W.L.; Stevenson, M.G.

This report discusses preliminary studies of the Three Mile Island Unit 2 (TMI-2) accident based on available methods and data. The work reported includes: (1) a TRAC base case calculation out to 3 hours into the accident sequence; (2) TRAC parametric calculations, these are the same as the base case except for a single hypothetical change in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident; (3) fuel rod cladding failure, cladding oxidation due to zirconium metal-steam reactions, hydrogen release due to cladding oxidation, cladding ballooning, cladding embrittlement,more » and subsequent cladding breakup estimates based on TRAC calculated cladding temperatures and system pressures. Some conclusions of this work are: the TRAC base case accident calculation agrees very well with known system conditions to nearly 3 hours into the accident; the parametric calculations indicate that, loss-of-core cooling was most influenced by the throttling of High-Pressure Injection (HPI) flows, given the accident initiating events and the pressurizer electromagnetic-operated valve (EMOV) failing to close as designed; failure of nearly all the rods and gaseous fission product gas release from the failed rods is predicted to have occurred at about 2 hours and 30 minutes; cladding oxidation (zirconium-steam reaction) up to 3 hours resulted in the production of approximately 40 kilograms of hydrogen.« less

Analytical measurements of fission products during a severe nuclear accident

NASA Astrophysics Data System (ADS)

Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

2018-01-01

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Mechanisms of and facility types involved in hazardous materials incidents.

PubMed Central

Kales, S N; Polyhronopoulos, G N; Castro, M J; Goldman, R H; Christiani, D C

1997-01-01

The purpose of this study was to systematically investigate hazardous materials (hazmat) releases and determine the mechanisms of these accidents, and the industries/activities and chemicals involved. We analyzed responses by Massachusetts' six district hazmat teams from their inception through May 1996. Information from incident reports was extracted onto standard coding sheets. The majority of hazardous materials incidents were caused by spills, leaks, or escapes of hazardous materials (76%) and occurred at fixed facilities (80%). Transportation-related accidents accounted for 20% of incidents. Eleven percent of hazardous materials incidents were at schools or health care facilities. Petroleum-derived fuels were involved in over half of transportation-related accidents, and these accounted for the majority of petroleum fuel releases. Chlorine derivatives were involved in 18% of all accidents and were associated with a wide variety of facility types and activities. In conclusion, systematic study of hazardous materials incidents allows the identification of preventable causes of these incidents. PMID:9300926

28 CFR 301.317 - Medical treatment following release.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 28 Judicial Administration 2 2012-07-01 2012-07-01 false Medical treatment following release. 301... INMATE ACCIDENT COMPENSATION Compensation for Work-Related Physical Impairment or Death § 301.317 Medical treatment following release. Federal Prison lndustries, Inc., may not pay the cost of medical, hospital...

28 CFR 301.317 - Medical treatment following release.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 28 Judicial Administration 2 2013-07-01 2013-07-01 false Medical treatment following release. 301... INMATE ACCIDENT COMPENSATION Compensation for Work-Related Physical Impairment or Death § 301.317 Medical treatment following release. Federal Prison lndustries, Inc., may not pay the cost of medical, hospital...

28 CFR 301.317 - Medical treatment following release.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Medical treatment following release. 301... INMATE ACCIDENT COMPENSATION Compensation for Work-Related Physical Impairment or Death § 301.317 Medical treatment following release. Federal Prison lndustries, Inc., may not pay the cost of medical, hospital...

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.O. Bader

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are releasedmore » from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.« less

Fission product release from fuel under LWR accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Norwood, K.S.

Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 2000/sup 0/C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gammamore » spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-02-01

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

Analysis on tank truck accidents involved in road hazardous materials transportation in china.

PubMed

Shen, Xiaoyan; Yan, Ying; Li, Xiaonan; Xie, Chenjiang; Wang, Lihua

2014-01-01

Due to the sheer size and capacity of the tanker and the properties of cargo transported in the tank, hazmat tanker accidents are more disastrous than other types of vehicle accidents. The aim of this study was to provide a current survey on the situation of accidents involving tankers transporting hazardous materials in China. Detailed descriptions of 708 tanker accidents associated with hazmat transportation in China from 2004 to 2011 were analyzed to identify causes, location, types, time of occurrence, hazard class for materials involved, consequences, and the corresponding probability. Hazmat tanker accidents mainly occurred in eastern (38.1%) and southwest China (12.3%). The most frequent hazmat tanker accidents involved classes 2, 3, and 8. The predominant accident types were rollover (29.10%), run-off-the-road (16.67%), and rear-end collisions (13.28%), with a high likelihood of a large spill occurring. About 55.93% of the accidents occurred on freeways and class 1 roads, with the spill percentage reaching 75.00% and the proportion of spills that occurred in the total accidents amounting to 77.82%, of which 61.72% are considered large spills. The month with the highest accident probability was July (12.29%), and most crashes occurred during the early morning (4:00-6:00 a.m.) and midday (10:00 a.m.-12:00 p.m.) hours, 19.63% versus 16.10%. Human-related errors (73.8%) and vehicle-related defects (19.6%) were the primary reasons for hazmat tanker crashes. The most common outcomes of a hazmat tanker accident was a spill without further events (55.51%), followed by a release with fire (7.77%), and release with an explosion (2.54%). The safety situation of China's hazmat tanker transportation is grim. Such accidents not only have high spill percentages and consistently large spills but they can also cause serious consequences, such as fires and explosions. Improving the training of drivers and the quality of vehicles, deploying roll stability aids, enhancing vehicle inspection and maintenance, and developing good delivery schedules may all be considered effective measures for mitigating hazmat tanker accidents, especially severe crashes.

Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

NASA Astrophysics Data System (ADS)

Lee, Chang Jae; Yun, Jae Hee

2017-06-01

Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

Heavy-section steel irradiation program. Semiannual progress report, October 1995--March 1996

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corwin, W.R.

1997-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents which have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPVs fracture resistance which occurs during service, since without thatmore » radiation damage, it is virtually impossible to postulate a realistic scenario that would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties.« less

New PANDA Tests to Investigate Effects of Light Gases on Passive Safety Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Auban, O.; Candreia, P.

The large- scale thermal-hydraulic PANDA facility (located at PSI in Switzerland), has been used over the last few years for investigating different passive decay- heat removal systems and containment phenomena for the next generation of light water reactors (Simplified Boiling Water Reactor: SBWR; European Simplified Boiling Water Reactor: ESBWR; Siedewasserreaktor: SWR-1000). Currently, as part of the European Commission 5. EURATOM Framework Programme project 'Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling' (TEMPEST), a new series of tests is being planned in the PANDA facility to experimentally investigate the distribution of non-condensable gases inside the containment andmore » their effect on the performance of the 'Passive Containment Cooling System' (PCCS). Hydrogen release caused by the metal-water reaction in the case of a postulated severe accident will be simulated in PANDA by injecting helium into the reactor pressure vessel. In order to provide suitable data for Computational Fluid Dynamic (CFD) code assessment and improvement, the instrumentation in PANDA has been upgraded for the new tests. In the present paper, a detailed discussion is given of the new PANDA tests to be performed to investigate the effects of light gas on passive safety systems. The tests are scheduled for the first half of the year 2002. (authors)« less

Risk evaluation framework and selected metrics for tank cars carrying hazardous materials.

DOT National Transportation Integrated Search

2015-05-01

This report presents an analysis of train accident and hazmat release data to quantify the likelihood of a hazmat release. The harm caused : by a hazmat release is characterized as the end result of a chain of events, with each link in the chain bein...

Behaviour of oceanic 137Cs following the Fukushima Daiichi Nuclear Power Plant accident for four years simulated numerically by a regional ocean model

NASA Astrophysics Data System (ADS)

Tsumune, D.; Tsubono, T.; Aoyama, M.; Misumi, K.; Tateda, Y.

2015-12-01

A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant (1F NPP) following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition.We reconstructed spatiotemporal variability of 137Cs activity in the regional ocean for four years by numerical model, such as a regional scale and the North Pacific scale oceanic dispersion models, an atmospheric transport model, a sediment transport model, a dynamic biological compartment model for marine biota and river runoff model. Direct release rate of 137Cs were estimated for four years after the accident by comparing simulated results and observed activities very close to the site. The estimated total amounts of directly release was 3.6±0.7 PBq. Directly release rate of 137Cs decreased exponentially with time by the end of December 2012 and then, was almost constant. Decrease rate were quite small after 2013. The daily release rate of 137Cs was estimated to be the order of magnitude of 1010 Bq/day by the end of March 2015. The activity of directly released 137Cs was detectable only in the coastal zone after December 2012. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable. There is no observed data of 137Cs activity in the ocean from 11 to 21 March 2011. Observed data of marine biota should reflect the history of 137Cs activity in this early period. We reconstructed the history of 137Cs activity in this early period by considering atmospheric deposition, river input, rain water runoff from the 1F NPP site. The comparisons between simulated 137Cs activity of marine biota by a dynamic biological compartment and observed data also suggest that simulated 137Cs activity attributable to atmospheric deposition was underestimated in this early period. The simulated river flux of 137Cs to the ocean did not effect on 137Cs activity in the ocean even if the parameters in this simulation have uncertainties because of the lack of observed data in rivers in the earlier period.

Behaviour of oceanic 137Cs following the Fukushima Daiichi Nuclear Power Plant accident for four years simulated numerically by a regional ocean model

NASA Astrophysics Data System (ADS)

Torn, M. S.; Koven, C. D.; Riley, W. J.; Zhu, B.; Hicks Pries, C.; Phillips, C. L.

2014-12-01

A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant (1F NPP) following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition.We reconstructed spatiotemporal variability of 137Cs activity in the regional ocean for four years by numerical model, such as a regional scale and the North Pacific scale oceanic dispersion models, an atmospheric transport model, a sediment transport model, a dynamic biological compartment model for marine biota and river runoff model. Direct release rate of 137Cs were estimated for four years after the accident by comparing simulated results and observed activities very close to the site. The estimated total amounts of directly release was 3.6±0.7 PBq. Directly release rate of 137Cs decreased exponentially with time by the end of December 2012 and then, was almost constant. Decrease rate were quite small after 2013. The daily release rate of 137Cs was estimated to be the order of magnitude of 1010 Bq/day by the end of March 2015. The activity of directly released 137Cs was detectable only in the coastal zone after December 2012. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable. There is no observed data of 137Cs activity in the ocean from 11 to 21 March 2011. Observed data of marine biota should reflect the history of 137Cs activity in this early period. We reconstructed the history of 137Cs activity in this early period by considering atmospheric deposition, river input, rain water runoff from the 1F NPP site. The comparisons between simulated 137Cs activity of marine biota by a dynamic biological compartment and observed data also suggest that simulated 137Cs activity attributable to atmospheric deposition was underestimated in this early period. The simulated river flux of 137Cs to the ocean did not effect on 137Cs activity in the ocean even if the parameters in this simulation have uncertainties because of the lack of observed data in rivers in the earlier period.

The Results From the First High-Pressure Melt Ejection Test Completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nitheanandan, T.; Kyle, G.; O'Connor, R.

2006-07-01

A high-pressure melt ejection test using prototypical corium was conducted at Atomic Energy of Canada Limited Chalk River Laboratories. This test was planned by the CANDU Owners Group to study the potential for an energetic interaction between molten fuel and water under postulated single-channel flow-blockage events. The experiments were designed to address regulator concerns surrounding this very low probability postulated accident events in CANDU Pressurized Heavy Water Reactors. The objective of the experimental program is to determine whether a highly energetic 'steam explosion' and associated high-pressure pulse, is possible when molten material is finely fragmented as it is ejected frommore » a fuel channel into the heavy-water moderator. The finely fragmented melt particles would transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection test consisted of heating up a {approx} 5 kg thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3} inside a 1.14-m length of insulated pressure tube. When the molten material reached the desired temperature of {approx} 2400 deg C, the pressure inside the tube was raised to 11.6 MPa, failing the pressure tube at a pre-machined flaw, and releasing the molten material into the surrounding tank of 68 deg C water. The experiment investigated the dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels. The measured mean particle size was 0.686 mm and the peak dynamic pressures were between 2.54 and 4.36 MPa, indicating that an energetic interaction between the melt and the water did not occur in the test. (authors)« less

Seveso 1986, Chernobyl 1976: a physicist' look at 2 ecological disasters

NASA Astrophysics Data System (ADS)

Ratti, S.

2004-05-01

Seveso suffered a chemical accident with a severe loss of supertoxic material (TCCD) released in the atmosphere; Chernobyl was a world known nuclear accident. The pollution induced by the two accident are analysed in term of fractal models. The first case involved a limited micro ecological system; the second one spread over a macro ecological system. The pollution is reproduced by means of simple Fractal Sum of Pulses models in the Seveso region; for the Chernobyl accident in northern Italy and in several european Countries. The 2 accidents are also analysed in terms of Universal Multifractals showing that thethe parameters α and C1 are those describing respectively rainfall (Seveso) and cloud formation (Chernobyl).

Accident at the Fukushima Dai-ichi nuclear power stations of TEPCO--outline & lessons learned.

PubMed

Tanaka, Shun-ichi

2012-01-01

The severe accident that broke out at Fukushima Dai-ichi nuclear power stations on March 11, 2011, caused seemingly infinite damage to the daily life of residents. Serious and wide-spread contamination of the environment occurred due to radioactive materials discharged from nuclear power stations (NPSs). At the same time, many issues were highlighted concerning countermeasures to severe nuclear accidents. The accident is outlined, and lessons learned are extracted with respect to the safety of NPSs, as well as radiation protection of residents under the emergency involving the accident. The materials of the current paper are those released by governmental agencies, academic societies, interim reports of committees under the government, and others.

Introduction of the Space Shuttle Columbia Accident, Investigation Details, Findings and Crew Survival Investigation Report

NASA Technical Reports Server (NTRS)

Chandler, Michael

2010-01-01

As the Space Shuttle Program comes to an end, it is important that the lessons learned from the Columbia accident be captured and understood by those who will be developing future aerospace programs and supporting current programs. Aeromedical lessons learned from the Accident were presented at AsMA in 2005. This Panel will update that information, closeout the lessons learned, provide additional information on the accident and provide suggestions for the future. To set the stage, an overview of the accident is required. The Space Shuttle Columbia was returning to Earth with a crew of seven astronauts on 1Feb, 2003. It disintegrated along a track extending from California to Louisiana and observers along part of the track filmed the breakup of Columbia. Debris was recovered from Littlefield, Texas to Fort Polk, Louisiana, along a 567 statute mile track; the largest ever recorded debris field. The Columbia Accident Investigation Board (CAIB) concluded its investigation in August 2003, and released their findings in a report published in February 2004. NASA recognized the importance of capturing the lessons learned from the loss of Columbia and her crew and the Space Shuttle Program managers commissioned the Spacecraft Crew Survival Integrated Investigation Team (SCSIIT) to accomplish this. Their task was to perform a comprehensive analysis of the accident, focusing on factors and events affecting crew survival, and to develop recommendations for improving crew survival, including the design features, equipment, training and procedures intended to protect the crew. NASA released the Columbia Crew Survival Investigation Report in December 2008. Key personnel have been assembled to give you an overview of the Space Shuttle Columbia accident, the medical response, the medico-legal issues, the SCSIIT findings and recommendations and future NASA flight surgeon spacecraft accident response training. Educational Objectives: Set the stage for the Panel to address the investigation, medico-legal issues, the Spacecraft Crew Survival Integrated Investigation Team report and training for accident response.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2016-12-01

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in thismore » study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU analyses. Additionally, current analyses suggest that the NUREG-1465 release fractions are conservative by about a factor of 2 in terms of release fractions and that release durations for in-vessel and late in-vessel release periods are in fact longer than the NUREG-1465 durations. It is currently planned that a subsequent report will further characterize these results using more refined statistical methods, permitting a more precise reformulation of the NUREG-1465 alternative source term for both LBU and HBU fuels, with the most important finding being that the NUREG-1465 formula appears to embody significant conservatism compared to current best-estimate analyses. ACKNOWLEDGEMENTS This work was supported by the United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The authors would like to thank Dr. Ian Gauld and Dr. Germina Ilas, of Oak Ridge National Laboratory, for their contributions to this work. In addition to development of core fission product inventory and decay heat information for use in MELCOR models, their insights related to fuel management practices and resulting effects on spatial distribution of fission products in the core was instrumental in completion of our work.« less

Ongoing environmental monitoring and assessment of the long-term impacts of the February 2014 radiological release from the waste isolation pilot plant.

PubMed

Thakur, Punam; Runyon, Tim

2018-04-09

Three years ago, the Waste Isolation Pilot Plant (WIPP) experienced its first minor accident involving a radiological release. Late in the evening on February 14, 2014, a waste container in the repository underwent a chemical reaction that caused the container to overheat and breach, releasing its contents into the underground. Following a lengthy recovery process, the facility recently resumed waste disposal operations. The accident released significant levels of radioactivity into the disposal room and adjacent exhaust drifts, and although no one was present in the underground at the time of the release, a total of 22 workers tested positive for very low level of radiation, presumably from some of the radioactive material that was released above ground through a small leak in the HEPA filtration system. The dominant radionuclides released were 241 Am and 239 + 240 Pu in a ratio that matched the content of the drum from Los Alamos National Laboratory (LANL) that was eventually identified as the breached container. From the air particulate monitoring and plume modeling, it was concluded that the dose, at the nearest location accessible to the general public, from this radiation release event would have been less than 0.01 mSv (< 1 mrem/year). This level is well below the 0.1 mSv/year (10 mrem/year) regulatory limit for DOE facilities established by the US Environmental Protection Agency (EPA).While no long-term impacts to public health or the environment are expected as a result of the WIPP radiation release, the limited ventilation and residual contamination levels in the underground are still a concern and pose a major challenge for the full recovery of WIPP. This article provides an up-to-date overview of environmental monitoring results through the WIPP recovery and an estimate of the long-term impacts of the accident on the natural and human environment.

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Comparison of Radiation Dose Studies of the 2011 Fukushima Nuclear Accident Prepared by the World Health Organization and the U.S. Department of Defense

DTIC Science & Technology

2012-11-01

R T Comparison of Radiation Dose Studies of the 2011 Fukushima Nuclear Accident Prepared by the World Health Organization and the U.S. Department...AND SUBTITLE Comparison of Radiation Dose Studies of the 2011 Fukushima Nuclear Accident Prepared by the World Health Organization and the U.S...in Japan on March 11, 2011 led to releases of radioactive materials from the Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Station

Pathways of proton release in the bacteriorhodopsin photocycle

NASA Technical Reports Server (NTRS)

Zimanyi, L.; Varo, G.; Chang, M.; Ni, B.; Needleman, R.; Lanyi, J. K.

1992-01-01

The pH dependencies of the rate constants in the photocycles of recombinant D96N and D115N/D96N bacteriorhodopsins were determined from time-resolved difference spectra between 70 ns and 420 ms after photoexcitation. The results were consistent with the model suggested earlier for proteins containing D96N substitution: BR hv----K----L----M1----M2----BR. Only the M2----M1 back-reaction was pH-dependent: its rate increased with increasing [H+] between pH 5 and 8. We conclude from quantitative analysis of this pH dependency that its reverse, the M1----M2 reaction, is linked to the release of a proton from a group with a pKa = 5.8. This suggests a model for wild-type bacteriorhodopsin in which at pH greater than 5.8 the transported proton is released on the extracellular side from this as yet unknown group and on the 100-microseconds time scale, but at pH less than 5.8, the proton release occurs from another residue and later in the photocycle most likely directly from D85 during the O----BR reaction. We postulate, on the other hand, that proton uptake on the cytoplasmic side will be by D96 and during the N----O reaction regardless of pH. The proton kinetics as measured with indicator dyes confirmed the unique prediction of this model: at pH greater than 6, proton release preceded proton uptake, but at pH less than 6, the release was delayed until after the uptake. The results indicated further that the overall M1----M2 reaction includes a second kinetic step in addition to proton release; this is probably the earlier postulated extracellular-to-cytoplasmic reorientation switch in the proton pump.

Environmental justice: frequency and severity of US chemical industry accidents and the socioeconomic status of surrounding communities.

PubMed

Elliott, M R; Wang, Y; Lowe, R A; Kleindorfer, P R

2004-01-01

The Clean Air Act Amendments of 1990 requires that chemical facilities in the US with specified quantities of certain toxic or flammable chemicals file a five year history of accidents. This study considers the relation between the reported accidents and surrounding community characteristics. This study is a retrospective analysis of the association between the demographics of counties in which facilities are located and the risk of accidental chemical release and resulting injuries at those facilities. The "location risk" (the risk that a facility having large volumes of hazardous chemicals is located in a community) and "operations risk" (the risk of an accident itself) are investigated. 1994-2000 accident history data from 15 083 US industrial facilities using one or more of 140 flammable or toxic substances above a threshold level. Demographic makeup of 2333 counties surrounding these facilities was determined from the 1990 US census. Larger and more chemical intensive facilities tend to be located in counties with larger African-American populations and in counties with both higher median incomes and high levels of income inequality. Even after adjusting for location risk there is greater risk of accidents for facilities in heavily African-American counties (OR of accident = 1.9, 95% CI = 1.5 to 2.4). Further research and policy interventions are required to reduce the probability of locating facilities in an inequitable fashion, as well as health surveillance, and regulatory monitoring and enforcement activities to ensure that hazardous facilities in minority communities prepare and prevent accidental chemical releases to the same standards as elsewhere.

The kinetics of aerosol particle formation and removal in NPP severe accidents

NASA Astrophysics Data System (ADS)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

2016-06-01

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

Accident at the Fukushima Dai-ichi Nuclear Power Stations of TEPCO —Outline & lessons learned—

PubMed Central

TANAKA, Shun-ichi

2012-01-01

The severe accident that broke out at Fukushima Dai-ichi nuclear power stations on March 11, 2011, caused seemingly infinite damage to the daily life of residents. Serious and wide-spread contamination of the environment occurred due to radioactive materials discharged from nuclear power stations (NPSs). At the same time, many issues were highlighted concerning countermeasures to severe nuclear accidents. The accident is outlined, and lessons learned are extracted with respect to the safety of NPSs, as well as radiation protection of residents under the emergency involving the accident. The materials of the current paper are those released by governmental agencies, academic societies, interim reports of committees under the government, and others. PMID:23138450

The kinetics of aerosol particle formation and removal in NPP severe accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.

2016-06-08

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal–hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into themore » KUPOL-M thermal–hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.« less

Hazardous material releases due to unsecured openings and lining failures, volume 1.

DOT National Transportation Integrated Search

1990-12-01

In response to the large number of unintentional releases of hazardous : materials from railroad tank cars for which accidents were not the cause, the : Federal Railroad Administration (FRA) initiated this study to (1) recommend : procedures to ensur...

Fact Sheet: Clean Air Act Section 112(r): Accidental Release Prevention / Risk Management Plan Rule

EPA Pesticide Factsheets

EPA is required to publish regulations and guidance for chemical accident prevention at facilities that pose the greatest risk of harm from accidental releases of regulated flammable and toxic substances above threshold quantities.

An overview of current knowledge concerning the health and environmental consequences of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident.

PubMed

Aliyu, Abubakar Sadiq; Evangeliou, Nikolaos; Mousseau, Timothy Alexander; Wu, Junwen; Ramli, Ahmad Termizi

2015-12-01

Since 2011, the scientific community has worked to identify the exact transport and deposition patterns of radionuclides released from the accident at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in Japan. Nevertheless, there still remain many unknowns concerning the health and environmental impacts of these radionuclides. The present paper reviews the current understanding of the FDNPP accident with respect to interactions of the released radionuclides with the environment and impacts on human and non-human biota. Here, we scrutinize existing literature and combine and interpret observations and modeling assessments derived after Fukushima. Finally, we discuss the behavior and applications of radionuclides that might be used as tracers of environmental processes. This review focuses on (137)Cs and (131)I releases derived from Fukushima. Published estimates suggest total release amounts of 12-36.7PBq of (137)Cs and 150-160PBq of (131)I. Maximum estimated human mortality due to the Fukushima nuclear accident is 10,000 (due to all causes) and the maximum estimates for lifetime cancer mortality and morbidity are 1500 and 1800, respectively. Studies of plants and animals in the forests of Fukushima have recorded a range of physiological, developmental, morphological, and behavioral consequences of exposure to radioactivity. Some of the effects observed in the exposed populations include the following: hematological aberrations in Fukushima monkeys; genetic, developmental and morphological aberrations in a butterfly; declines in abundances of birds, butterflies and cicadas; aberrant growth forms in trees; and morphological abnormalities in aphids. These findings are discussed from the perspective of conservation biology. Copyright © 2015 Elsevier Ltd. All rights reserved.

Computational Aerodynamics of Shuttle Orbiter Damage Scenarios in Support of the Columbia Accident Investigation

NASA Technical Reports Server (NTRS)

Bibb, Karen L.; Prabhu, Ramadas K.

2004-01-01

In support of the Columbia Accident Investigation, inviscid computations of the aerodynamic characteristics for various Shuttle Orbiter damage scenarios were performed using the FELISA unstructured CFD solver. Computed delta aerodynamics were compared with the reconstructed delta aerodynamics in order to postulate a progression of damage through the flight trajectory. By performing computations at hypervelocity flight and CF4 tunnel conditions, a bridge was provided between wind tunnel testing in Langley's 20-Inch CF4 facility and the flight environment experienced by Columbia during re-entry. The rapid modeling capability of the unstructured methodology allowed the computational effort to keep pace with the wind tunnel and, at times, guide the wind tunnel efforts. These computations provided a detailed view of the flowfield characteristics and the contribution of orbiter components (such as the vertical tail and wing) to aerodynamic forces and moments that were unavailable from wind tunnel testing. The damage scenarios are grouped into three categories. Initially, single and multiple missing full RCC panels were analyzed to determine the effect of damage location and magnitude on the aerodynamics. Next is a series of cases with progressive damage, increasing in severity, in the region of RCC panel 9. The final group is a set of wing leading edge and windward surface deformations that model possible structural deformation of the wing skin due to internal heating of the wing structure. By matching the aerodynamics from selected damage scenarios to the reconstructed flight aerodynamics, a progression of damage that is consistent with the flight data, debris forensics, and wind tunnel data is postulated.

Reconstruction of 137Cs activity in the ocean following the Fukushima Daiichi Nuclear Power Plant Accident

NASA Astrophysics Data System (ADS)

Tsumune, Daisuke; Aoyama, Michio; Tsubono, Takaki; Tateda, Yutaka; Misumi, Kazuhiro; Hayami, Hiroshi; Toyoda, Yasuhiro; Maeda, Yoshiaki; Yoshida, Yoshikatsu; Uematsu, Mitsuo

2014-05-01

A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition. We reconstructed spatiotemporal variability of 137Cs activity in the ocean by the comparison model simulations and observed data. We employed a regional scale and the North Pacific scale oceanic dispersion models, an atmospheric transport model, a sediment transport model, a dynamic biological compartment model for marine biota and river runoff model to investigate the oceanic contamination. Direct releases of 137Cs were estimated for more than 2 years after the accident by comparing simulated results and observed activities very close to the site. The estimated total amounts of directly released 137Cs was 3.6±0.7 PBq. Directly release rate of 137Cs decreased exponentially with time by the end of December 2012 and then, was almost constant. The daily release rate of 137Cs was estimated to be 3.0 x 1010 Bq day-1 by the end of September 2013. The activity of directly released 137Cs was detectable only in the coastal zone after December 2012. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable, while simulated 137Cs activities attributable to atmospheric deposition were low compared to measured activities. The rate of atmospheric deposition onto the ocean was underestimated because of a lack of measurements of dose rate and air activity of 137Cs over the ocean when atmospheric deposition rates were being estimated. Observed 137Cs activities attributable to atmospheric deposition in the ocean helped to improve the accuracy of simulated atmospheric deposition rates. Although there is no observed data of 137Cs activity in the ocean from 11 to 21 March 2011, observed data of marine biota should reflect the history of 137Cs activity in this early period. The comparisons between simulated 137Cs activity of marine biota by a dynamic biological compartment and observed data also suggest that simulated 137Cs activity attributable to atmospheric deposition was underestimated in this early period. In addition, river runoff model simulations suggest that the river flux of 137Cs to the ocean was effective to the 137Cs activity in the ocean in this early period. The sediment transport model simulations suggests that the inventory of 137Cs in sediment was less than 10

Space Propulsion Hazards Analysis Manual (SPHAM). Volume 2. Appendices

DTIC Science & Technology

1988-10-01

lb. RESTRICTIVE MARKINGS UNCLASSIFIED 2a. SECURITY CLASSIFICATION AUTHORITY 3 . DISTRIBUTION/AVAILABILITY OF REPORT Approved for public release...Volume I Chapter 2 - Requirementb and the Hazards Analysis Process .... Volume I Chapter 3 - Accident Scenarios...list of the hazardous materials that are discussed; 3 ) description of the failure scenarios; 4) type of post-accident environment that is discussed

49 CFR 40.331 - To what additional parties must employers and service agents release information?

Code of Federal Regulations, 2012 CFR

2012-10-01

... information about that employee's drug or alcohol tests to an identified person, you must provide the information to the identified person. For example, as an employer, when you receive a written request from a... Safety Board as part of an accident investigation, you must provide information concerning post-accident...

49 CFR 40.331 - To what additional parties must employers and service agents release information?

Code of Federal Regulations, 2013 CFR

2013-10-01

... information about that employee's drug or alcohol tests to an identified person, you must provide the information to the identified person. For example, as an employer, when you receive a written request from a... Safety Board as part of an accident investigation, you must provide information concerning post-accident...

Atmospheric transport of radioactive debris to Norway in case of a hypothetical accident related to the recovery of the Russian submarine K-27.

PubMed

Bartnicki, Jerzy; Amundsen, Ingar; Brown, Justin; Hosseini, Ali; Hov, Øystein; Haakenstad, Hilde; Klein, Heiko; Lind, Ole Christian; Salbu, Brit; Szacinski Wendel, Cato C; Ytre-Eide, Martin Album

2016-01-01

The Russian nuclear submarine K-27 suffered a loss of coolant accident in 1968 and with nuclear fuel in both reactors it was scuttled in 1981 in the outer part of Stepovogo Bay located on the eastern coast of Novaya Zemlya. The inventory of spent nuclear fuel on board the submarine is of concern because it represents a potential source of radioactive contamination of the Kara Sea and a criticality accident with potential for long-range atmospheric transport of radioactive particles cannot be ruled out. To address these concerns and to provide a better basis for evaluating possible radiological impacts of potential releases in case a salvage operation is initiated, we assessed the atmospheric transport of radionuclides and deposition in Norway from a hypothetical criticality accident on board the K-27. To achieve this, a long term (33 years) meteorological database has been prepared and used for selection of the worst case meteorological scenarios for each of three selected locations of the potential accident. Next, the dispersion model SNAP was run with the source term for the worst-case accident scenario and selected meteorological scenarios. The results showed predictions to be very sensitive to the estimation of the source term for the worst-case accident and especially to the sizes and densities of released radioactive particles. The results indicated that a large area of Norway could be affected, but that the deposition in Northern Norway would be considerably higher than in other areas of the country. The simulations showed that deposition from the worst-case scenario of a hypothetical K-27 accident would be at least two orders of magnitude lower than the deposition observed in Norway following the Chernobyl accident. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

The potential for disease initiation by inhaled beta-emitting nuclear particles.

PubMed

Aamodt, Norman O

2018-07-01

There were many anecdotal reports of injuries to humans, animals and plants following the Three Mile Island, Chernobyl and Fukushima accidents that were indicative of radiation exposures that delivered a dose of at least 0.5 Sieverts, but studies that attempted to relate observed increases of cancer rates and other injuries with exposure to the radioactive releases from these accidents have failed to find an association. To resolve this dissonance, it was assumed that an analysis of knowledge about accident releases and health effects gathered from one of these accidents could lead to the identification of an unrecognized exposure that could be inferred to have caused a specific observed injury that required a dose of at least 0.5 Sieverts. Because there is considerably more useful knowledge of reactor releases of radionuclides and observed health effects related to the Three Mile Island accident, that knowledge is analyzed. A relationship is inferred to exist between exposure to particulates in plumes released from the station vent stack and severe prolonged immunosuppression, a known effect of an exposure that delivers a dose of at least 0.5 Sieverts. More than ninety percent of the particulates were comprised of Strontium 89 nuclides, essentially pure beta emitters. Because Strontium is a metal, the nuclides in the particulates were configured as stable crystals which, when coming to rest in body tissue, functioned as intense point sources of chronic beta irradiation. The inference led to the hypothesis "particulates comprised of Strontium 89 nuclides provided the exposure that caused the health effects that were observed following the Three Mile Island accident". The hypothesis was tested for validity against two requirements; that only the humans beneath the plumes, who would have inhaled the particulates, expressed the abrupt and persistent rise in the health parameter Deaths from all Causes that would result from the severe prolonged immunosuppression that follows a 0.5 Sievert dose, and that they expressed the excess cancers that would be expected following organ doses of 0.5 or more Sieverts. These effects were found. The Hypothesis is therefore valid and leads to knowledge of the heretofore unknown mechanisms and effects of low LET beta irradiation by a particulate. Copyright © 2018. Published by Elsevier Ltd.

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE PAGES

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald; ...

2014-05-01

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

An overview of Fukushima radionuclides measured in the northern hemisphere.

PubMed

Thakur, P; Ballard, S; Nelson, R

2013-08-01

The Great East Japan Earthquake and tsunami on March 11, 2011 resulted in the tragic accident at the Fukushima Nuclear Power Plant (NPP) and subsequently uncontrolled release of radioactive contaminants into the atmosphere. This review article attempts to compile and interpret data collected by various national and international monitoring networks in response to the Fukushima releases across the northern hemisphere. The majority of the releases occurred during the period March 12-22 with a maximum release phase from March 14-17, 2011. The radioactivity released was dominated by volatile fission products including isotopes of the noble gases (xenon and krypton), iodine, cesium, and tellurium. The radioactive gases and particles released in the accident were dispersed over the middle latitudes of the entire northern hemisphere and for the first time also measured in the southern Hemisphere. Isotopes of iodine and cesium were detected in air, water, milk and food samples collected across the entire northern hemisphere. Elevated levels of fission products were detected from March to May 2011 at many locations over the northern hemisphere. This article focuses on the most prevalent cesium and iodine isotopes, but other secondary isotopes are also discussed. Spatial and temporal patterns and differences are contrasted. The activity ratios of (131)I/(137)Cs and (134)Cs/(137)Cs measured at several locations are evaluated to gain an insight into the fuel burn-up, the inventory of radionuclides in the reactor and the isotopic signature of the accident. It is important to note that all of the radiation levels detected outside of Japan have been very low and are well below any level of public and environmental hazard. Published by Elsevier B.V.

Surveys of Food Intake Just after the Nuclear Accident at the Fukushima Daiichi Nuclear Power Station.

PubMed

Hirakawa, Sachiko; Yoshizawa, Nobuaki; Murakami, Kana; Takizawa, Mari; Kawai, Masaki; Sato, Osamu; Takagi, Shunji; Suzuki, Gen

2017-01-01

As a result of the nuclear accident at the Fukushima Daiichi nuclear power station (FDNPS) after the Great East Japan Earthquake on March 11, 2011, volatile radionuclides including iodine-131 were released into the environment and contaminated open-field vegetables, raw milk, tap water, etc. It is important for the health care of residents to correctly comprehend the level of their exposure to radioactive substances released following the accident. However, an evaluation of the internal exposure doses of residents of Fukushima Prefecture as a result of the ingestion of foods, which is indicated in the report issued by United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR)1 is based on a number of assumptions. For instance, the estimation assumes that foods were ingested as usual, without regard to the places to which residents were evacuated after the accident, the places where food shipment restrictions were imposed, and so forth. The present report aims to improve the accuracy of estimation of the amount of food actually ingested at evacuation areas, in order to reduce as much as possible the level of uncertainty in conventional values estimated directly after the accident, which were in fact values based on conservative assumptions. More concretely, as basic source material to more accurately estimate internal exposure doses from food ingestion, various patterns of evacuation and dietary habits at the time of the accident of the residents of 13 municipalities in Fukushima Prefecture who were evacuated during the period from directly after the accident of March 11, 2011 until the end of March are clarified in this report. From survey results, most of the food that evacuees took immediately after the accident was confirmed to have been sourced from either stockpiles prepared before the accident, or relief supplies from outside of the affected areas. The restriction orders of food supplies such as contaminated vegetables and milk, and tap water intake were implemented within several days after the major release of radionuclides on March 15, 2011. In addition, collapse in supply chains, i.e., damage to distribution facilities, lack of transportation vehicles or electricity, and the closure of retail stores, contributed to a situation where food or supplies contaminated with iodine -131 were not consumed in large quantities in general, even before the food restriction order. Since people consumed tap water and water from other sources before the implementation of restriction orders in affected areas, we surveyed the status of water as a potential route of internal exposure.

Columbia Accident Investigation Board Report. Volume 3

NASA Technical Reports Server (NTRS)

White, Donald J. (Editor); Goodman, Patrick A. (Editor); Reingold, Lester A. (Editor); Kirchhoff, Christopher M. (Editor); Simon, Ariel H. (Editor)

2003-01-01

This report describes the results of an investigative analysis performed by the Air Force Research Laboratory Sensors Directorate at th the specific request of the Defense Columbia Investigation Support Team (DCIST) who was supporting the Columbia Accident Investigation Board (CAIB). The work was performed during the period February 20, 2003 through 20 July 2003. An interim release of measurement findings was provided the CAIB on 24 April 2003, and the information was released in public testimony to the CAIB on May 6, 2003 at the Hilton Hotel, Houston, Texas. The overall assessment and conclusions of this report are consistent with the CAIB 6 May 2003 testimony, with one notable exception discussed in Section VI. This report has been reviewed by the AFRL/SN Flight Day Two DCIST appointed assessment team, and is hereby released to the CAIB and DCIST for final disposition.

Inverse modelling of radionuclide release rates using gamma dose rate observations

NASA Astrophysics Data System (ADS)

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

2015-04-01

Severe accidents in nuclear power plants such as the historical accident in Chernobyl 1986 or the more recent disaster in the Fukushima Dai-ichi nuclear power plant in 2011 have drastic impacts on the population and environment. Observations and dispersion modelling of the released radionuclides help to assess the regional impact of such nuclear accidents. Modelling the increase of regional radionuclide activity concentrations, which results from nuclear accidents, underlies a multiplicity of uncertainties. One of the most significant uncertainties is the estimation of the source term. That is, the time dependent quantification of the released spectrum of radionuclides during the course of the nuclear accident. The quantification of the source term may either remain uncertain (e.g. Chernobyl, Devell et al., 1995) or rely on estimates given by the operators of the nuclear power plant. Precise measurements are mostly missing due to practical limitations during the accident. The release rates of radionuclides at the accident site can be estimated using inverse modelling (Davoine and Bocquet, 2007). The accuracy of the method depends amongst others on the availability, reliability and the resolution in time and space of the used observations. Radionuclide activity concentrations are observed on a relatively sparse grid and the temporal resolution of available data may be low within the order of hours or a day. Gamma dose rates, on the other hand, are observed routinely on a much denser grid and higher temporal resolution and provide therefore a wider basis for inverse modelling (Saunier et al., 2013). We present a new inversion approach, which combines an atmospheric dispersion model and observations of radionuclide activity concentrations and gamma dose rates to obtain the source term of radionuclides. We use the Lagrangian particle dispersion model FLEXPART (Stohl et al., 1998; Stohl et al., 2005) to model the atmospheric transport of the released radionuclides. The inversion method uses a Bayesian formulation considering uncertainties for the a priori source term and the observations (Eckhardt et al., 2008, Stohl et al., 2012). The a priori information on the source term is a first guess. The gamma dose rate observations are used to improve the first guess and to retrieve a reliable source term. The details of this method will be presented at the conference. This work is funded by the Bundesamt für Strahlenschutz BfS, Forschungsvorhaben 3612S60026. References Davoine, X. and Bocquet, M., Atmos. Chem. Phys., 7, 1549-1564, 2007. Devell, L., et al., OCDE/GD(96)12, 1995. Eckhardt, S., et al., Atmos. Chem. Phys., 8, 3881-3897, 2008. Saunier, O., et al., Atmos. Chem. Phys., 13, 11403-11421, 2013. Stohl, A., et al., Atmos. Environ., 32, 4245-4264, 1998. Stohl, A., et al., Atmos. Chem. Phys., 5, 2461-2474, 2005. Stohl, A., et al., Atmos. Chem. Phys., 12, 2313-2343, 2012.

40 CFR 68.87 - Contractors.

Code of Federal Regulations, 2011 CFR

2011-07-01

... ACCIDENT PREVENTION PROVISIONS Program 3 Prevention Program § 68.87 Contractors. (a) Application. This... known potential fire, explosion, or toxic release hazards related to the contractor's work and the... instructed in the known potential fire, explosion, or toxic release hazards related to his/her job and the...

40 CFR 68.87 - Contractors.

Code of Federal Regulations, 2012 CFR

2012-07-01

... ACCIDENT PREVENTION PROVISIONS Program 3 Prevention Program § 68.87 Contractors. (a) Application. This... known potential fire, explosion, or toxic release hazards related to the contractor's work and the... instructed in the known potential fire, explosion, or toxic release hazards related to his/her job and the...

40 CFR 68.87 - Contractors.

Code of Federal Regulations, 2014 CFR

2014-07-01

... ACCIDENT PREVENTION PROVISIONS Program 3 Prevention Program § 68.87 Contractors. (a) Application. This... known potential fire, explosion, or toxic release hazards related to the contractor's work and the... instructed in the known potential fire, explosion, or toxic release hazards related to his/her job and the...

40 CFR 68.87 - Contractors.

Code of Federal Regulations, 2013 CFR

2013-07-01

... ACCIDENT PREVENTION PROVISIONS Program 3 Prevention Program § 68.87 Contractors. (a) Application. This... known potential fire, explosion, or toxic release hazards related to the contractor's work and the... instructed in the known potential fire, explosion, or toxic release hazards related to his/her job and the...

Assessment of radiation doses in the UK from the Fukushima Daiichi Nuclear accident.

PubMed

Brown, J

2014-06-01

PHE has undertaken a simple dose assessment for members of the public living in the UK at the time of the accident at the Fukushima Daiichi nuclear power station in March 2011. PHE reported that there was no public health risk to the UK from the release of material from the accident in a statement made on 29 March 2013. This assessment confirms the initial estimate of the doses which were about the same as a person in the UK would receive in an hour from natural background.

[Nuclear-power-plant accidents: thyroid cancer incidence and radiation-related health effects from the Chernobyl accident].

PubMed

Schlumberger, Martin; Le Guen, Bernard

2012-01-01

Following the Chernobyl accident, enormous amounts of radioisotopes were released in the atmosphere and have contaminated surrounding populations in the absence of rapid protective countermeasures. The highest radiation doses were delivered to the thyroid gland, and the only direct consequence of radiation exposure observed among contaminated population is the increased incidence of thyroid cancers among subjects who were children in 1986 and who lived at that time in Belarus, Ukraine or Russia. © 2012 médecine/sciences – Inserm / SRMS.

Environmental justice: frequency and severity of US chemical industry accidents and the socioeconomic status of surrounding communities

PubMed Central

Elliott, M; Wang, Y; Lowe, R; Kleindorfer, P

2004-01-01

Study objectives: The Clean Air Act Amendments of 1990 requires that chemical facilities in the US with specified quantities of certain toxic or flammable chemicals file a five year history of accidents. This study considers the relation between the reported accidents and surrounding community characteristics. Design: This study is a retrospective analysis of the association between the demographics of counties in which facilities are located and the risk of accidental chemical release and resulting injuries at those facilities. The "location risk" (the risk that a facility having large volumes of hazardous chemicals is located in a community) and "operations risk" (the risk of an accident itself) are investigated. Setting:1994–2000 accident history data from 15 083 US industrial facilities using one or more of 140 flammable or toxic substances above a threshold level. Demographic makeup of 2333 counties surrounding these facilities was determined from the 1990 US census. Main results: Larger and more chemical intensive facilities tend to be located in counties with larger African-American populations and in counties with both higher median incomes and high levels of income inequality. Even after adjusting for location risk there is greater risk of accidents for facilities in heavily African-American counties (OR of accident = 1.9, 95% CI = 1.5 to 2.4). Conclusions: Further research and policy interventions are required to reduce the probability of locating facilities in an inequitable fashion, as well as health surveillance, and regulatory monitoring and enforcement activities to ensure that hazardous facilities in minority communities prepare and prevent accidental chemical releases to the same standards as elsewhere. PMID:14684723

The Chernobyl accident and its consequences.

PubMed

Saenko, V; Ivanov, V; Tsyb, A; Bogdanova, T; Tronko, M; Demidchik, Yu; Yamashita, S

2011-05-01

The accident at the Chernobyl nuclear power plant was the worst industrial accident of the last century that involved radiation. The unprecedented release of multiple different radioisotopes led to radioactive contamination of large areas surrounding the accident site. The exposure of the residents of these areas was varied and therefore the consequences for health and radioecology could not be reliably estimated quickly. Even though some studies have now been ongoing for 25 years and have provided a better understanding of the situation, these are yet neither complete nor comprehensive enough to determine the long-term risk. A true assessment can only be provided after following the observed population for their natural lifespan. Here we review the technical aspects of the accident and provide relevant information on radioactive releases that resulted in exposure of this large population to radiation. A number of different groups of people were exposed to radiation: workers involved in the initial clean-up response, and members of the general population who were either evacuated from the settlements in the Chernobyl nuclear power plant vicinity shortly after the accident, or continued to live in the affected territories of Belarus, Russia and Ukraine. Through domestic efforts and extensive international co-operation, essential information on radiation dose and health status for this population has been collected. This has permitted the identification of high-risk groups and the use of more specialised means of collecting information, diagnosis, treatment and follow-up. Because radiation-associated thyroid cancer is one of the major health consequences of the Chernobyl accident, a particular emphasis is placed on this malignancy. The initial epidemiological studies are reviewed, as are the most significant studies and/or aid programmes in the three affected countries. Copyright © 2011 The Royal College of Radiologists. Published by Elsevier Ltd. All rights reserved.

EMERALD REV. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1974-03-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Root causes and impacts of severe accidents at large nuclear power plants.

PubMed

Högberg, Lars

2013-04-01

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Loss of DHR sequences at Browns Ferry Unit One - accident-sequence analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cook, D.H.; Grene, S.R.; Harrington, R.M.

1983-05-01

This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated loss of decay heat removal (DHR) capability following scram from full power with the power conversion system unavailable. In accident sequences without DHR capability, the residual heat removal (RHR) system functions of pressure suppression pool cooling and reactor vessel shutdown cooling are unavailable. Consequently, all decay heat energy is stored in the pressure suppression pool with a concomitant increase in pool temperature and primary containment pressure. With the assumption that DHR capability is not regained during the lengthy course of this accidentmore » sequence, the containment ultimately fails by overpressurization. Although unlikely, this catastrophic failure might lead to loss of the ability to inject cooling water into the reactor vessel, causing subsequent core uncovery and meltdown. The timing of these events and the effective mitigating actions that might be taken by the operator are discussed in this report.« less

Estimated rate of fatal automobile accidents attributable to acute solvent exposure at low inhaled concentrations.

PubMed

Benignus, Vernon A; Bushnell, Philip J; Boyes, William K

2011-12-01

Acute solvent exposures may contribute to automobile accidents because they increase reaction time and decrease attention, in addition to impairing other behaviors. These effects resemble those of ethanol consumption, both with respect to behavioral effects and neurological mechanisms. These observations, along with the extensive data on the relationship between ethanol consumption and fatal automobile accidents, suggested a way to estimate the probability of fatal automobile accidents from solvent inhalation. The problem can be approached using the logic of the algebraic transitive postulate of equality: if A=B and B=C, then A=C. We first calculated a function describing the internal doses of solvent vapors that cause the same magnitude of behavioral impairment as ingestion of ethanol (A=B). Next, we fit a function to data from the literature describing the probability of fatal car crashes for a given internal dose of ethanol (B=C). Finally, we used these two functions to generate a third function to estimate the probability of a fatal car crash for any internal dose of organic solvent vapor (A=C). This latter function showed quantitatively (1) that the likelihood of a fatal car crash is increased by acute exposure to organic solvent vapors at concentrations less than 1.0 ppm, and (2) that this likelihood is similar in magnitude to the probability of developing leukemia from exposure to benzene. This approach could also be applied to other potentially adverse consequences of acute exposure to solvents (e.g., nonfatal car crashes, property damage, and workplace accidents), if appropriate data were available. © 2011 Society for Risk Analysis Published 2011. This article is a U.S. Government work and is in the public domain for the U.S.A.

Nuclear Fuels & Materials Spotlight Volume 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

Assessment of risk due to the use of carbon fiber composites in commercial and general aviation

NASA Technical Reports Server (NTRS)

Fiksel, J.; Rosenfield, D.; Kalelkar, A.

1980-01-01

The development of a national risk profile for the total annual aircraft losses due to carbon fiber composite (CFC) usage through 1993 is discussed. The profile was developed using separate simulation methods for commercial and general aviation aircraft. A Monte Carlo method which was used to assess the risk in commercial aircraft is described. The method projects the potential usage of CFC through 1993, investigates the incidence of commercial aircraft fires, models the potential release and dispersion of carbon fibers from a fire, and estimates potential economic losses due to CFC damaging electronic equipment. The simulation model for the general aviation aircraft is described. The model emphasizes variations in facility locations and release conditions, estimates distribution of CFC released in general aviation aircraft accidents, and tabulates the failure probabilities and aggregate economic losses in the accidents.

As assessment of power system vulnerability to release of carbon fibers during commercial aviation accidents

NASA Technical Reports Server (NTRS)

Larocque, G. R.

1980-01-01

The vulnerability of a power distribution system in Bedford and Lexington, Massachusetts to power outages as a result of exposure to carbon fibers released in a commercial aviation accident in 1993 was examined. Possible crash scenarios at Logan Airport based on current operational data and estimated carbon fiber usage levels were used to predict exposure levels and occurrence probabilities. The analysis predicts a mean time between carbon fiber induced power outages of 2300 years with an expected annual consequence of 0.7 persons losing power. In comparison to historical outage data for the system, this represents a 0.007% increase in outage rate and 0.07% increase in consequence.

French policy for managing the post-accident phase of a nuclear accident.

PubMed

Gallay, F; Godet, J L; Niel, J C

2015-06-01

In 2005, at the request of the French Government, the Nuclear Safety Authority (ASN) established a Steering Committee for the Management of the Post-Accident Phase of a Nuclear Accident or a Radiological Emergency, with the objective of establishing a policy framework. Under the supervision of ASN, this Committee, involving several tens of experts from different backgrounds (e.g. relevant ministerial offices, expert agencies, local information commissions around nuclear installations, non-governmental organisations, elected officials, licensees, and international experts), developed a number of recommendations over a 7-year period. First published in November 2012, these recommendations cover the immediate post-emergency situation, and the transition and longer-term periods of the post-accident phase in the case of medium-scale nuclear accidents causing short-term radioactive release (less than 24 h) that might occur at French nuclear facilities. They also apply to actions to be undertaken in the event of accidents during the transportation of radioactive materials. These recommendations are an important first step in preparation for the management of a post-accident situation in France in the case of a nuclear accident. © The Chartered Institution of Building Services Engineers 2014.

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

Tailoring the response of Autonomous Reactivity Control (ARC) systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qvist, Staffan A.; Hellesen, Carl; Gradecka, Malwina

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, weremore » performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.« less

Two examples of industrial applications of shock physics research

NASA Astrophysics Data System (ADS)

Sanai, Mohsen

1996-05-01

An in-depth understanding of shock physics phenomena has led to many industrial applications. Two recent applications discussed in this paper are a method for assessing explosion safety in industrial plants and a bomb-resistant luggage container for widebody aircraft. Our explosion safety assessment is based on frequent use of computer simulation of postulated accidents to model in detail the detonation of energetic materials, the formation and propagation of the resulting airblast, and the projection of fragments of known material and mass. Using a general load-damage analysis technique referred to as the pressure-impulse (PI) method, we have developed a PC-based computer algorithm that includes a continually expanding library of PI load and damage curves, which can predict and graphically display common structural damage modes and the response of humans to postulated explosion accidents. A second commercial application of shock physics discussed here is a bomb-resistant luggage container for widebody aircraft that can protect the aircraft from a terrorist bomb hidden inside the luggage. This hardened luggage container (HLC) relies on blast management and debris containment provided by a flexible flow-through blanket woven from threads made with a strong lightweight material, such as Spectra or Kevlar. This mitigation blanket forms a continuous and seamless shell around the sides of the luggage container that are parallel to the aircraft axis, leaving the two ends of the container unprotected. When an explosion occurs, the mitigation blanket expands into a nearly circular shell that contains the flying debris while directing the flow into the adjacent containers. The HLC concept has been demonstrated through full-scale experiments conducted at SRI. We believe that these two examples represent a broad class of potential industrial hazard applications of the experimental, analytical, and computational tools possessed by the shock physics community.

Persons with Mild or Moderate Alzheimer's Disease Learn to Use Urine Alarms and Prompts to Avoid Large Urinary Accidents

ERIC Educational Resources Information Center

Lancioni, Giulio E.; Singh, Nirbhay N.; O'Reilly, Mark F.; Sigafoos, Jeff; Bosco, Andrea; Zonno, Nadia; Badagliacca, Francesco

2011-01-01

This study assessed whether three patients with Alzheimer's disease could learn to use urine alarms and caregivers' prompts to eliminate large urinary accidents. As soon as the patient began to release urine, the alarm system presented auditory and vibratory signals. In relation to those signals, the caregiver would prompt/encourage the patient to…

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

PubMed

Zhang, X L; Su, G F; Yuan, H Y; Chen, J G; Huang, Q Y

2014-09-15

Atmospheric dispersion models play an important role in nuclear power plant accident management. A reliable estimation of radioactive material distribution in short range (about 50 km) is in urgent need for population sheltering and evacuation planning. However, the meteorological data and the source term which greatly influence the accuracy of the atmospheric dispersion models are usually poorly known at the early phase of the emergency. In this study, a modified ensemble Kalman filter data assimilation method in conjunction with a Lagrangian puff-model is proposed to simultaneously improve the model prediction and reconstruct the source terms for short range atmospheric dispersion using the off-site environmental monitoring data. Four main uncertainty parameters are considered: source release rate, plume rise height, wind speed and wind direction. Twin experiments show that the method effectively improves the predicted concentration distribution, and the temporal profiles of source release rate and plume rise height are also successfully reconstructed. Moreover, the time lag in the response of ensemble Kalman filter is shortened. The method proposed here can be a useful tool not only in the nuclear power plant accident emergency management but also in other similar situation where hazardous material is released into the atmosphere. Copyright © 2014 Elsevier B.V. All rights reserved.

Risk to the public from carbon fibers released in civil aircraft accidents

NASA Technical Reports Server (NTRS)

1980-01-01

Because carbon fibers are strong, stiff, and lightweight, they are attractive for use in composite structures. Because they also have high electrical conductivity, free carbon fibers settling on electrical conductors can cause malfunctions. If released from the composite by burning, the fibers may become a hazard to exposed electrical and electronic equipment. As part of a Federal study of the potential hazard associated with the use of carbon fibers, NASA assessed the public risk associated with crash fire accidents of civil aircraft. The NASA study projected a dramatic increase in the use of carbon composites in civil aircraft and developed technical data to support the risk assessment. Personal injury was found to be extremely unlikely. In 1993, the year chosen as a focus for the study, the expected annual cost of damage caused by released carbon fibers is only $1000. Even the worst-case carbon fiber incident simulated (costing $178,000 once in 34,000 years) was relatively low-cost compared with the usual air transport accident cost. On the basis of these observations, the NASA study concluded that exploitation of composites should continue, that additional protection of avionics is unnecessary, and that development of alternate materials specifically to overcome this problem is not justified.

A Review of Criticality Accidents 2000 Revision

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost

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. Themore » second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.« less

Levels, distribution and bioavailability of transuranic elements released in the Palomares accident (Spain).

PubMed

Jiménez-Ramos, M C; Vioque, I; García-Tenorio, R; García León, M

2008-11-01

The current levels and distribution of the remaining transuranic contamination present in the terrestrial area affected by the nuclear Palomares accident have been evaluated through the determination of the Pu-isotopes and (241)Am concentrations in soils collected 35 years after the accident. In addition, after confirming that most of the contamination is present in particulate form, some bioavailability laboratory-based experiments, based on the use of single extractants, were performed as an essential step in order to study the behaviour of the Pu contamination in the soils from the affected areas.

Rescue complex for coal mines

NASA Astrophysics Data System (ADS)

Yungmeyster, D. A.; Urazbakhtin, R. Yu

2017-10-01

The mining industry was potentially dangerous at all times, even with the use of modern equipment in mines, accidents continue to occur, including catastrophic ones. Accidents in mines are due to the presence of specific features in the conduct of mining operations. These include the inconsistency of mining and geological conditions, the contamination of the mine atmosphere due to the release of gases from minerals, the presence of self-igniting coal strata, which creates the danger of underground fires, gas explosions. The main cause of accidents is the irresponsibility of both the manager and the personnel who violate the safety rules during mining operations.

Radioxenon monitoring in Beijing following the Fukushima Daiichi NPP accident.

PubMed

Shilian, Wang; Qi, Li; Qinghua, Meng; Zhanying, Chen; Yungang, Zhao; Huijuan, Li; Huaimao, Jia; Yinzhong, Chang; Shujiang, Liu; Xinjun, Zhang; Yuanqing, Fan; Ling, Wan; Yun, Lou

2013-11-01

This paper reports the brief process and results of radioxenon monitoring and analysis in Beijing following the Fukushima Daiichi nuclear power plant accident. The accident and release of volatile radionuclides were caused by 9.0 magnitude earthquake and tsunami on March 11, 2011. The maximum concentrations of (133)Xe and (131 m)Xe were in excess of 0.90 Bq.m(-3) and 0.047 Bq.m(-3), respectively. The activity ratio of (131 m)Xe to (133)Xe and the dynamic trend of (133)Xe activity concentration were analyzed. Copyright © 2013 Elsevier Ltd. All rights reserved.

Source Term Estimation of Radioxenon Released from the Fukushima Dai-ichi Nuclear Reactors Using Measured Air Concentrations and Atmospheric Transport Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eslinger, Paul W.; Biegalski, S.; Bowyer, Ted W.

2014-01-01

Systems designed to monitor airborne radionuclides released from underground nuclear explosions detected radioactive fallout from the Fukushima Daiichi nuclear accident in March 2011. Atmospheric transport modeling (ATM) of plumes of noble gases and particulates were performed soon after the accident to determine plausible detection locations of any radioactive releases to the atmosphere. We combine sampling data from multiple International Modeling System (IMS) locations in a new way to estimate the magnitude and time sequence of the releases. Dilution factors from the modeled plume at five different detection locations were combined with 57 atmospheric concentration measurements of 133-Xe taken from Marchmore » 18 to March 23 to estimate the source term. This approach estimates that 59% of the 1.24×1019 Bq of 133-Xe present in the reactors at the time of the earthquake was released to the atmosphere over a three day period. Source term estimates from combinations of detection sites have lower spread than estimates based on measurements at single detection sites. Sensitivity cases based on data from four or more detection locations bound the source term between 35% and 255% of available xenon inventory.« less

How "lucky" we are that the Fukushima disaster occurred in early spring: predictions on the contamination levels from various fission products released from the accident and updates on the risk assessment for solid and thyroid cancers.

PubMed

Evangeliou, Nikolaos; Balkanski, Yves; Cozic, Anne; Møller, Anders Pape

2014-12-01

The present paper studies how a random event (earthquake) and the subsequent disaster in Japan affect transport and deposition of fallout and the resulting health consequences. Therefore, except for the original accident in March 2011, three additional scenarios are assessed assuming that the same releases took place in winter 2010, summer 2011 and autumn 2011 in order to cover a full range of annual seasonality. This is also the first study where a large number of fission products released from the accident are used to assess health risks with the maximum possible efficiency. Xenon-133 and (137)Cs are directly estimated within the model, whereas 15 other radionuclides are calculated indirectly using reported isotopic ratios. As much as 85% of the released (137)Cs would be deposited in continental regions worldwide if the accident occurred in winter 2010, 22% in spring 2011 (when it actually happened), 55% in summer 2011 and 48% if it occurred during autumn 2011. Solid cancer incidents and mortalities from Fukushima are estimated to be between 160 and 880 and from 110 to 640 close to previous estimations. By adding thyroid cancers, the total number rises from 230 to 850 for incidents and from 120 to 650 for mortalities. Fatalities due to worker exposure and mandatory evacuation have been reported to be around 610 increasing total estimated mortalities to 730-1260. These estimates are 2.8 times higher than previously reported ones for radiocaesium and (131)I and 16% higher than those reported based on radiocaesium only. Total expected fatalities from Fukushima are 32% lower than in the winter scenario, 5% that in the summer scenario and 30% lower than in the autumn scenario. Nevertheless, cancer fatalities are expected to be less than 5% of those from the tsunami (~20,000). Copyright © 2014 Elsevier B.V. All rights reserved.

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

Consequences of the radiation accident at the Mayak production association in 1957 (the 'Kyshtym Accident').

PubMed

Akleyev, A V; Krestinina, L Yu; Degteva, M O; Tolstykh, E I

2017-09-01

This paper presents an overview of the nuclear accident that occurred at the Mayak Production Association (PA) in the Russian Federation on 29 September 1957, often referred to as 'Kyshtym Accident', when 20 MCi (740 PBq) of radionuclides were released by a chemical explosion in a radioactive waste storage tank. 2 MCi (74 PBq) spread beyond the Mayak PA site to form the East Urals Radioactive Trace (EURT). The paper describes the accident and gives brief characteristics of the efficacy of the implemented protective measures that made it possible to considerably reduce doses to the exposed population. The paper also provides retrospective dosimetry estimates for the members of the EURT Cohort (EURTC) which comprises approximately 21 400 people. During the first two years after the accident a decrease in the group average leukocyte (mainly due to neutrophils and lymphocytes) and thrombocyte count was observed in the population. At later dates an increased excess relative risk of solid cancer incidence and mortality was found in the EURTC.

Can Vitex Agnus Castus be Used for the Treatment of Mastalgia? What is the Current Evidence?

PubMed Central

2008-01-01

There have been many treatments suggested for the management of mastalgia; one of these is the fruit extract of Vitex Agnus castus L. commonly known as Agnus castus, an extract of a deciduous shrub native to Mediterranean Europe and Central Asia. It is postulated that A. castus suppresses the stress-induced latent hyperprolactinemia which is a release of supra-physiological levels of prolactin in some patients in response to stressful stimuli. It is postulated that A. castus could be effective in the treatment of cyclical mastalgia by inhibiting the release of excess prolactin by blocking Dopamine-2 receptor type on pituitary. The adverse events following A. castus treatment are mild and reversible. The aim of this review is assess the efficacy of A. castus in the treatment of mastalgia. Data from randomized and non-randomized studies regarding the efficacy and safety of A. castus is reviewed in a systematic fashion. It is concluded that A. castus can be considered as an efficient alternative phytotherapeutic agent in the treatment of mastalgia. PMID:18830450

Can Vitex Agnus Castus be Used for the Treatment of Mastalgia? What is the Current Evidence?

PubMed

Carmichael, A R

2008-09-01

There have been many treatments suggested for the management of mastalgia; one of these is the fruit extract of Vitex Agnus castus L. commonly known as Agnus castus, an extract of a deciduous shrub native to Mediterranean Europe and Central Asia. It is postulated that A. castus suppresses the stress-induced latent hyperprolactinemia which is a release of supra-physiological levels of prolactin in some patients in response to stressful stimuli. It is postulated that A. castus could be effective in the treatment of cyclical mastalgia by inhibiting the release of excess prolactin by blocking Dopamine-2 receptor type on pituitary. The adverse events following A. castus treatment are mild and reversible. The aim of this review is assess the efficacy of A. castus in the treatment of mastalgia. Data from randomized and non-randomized studies regarding the efficacy and safety of A. castus is reviewed in a systematic fashion. It is concluded that A. castus can be considered as an efficient alternative phytotherapeutic agent in the treatment of mastalgia.

Regulatory Information by Topic: Emergency Management

EPA Pesticide Factsheets

Regulatory information about emergencies, including chemical accident prevention, risk management plans (RMPs), chemical reporting, community right to know, and oil spills and hazardous substances releases.

76 FR 800 - Ameren Missouri, Combined License Application for Callaway Plant Unit 2; Environmental Assessment...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-06

... significantly increase the probability or consequences of accidents. No changes are being made in the types of effluents that may be released offsite. There is no significant increase in the amount of any effluent released offsite. There is no significant increase in occupational or public radiation exposure. Therefore...

75 FR 61226 - Exemption; Entergy Operations, Inc.; Arkansas Nuclear One, Units 1 and 2

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-04

... maximum potential annual radiation doses to the public resulting from effluent releases. The report must... radiation doses to the public resulting from effluent releases. This exemption does not affect the... submitted. Based on the above, no new accident precursors are created by extending the submittal date for...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

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 amore » 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.« less

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

ScienceCinema

None

2018-01-16

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.

First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations.

PubMed

Tsuruta, Haruo; Oura, Yasuji; Ebihara, Mitsuru; Ohara, Toshimasa; Nakajima, Teruyuki

2014-10-22

No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric (137)Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12-23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric (137)Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models.

Offsite radiological consequence analysis for the bounding flammable gas accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

CARRO, C.A.

2003-03-19

The purpose of this analysis is to calculate the offsite radiological consequence of the bounding flammable gas accident. DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', requires the formal quantification of a limited subset of accidents representing a complete set of bounding conditions. The results of these analyses are then evaluated to determine if they challenge the DOE-STD-3009-94, Appendix A, ''Evaluation Guideline,'' of 25 rem total effective dose equivalent in order to identify and evaluate safety class structures, systems, and components. The bounding flammable gas accident is a detonation in a single-shell tank (SST).more » A detonation versus a deflagration was selected for analysis because the faster flame speed of a detonation can potentially result in a larger release of respirable material. As will be shown, the consequences of a detonation in either an SST or a double-shell tank (DST) are approximately equal. A detonation in an SST was selected as the bounding condition because the estimated respirable release masses are the same and because the doses per unit quantity of waste inhaled are generally greater for SSTs than for DSTs. Appendix A contains a DST analysis for comparison purposes.« less

Safety distance assessment of industrial toxic releases based on frequency and consequence: a case study in Shanghai, China.

PubMed

Yu, Q; Zhang, Y; Wang, X; Ma, W C; Chen, L M

2009-09-15

A case study on the safety distance assessment of a chemical industry park in Shanghai, China, is presented in this paper. Toxic releases were taken into consideration. A safety criterion based on frequency and consequence of major hazard accidents was set up for consequence analysis. The exposure limits for the accidents with the frequency of more than 10(-4), 10(-5)-10(-4) and 10(-6)-10(-5) per year were mortalities of 1% (or SLOT), 50% (SLOD) and 75% (twice of SLOD) respectively. Accidents with the frequency of less than 10(-6) per year were considered incredible and ignored in the consequence analysis. Taking the safety distance of all the hazard installations in a chemical plant into consideration, the results based on the new criterion were almost smaller than those based on LC50 or SLOD. The combination of the consequence and risk based results indicated that the hazard installations in two of the chemical plants may be dangerous to the protection targets and measurements had to be taken to reduce the risk. The case study showed that taking account of the frequency of occurrence in the consequence analysis would give more feasible safety distances for major hazard accidents and the results were more comparable to those calculated by risk assessment.

Fukushima Daiichi Nuclear Plant accident: Atmospheric and oceanic impacts over the five years.

PubMed

Hirose, Katsumi

2016-06-01

The Fukushima Daiichi Nuclear Plant (FDNPP) accident resulted in huge environmental and socioeconomic impacts to Japan. To document the actual environmental and socioeconomic effects of the FDNPP accident, we describe here atmospheric and marine contamination due to radionuclides released from the FDNPP accident using papers published during past five years, in which temporal and spatial variations of FDNPP-derived radionuclides in air, deposition and seawater and their mapping are recorded by local, regional and global monitoring activities. High radioactivity-contaminated area in land were formed by the dispersion of the radioactive cloud and precipitation, depending on land topography and local meteorological conditions, whereas extremely high concentrations of (131)I and radiocesium in seawater occurred due to direct release of radioactivity-contaminated stagnant water in addition to atmospheric deposition. For both of atmosphere and ocean, numerical model simulations, including local, regional and global-scale modeling, were extensively employed to evaluate source terms of the FDNPP-derived radionuclides from the monitoring data. These models also provided predictions of the dispersion and high deposition areas of the FDNPP-derived radionuclides. However, there are significant differences between the observed and simulated values. Then, the monitoring data would give a good opportunity to improve numerical modeling. Copyright © 2016 Elsevier Ltd. All rights reserved.

Predicting Consequences of Technological Disasters from Natural Hazard Events: Challenges and Opportunities Associated with Industrial Accident Data Sources

NASA Astrophysics Data System (ADS)

Wood, M.

2009-04-01

The increased focus on the possibility of technological accidents caused by natural events (Natech) is foreseen to continue for years to come. In this case, experts in prevention, mitigation and preparation activities associated with natural events will increasingly need to borrow data and expertise traditionally associated with the technological fields to carry out the work. An important question is how useful is the data for understanding consequences from such natech events. Data and case studies provided on major industrial accidents tend to focus on lessons learned for re-engineering the process. While consequence data are reported at least nominally in most reports, their precision, quality and completeness is often lacking. Consequences that are often or sometimes available but not provided can include severity and type of injuries, distance of victims from the source, exposure measurements, volume of the release, population in potentially affected zones, and weather conditions. Yet these are precisely the type of data that will aid natural hazard experts in land-use planning and emergency response activities when a Natech event may be foreseen. This work discusses the results of a study of consequence data from accidents involving toxic releases reported in the EU's MARS accident database. The study analysed the precision, quality and completeness of three categories of consequence data reported: the description of health effects, consequence assessment and chemical risk assessment factors, and emergency response information. This work reports on the findings from this study and discusses how natural hazards experts might interact with industrial accident experts to promote more consistent and accurate reporting of the data that will be useful in consequence-based activities.

The magnitude and relevance of the February 2014 radiation release from the Waste Isolation Pilot Plant repository in New Mexico, USA.

PubMed

Thakur, P; Lemons, B G; White, C R

2016-09-15

After almost fifteen years of successful waste disposal operations, the first unambiguous airborne radiation release from the Waste Isolation Pilot Plant (WIPP) was detected beyond the site boundary on February 14, 2014. It was the first accident of its kind in the 15-year operating history of the WIPP. The accident released moderate levels of radioactivity into the underground air. A small but measurable amount of radioactivity also escaped to the surface through the ventilation system and was detected above ground. The dominant radionuclides released were americium and plutonium, in a ratio consistent with the known content of a breached drum. The radiation release was caused by a runaway chemical reaction inside a transuranic (TRU) waste drum which experienced a seal and lid failure, spewing radioactive materials into the repository. According to source-term estimation, approximately 2 to 10Ci of radioactivity was released from the breached drum into the underground, and an undetermined fraction of that source term became airborne, setting off an alarm and triggering the closure of seals designed to force exhausting air through a system of filters including high-efficiency-particulate-air (HEPA) filters. Air monitoring across the WIPP site intensified following the first reports of radiation detection underground to determine the extent of impact to WIPP personnel, the public, and the environment, if any. This article attempts to compile and interpret analytical data collected by an independent monitoring program conducted by the Carlsbad Environmental Monitoring & Research Center (CEMRC) and by a compliance-monitoring program conducted by the WIPP's management and operating contractor, the Nuclear Waste Partnership (NWP), LLC., in response to the accident. Both the independent and the WIPP monitoring efforts concluded that the levels detected were very low and localized, and no radiation-related health effects among local workers or the public would be expected. Published by Elsevier B.V.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 safetymore » 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)« less

The atmospheric transport of iodine-129 from Fukushima to British Columbia, Canada and its deposition and transport into groundwater

NASA Astrophysics Data System (ADS)

Herod, Matt N.; Suchy, Martin; Cornett, R. Jack; Kieser, W. E.; Clark, Ian D.; Graham, Gwyn

2015-12-01

The Fukushima-Daiichi nuclear accident (FDNA) released iodine-129 (15.7 million year half-life) and other fission product radionuclides into the environment in the spring and summer of 2011. 129I is recognized as a useful tracer for the short-lived radiohazard 131I, which has a mobile geochemical behavior with potential to contaminate water resources. To trace 129I released by the FDNA reaching Canada, pre-accident and post-accident rain samples collected in Vancouver, on Saturna Island and from the National Atmospheric Deposition Program in Washington State were measured. Groundwater from the Abbotsford-Sumas Aquifer was sampled to determine the fate of 129I that infiltrates below the root zone. Modeling of vadose zone transport was performed to constrain the travel time and retardation of 129I. The mean pre-accident 129I concentration in rain was 31 × 106 atoms/L (n = 4). Immediately following the FDNA, 129I values increased to 211 × 106 atoms/L and quickly returned to near-background levels. However, pulses of elevated 129I continued for several months. The increases in 129I concentrations from both Vancouver and Saturna Island were synchronized, and occurred directly after the initial release from the FDNA. The 129I in shallow (3H/3He age <1.4 years) Wassenaar et al. (2006) groundwater showed measurable variability through March 2013 with an average of 3.2 × 106 atoms/L (n = 32) that was coincident with modeled travel times for Fukushima 129I. The groundwater response and the modeling results suggest that 129I was partially attenuated in soil, which is consistent with its geochemical behavior; however, we conclude that the measured variability may be due to Fukushima 129I entering groundwater.

RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

KOZLOWSKI, S.D.

2007-05-30

This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditionsmore » for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below.« less

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2012 CFR

2012-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2013 CFR

2013-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2011 CFR

2011-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2010 CFR

2010-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2014 CFR

2014-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

Third generation nuclear plants

NASA Astrophysics Data System (ADS)

Barré, Bertrand

2012-05-01

After the Chernobyl accident, a new generation of Light Water Reactors has been designed and is being built. Third generation nuclear plants are equipped with dedicated systems to insure that if the worst accident were to occur, i.e. total core meltdown, no matter how low the probability of such occurrence, radioactive releases in the environment would be minimal. This article describes the EPR, representative of this "Generation III" and a few of its competitors on the world market.

Evaluation of external hazards to nuclear power plants in the United States

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kimura, C.Y.; Budnitz, R.J.

1987-12-01

The Lawrence Livermore National Laboratory (LLNL) has performed a study of the risk of core damage to nuclear power plants in the United States due to externally initiated events. The broad objective has been to gain an understanding of whether or not each external initiator is among the major potential accident initiators that may pose a threat of severe reactor core damage or of large radioactive release to the environment from the reactor. Four external hazards were investigated in this report. These external hazards are internal fires, high winds/tornadoes, external floods, and transportation accidents. Analysis was based on two figures-of-merit,more » one based on core damage frequency and the other based on the frequency of large radioactive releases. Using these two figures-of-merit as evaluation criteria, it has been feasible to ascertain whether the risk from externally initiated accidents is, or is not, an important contributor to overall risk for the US nuclear power plants studied. This has been accomplished for each initiator separately. 208 refs., 17 figs., 45 tabs.« less

Analysis of Radionuclide Releases from the Fukushima Dai-Ichi Nuclear Power Plant Accident Part I

NASA Astrophysics Data System (ADS)

Le Petit, G.; Douysset, G.; Ducros, G.; Gross, P.; Achim, P.; Monfort, M.; Raymond, P.; Pontillon, Y.; Jutier, C.; Blanchard, X.; Taffary, T.; Moulin, C.

2014-03-01

Part I of this publication deals with the analysis of fission product releases consecutive to the Fukushima Dai-ichi accident. Reactor core damages are assessed relying on radionuclide detections performed by the CTBTO radionuclide network, especially at the particulate station located at Takasaki, 210 km away from the nuclear power plant. On the basis of a comparison between the reactor core inventory at the time of reactor shutdowns and the fission product activities measured in air at Takasaki, especially 95Nb and 103Ru, it was possible to show that the reactor cores were exposed to high temperature for a prolonged time. This diagnosis was confirmed by the presence of 113Sn in air at Takasaki. The 133Xe assessed release at the time of reactor shutdown (8 × 1018 Bq) turned out to be in the order of 80 % of the amount deduced from the reactor core inventories. This strongly suggests a broad meltdown of reactor cores.

Operational level for unconditional release of contaminated property from affected areas around Fukushima Daiichi nuclear power plant

PubMed Central

Ogino, Haruyuki; Hattori, Takatoshi

2013-01-01

This paper focuses on the surface contamination control of slightly contaminated property after the Fukushima nuclear accident. The operational level for the unconditional release of contaminated properties is calculated in counts per minute (cpm) to enable the use of a typical Geiger-Muller (GM) survey meter with a 50-mm bore, on the basis of the surficial clearance level of 10 Bq cm−2 for 134Cs and 137Cs derived in the previous studies of the authors. By applying a factor for the conversion of the unit surface contamination to the count rate of a survey meter widely used after the Fukushima accident, the operational level for the unconditional release of contaminated properties was calculated to be 2300 cpm on average and 23 000 cpm at the highest-contamination part. The calculated numerical values of the operational levels are effective as long as the typical GM survey meter is used in the radiation measurement. PMID:23778575

Fukushima Accident: Sequence of Events and Lessons Learned

NASA Astrophysics Data System (ADS)

Morse, Edward C.

2011-10-01

The Fukushima Dai-Ichi nuclear power station suffered a devastating Richter 9.0 earthquake followed by a 14.0 m tsunami on 11 March 2011. The subsequent loss of power for emergency core cooling systems resulted in damage to the fuel in the cores of three reactors. The relief of pressure from the containment in these three reactors led to sufficient hydrogen gas release to cause explosions in the buildings housing the reactors. There was probably subsequent damage to a spent fuel pool of a fourth reactor caused by debris from one of these explosions. Resultant releases of fission product isotopes in air were significant and have been estimated to be in the 3 . 7 --> 6 . 3 ×1017 Bq range (~10 MCi) for 131I and 137Cs combined, or approximately one tenth that of the Chernobyl accident. A synopsis of the sequence of events leading up to this large release of radioactivity will be presented, along with likely scenarios for stabilization and site cleanup in the future. Some aspects of the isotope monitoring programs, both locally and at large, will also be discussed. An assessment of radiological health risk for the plant workers as well as the general public will also be presented. Finally, the impact of this accident on design and deployment of nuclear generating stations in the future will be discussed.

Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Released to the Atmosphere from Lawrence Livermore National Laboratory (LLNL). Part 5. Accidental Releases

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, S

2007-08-15

Over the course of fifty-three years, LLNL had six acute releases of tritiated hydrogen gas (HT) and one acute release of tritiated water vapor (HTO) that were too large relative to the annual releases to be included as part of the annual releases from normal operations detailed in Parts 3 and 4 of the Tritium Dose Reconstruction (TDR). Sandia National Laboratories/California (SNL/CA) had one such release of HT and one of HTO. Doses to the maximally exposed individual (MEI) for these accidents have been modeled using an equation derived from the time-dependent tritium model, UFOTRI, and parameter values based onmore » expert judgment. All of these acute releases are described in this report. Doses that could not have been exceeded from the large HT releases of 1965 and 1970 were calculated to be 43 {micro}Sv (4.3 mrem) and 120 {micro}Sv (12 mrem) to an adult, respectively. Two published sets of dose predictions for the accidental HT release in 1970 are compared with the dose predictions of this TDR. The highest predicted dose was for an acute release of HTO in 1954. For this release, the dose that could not have been exceeded was estimated to have been 2 mSv (200 mrem), although, because of the high uncertainty about the predictions, the likely dose may have been as low as 360 {micro}Sv (36 mrem) or less. The estimated maximum exposures from the accidental releases were such that no adverse health effects would be expected. Appendix A lists all accidents and large routine puff releases that have occurred at LLNL and SNL/CA between 1953 and 2005. Appendix B describes the processes unique to tritium that must be modeled after an acute release, some of the time-dependent tritium models being used today, and the results of tests of these models.« less

Distribution of oceanic 137Cs from the Fukushima Dai-ichi Nuclear Power Plant simulated numerically by a regional ocean model.

PubMed

Tsumune, Daisuke; Tsubono, Takaki; Aoyama, Michio; Hirose, Katsumi

2012-09-01

Radioactive materials were released to the environment from the Fukushima Dai-ichi Nuclear Power Plant as a result of the reactor accident after the Tohoku earthquake and tsunami of 11 March 2011. The measured (137)Cs concentration in a seawater sample near the Fukushima Dai-ichi Nuclear Power Plant site reached 68 kBq L(-1) (6.8 × 10(4)Bq L(-1)) on 6 April. The two major likely pathways from the accident site to the ocean existed: direct release of high radioactive liquid wastes to the ocean and the deposition of airborne radioactivity to the ocean surface. By analysis of the (131)I/(137)Cs activity ratio, we determined that direct release from the site contributed more to the measured (137)Cs concentration than atmospheric deposition did. We then used a regional ocean model to simulate the (137)Cs concentrations resulting from the direct release to the ocean off Fukushima and found that from March 26 to the end of May the total amount of (137)Cs directly released was 3.5 ± 0.7 PBq ((3.5 ± 0.7) × 10(15)Bq). The simulated temporal change in (137)Cs concentrations near the Fukushima Daini Nuclear Power Plant site agreed well with observations. Our simulation results showed that (1) the released (137)Cs advected southward along the coast during the simulation period; (2) the eastward-flowing Kuroshio and its extension transported (137)C during May 2011; and (3) (137)Cs concentrations decreased to less than 10 BqL(-1) by the end of May 2011 in the whole simulation domain as a result of oceanic advection and diffusion. We compared the total amount and concentration of (137)Cs released from the Fukushima Dai-ichi reactors to the ocean with the (137)Cs released to the ocean by global fallout. Even though the measured (137)Cs concentration from the Fukushima accident was the highest recorded, the total released amount of (137)Cs was not very large. Therefore, the effect of (137)Cs released from the Fukushima Dai-ichi reactors on concentration in the whole North Pacific was smaller than that of past release events such as global fallout, and the amount of (137)Cs expected to reach other oceanic basins is negligible comparing with the past radioactive input. Copyright © 2011 Elsevier Ltd. All rights reserved.

Report of Apollo 204 Review Board to the Administrator, National Aeronautics and Space Administration . Appendix F ; Schedule of Physical Evidence

NASA Technical Reports Server (NTRS)

1967-01-01

Immediately following the Apollo 204 accident of January 27, 1961. all associated equipment and material were impounded. Release of this equipment and material for normal use was under the close control of the Apollo 204 Review Board. Apollo Review Board Administrative Procedure No. 11, February 11, 1961, established the Apollo 204 Review Board Material Release Record (MRR). This MRR was the official form used to release material from full impoundment and was valid only after being approved by the Board and signed by a Member. The form was used as the authority to place any impounded item into one of the three Categories defined in Administrative Procedure No. 11. This appendix contains all of the authorized MRR's. Each item submitted on an MRR was given a control number; a description, including the part number and serial number; the relevance and location to the accident; any constraints before release; and the control category. The categories placed on the equipment were as follows: Category A - Items which may have a significant influence or bearing on the results or findings of the Apollo 204 Review Board; Category B - All material other than Category A which is considered relevant to the Apollo 204 Review Board investigation; Category C - Material released from Board jurisdiction. Several classes of equipment were released by special Board action prior to the establishment of the MRR system. The operating procedure for release of these classes is Enclosure F-l to this appendix.

Traits and causes of environmental loss-related chemical accidents in China based on co-word analysis.

PubMed

Wu, Desheng; Song, Yu; Xie, Kefan; Zhang, Baofeng

2018-04-25

Chemical accidents are major causes of environmental losses and have been debated due to the potential threat to human beings and environment. Compared with the single statistical analysis, co-word analysis of chemical accidents illustrates significant traits at various levels and presents data into a visual network. This study utilizes a co-word analysis of the keywords extracted from the Web crawling texts of environmental loss-related chemical accidents and uses the Pearson's correlation coefficient to examine the internal attributes. To visualize the keywords of the accidents, this study carries out a multidimensional scaling analysis applying PROXSCAL and centrality identification. The research results show that an enormous environmental cost is exacted, especially given the expected environmental loss-related chemical accidents with geographical features. Meanwhile, each event often brings more than one environmental impact. Large number of chemical substances are released in the form of solid, liquid, and gas, leading to serious results. Eight clusters that represent the traits of these accidents are formed, including "leakage," "poisoning," "explosion," "pipeline crack," "river pollution," "dust pollution," "emission," and "industrial effluent." "Explosion" and "gas" possess a strong correlation with "poisoning," located at the center of visualization map.

Maximal design basis accident of fusion neutron source DEMO-TIN

NASA Astrophysics Data System (ADS)

Kolbasov, B. N.

2015-12-01

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

Analysis of 129I in the soils of Fukushima Prefecture: preliminary reconstruction of 131I deposition related to the accident at Fukushima Daiichi Nuclear Power Plant (FDNPP).

PubMed

Muramatsu, Yasuyuki; Matsuzaki, Hiroyuki; Toyama, Chiaki; Ohno, Takeshi

2015-01-01

Iodine-131 is one of the most critical radionuclides to be monitored after release from reactor accidents due to the tendency for this nuclide to accumulate in the human thyroid gland. However, there are not enough data related to the reactor accident in Fukushima, Japan to provide regional information on the deposition of this short-lived nuclide (half-life = 8.02 d). In this study we have focused on the long-lived iodine isotope, (129)I (half-life of 1.57 × 10(7) y), and analyzed it by accelerator mass spectrometry (AMS) for surface soil samples collected at various locations in Fukushima Prefecture. In order to obtain information on the (131)I/(129)I ratio released from the accident, we have determined (129)I concentrations in 82 soil samples in which (131)I concentrations were previously determined. There was a strong correlation (R(2) = 0.84) between the two nuclides, suggesting that the (131)I levels in soil samples following the accident can be estimated through the analysis of (129)I. We have also examined the possible influence from (129m)Te on (129)I, and found no significant effect. In order to construct a deposition map of (131)I, we determined the (129)I concentrations (Bq/kg) in 388 soil samples collected from different locations in Fukushima Prefecture and the deposition densities (Bq/m(2)) of (131)I were reconstructed from the results. Copyright © 2014 Elsevier Ltd. All rights reserved.

[Pedestrian accidents].

PubMed

Bugmann, I; Röthlisberger, M; Friederich, N F; Müller, W

1998-12-03

The epidemiology of non-traffic-related, pedestrian injuries in the winter sports resort of Arosa (Switzerland) between 1968 and 1995 (beginning december to end april) was studied. A total of 2813 patients (1140 male and 1673 female patients) with 3010 injuries was surveyed in a general practice. The study showed in the examined period not only an increasing number of injuries and a culmination in February, but also a higher risk for females (60 versus 40 per cent) and an increasing incidence with age. 6% of all patients had more than one injury and 5% required hospitalization. Most pedestrians had injuries of the upper and lower extremity; fracture of the distal radius was the most often diagnosed injury, followed by distortions of the foot and wrist and fractures of the ankle. For accident prevention it is important, that the authorities do not abstain from mechanical cleaning of walking surfaces as soon as they become slippery and from giving out warnings via the mass media. For outdoor walking we postulate boots for ankle protection. The results are discussed and compared with already existing publications.

Performance modeling of Deep Burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong

2010-01-01

The effects of design choices for the TRISO particle fuel were explored in order to determine their contribution to attaining high-burnup in Deep Burn modular helium reactor fuels containing transuranics from light water reactor spent fuel. The new design features were: (1) ZrC coating substituted for the SiC, allowing the fuel to survive higher accident temperatures; (2) pyrocarbon/SiC "alloy" substituted for the inner pyrocarbon coating to reduce layer failure and (3) pyrocarbon seal coat and thin ZrC oxygen getter coating on the kernel to eliminate CO. Fuel performance was evaluated using General Atomics Company's PISA code. The only acceptable design has a 200-μm kernel diameter coupled with at least 150-μm thick, 50% porosity buffer, a 15-μm ZrC getter over a 10-μm pyrocarbon seal coat on the kernel, an alloy inner pyrocarbon, and ZrC substituted for SiC. The code predicted that during a 1600 °C postulated accident at 70% FIMA, the ZrC failure probability is <10-4.

Monitoring Cs-134 and 137 released by Fukushima Dai-ichi Nuclear Power Plant accident in ground, soil, and stream waters

NASA Astrophysics Data System (ADS)

Tsujimura, Maki; Onda, Yuichi; Hada, Manami; Ishwar, Pun; Abe, Yutaka

2013-04-01

Due to Fukushima Dai-ichi Nuclear power plant accident occurred in March 2011, large amount of radionuclides was released into the atmosphere and was fallen onto ground by rainfall. Few researches have monitored radioactive cesium dynamics in whole hydrological cycle system such as groundwater, soil water, spring water and stream water. Thus, the purpose of this study is to monitor concentration of radioactive cesium in those waters in time series in the headwaters. We have performed an intensive monitoring at three small mountainous catchments in Yamakiya district, Kawamata town, Fukushima prefecture, locating 35 km northwest from Fukushima Dai-ichi Nuclear Power Plant since June 2011, also we consider the movement of radioactive cesium and its relation with the hydrological cycle.

Toxic chemical considerations for tank farm releases

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Keuren, J.C.; Davis, J.S., Westinghouse Hanford

1996-08-01

This topical report contains technical information used to determine the accident consequences of releases of toxic chemical and gases for the Tank Farm Final Safety Analysis report (FSAR).It does not provide results for specific accident scenarios but does provide information for use in those calculations including chemicals to be considered, chemical concentrations, chemical limits and a method of summing the fractional contributions of each chemical. Tank farm composites evaluated were liquids and solids for double shell tanks, single shell tanks, all solids,all liquids, headspace gases, and 241-C-106 solids. Emergency response planning guidelines (ERPGs) were used as the limits.Where ERPGs weremore » not available for the chemicals of interest, surrogate ERPGs were developed. Revision 2 includes updated sample data, an executive summary, and some editorial revisions.« less

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

NASA Astrophysics Data System (ADS)

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

2013-04-01

The Chernobyl nuclear accident and more recently the Fukushima accident highlighted that the largest source of error on consequences assessment is the source term estimation including the time evolution of the release rate and its distribution between radioisotopes. Inverse modelling methods have proved to be efficient to assess the source term due to accidental situation (Gudiksen, 1989, Krysta and Bocquet, 2007, Stohl et al 2011, Winiarek et al 2012). These methods combine environmental measurements and atmospheric dispersion models. They have been recently applied to the Fukushima accident. Most existing approaches are designed to use air sampling measurements (Winiarek et al, 2012) and some of them use also deposition measurements (Stohl et al, 2012, Winiarek et al, 2013). During the Fukushima accident, such measurements are far less numerous and not as well distributed within Japan than the dose rate measurements. To efficiently document the evolution of the contamination, gamma dose rate measurements were numerous, well distributed within Japan and they offered a high temporal frequency. However, dose rate data are not as easy to use as air sampling measurements and until now they were not used in inverse modelling approach. Indeed, dose rate data results from all the gamma emitters present in the ground and in the atmosphere in the vicinity of the receptor. They do not allow one to determine the isotopic composition or to distinguish the plume contribution from wet deposition. The presented approach proposes a way to use dose rate measurement in inverse modeling approach without the need of a-priori information on emissions. The method proved to be efficient and reliable when applied on the Fukushima accident. The emissions for the 8 main isotopes Xe-133, Cs-134, Cs-136, Cs-137, Ba-137m, I-131, I-132 and Te-132 have been assessed. The Daiichi power plant events (such as ventings, explosions…) known to have caused atmospheric releases are well identified in the retrieved source term, except for unit 3 explosion where no measurement was available. The comparisons between the simulations of atmospheric dispersion and deposition of the retrieved source term show a good agreement with environmental observations. Moreover, an important outcome of this study is that the method proved to be perfectly suited to crisis management and should contribute to improve our response in case of a nuclear accident.

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

NASA Astrophysics Data System (ADS)

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

2015-01-01

Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of cumulative surface deposition of total 131I and 137Cs and air dose rate obtained by airborne surveys. The new source term was also tested using three atmospheric dispersion models (Modèle Lagrangien de Dispersion de Particules d'ordre zéro: MLDP0, Hybrid Single Particle Lagrangian Integrated Trajectory Model: HYSPLIT, and Met Office's Numerical Atmospheric-dispersion Modelling Environment: NAME) for regional and global calculations, and the calculated results showed good agreement with observed air concentration and surface deposition of 137Cs in eastern Japan.

Analysis of Air Activity Concentration Data Collected in the Kanto Plain, Japan, following the 2011 Fukushima Daiichi Nuclear Accident

DTIC Science & Technology

2015-06-09

mi) northeast of Tokyo off the coast of Honshu Island. The tsunami triggered by the earthquake damaged the Fukushima Daiichi Nuclear Power Station...Concentration Data Collected in the Kanto Plain, Japan, following the 2011 Fukushima Daiichi Nuclear Accident 5a. CONTRACT NUMBER 5b. GRANT NUMBER...collected at three locations following the release of radioactive materials from the Fukushima Daiichi Nuclear Power Station. The time series at

First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations

PubMed Central

Tsuruta, Haruo; Oura, Yasuji; Ebihara, Mitsuru; Ohara, Toshimasa; Nakajima, Teruyuki

2014-01-01

No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric 137Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12–23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric 137Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models. PMID:25335435

An overview of the Gold King Mine Release and its Transport and Fate in the Animas and San Juan Rivers

EPA Science Inventory

On August 5, 2015, a large acidic mine pool trapped behind a collapsed mine structure and rock debris in the Gold King Mine (GKM) was accidently breached releasing approximately 11.3 million liters (3 million gallons) of low pH (~3) metal contaminated mine drainage into a small t...

Federal Aviation Administration Plan for Research, Engineering & Development, 1998.

DTIC Science & Technology

1998-02-01

release rate. • Improved fracture toughness of non-combustible geopolymer composite fire barriers to enable use as interior and secondary composites ...Fire Resistant, Non-Toxic Interior Panels for Evaluation of Heat Release Rate ♦ Improved Fracture Toughness of Non-Combustible Geopolymer Composite ... composites , atmospheric hazards, crash worthiness, fire safety, and forensics capabilities to support accident investigations. Aviation Security R,E&D

Chemical Accident Prevention: Site Security

EPA Pesticide Factsheets

This chemical safety alert assists facilities that routinely handle extremely hazardous substances, along with SERCs, LEPCs, and emergency responders, in their efforts to reduce criminally caused releases and vulnerability to terrorist activity.

78 FR 6807 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-31

... pressured and emitted powdered debris, hot gases, and molten calcium carbide. The hot gases and debris blown... investigating industrial accidents that result in the release of extremely hazardous substances. The agency's...

Effect of the Duration Time of a Nuclear Accident on Radiological Health Consequences

PubMed Central

Jeong, Hyojoon; Park, Misun; Jeong, Haesun; Hwang, Wontae; Kim, Eunhan; Han, Moonhee

2014-01-01

This study aimed to quantify the effect of duration time of a nuclear accident on the radiation dose of a densely populated area and the resulting acute health effects. In the case of nuclear accidents, the total emissions of radioactive materials can be classified into several categories. Therefore, the release information is very important for the assessment of risk to the public. We confirmed that when the duration time of the emissions are prolonged to 7 hours, the concentrations of radioactive substances in the ambient air are reduced by 50% compared to that when the duration time of emission is one hour. This means that the risk evaluation using only the first wind direction of an accident is very conservative, so it has to be used as a screening level for the risk assessment. Furthermore, it is judged that the proper control of the emission time of a nuclear accident can minimize the health effects on residents. PMID:24619120

Development of a chitosan based double layer-coated tablet as a platform for colon-specific drug delivery

PubMed Central

Kim, Min Soo; Yeom, Dong Woo; Kim, Sung Rae; Yoon, Ho Yub; Kim, Chang Hyun; Son, Ho Yong; Kim, Jin Han; Lee, Sangkil; Choi, Young Wook

2017-01-01

A double layer-coated colon-specific drug delivery system (DL-CDDS) was developed, which consisted of chitosan (CTN) based polymeric subcoating of the core tablet containing citric acid for microclimate acidification, followed by an enteric coating. The polymeric composition ratio of Eudragit E100 and ethyl cellulose and amount of subcoating were optimized using a two-level factorial design method. Drug-release characteristics in terms of dissolution efficiency and controlled-release duration were evaluated in various dissolution media, such as simulated colonic fluid in the presence or absence of CTNase. Microflora activation and a stepwise mechanism for drug release were postulated. Consequently, the optimized DL-CDDS showed drug release in a controlled manner by inhibiting drug release in the stomach and intestine, but releasing the drug gradually in the colon (approximately 40% at 10 hours and 92% at 24 hours in CTNase-supplemented simulated colonic fluid), indicating its feasibility as a novel platform for CDD. PMID:28053506

Inverse modelling of radionuclide release rates using gamma dose rate observations

NASA Astrophysics Data System (ADS)

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

2014-05-01

Severe accidents in nuclear power plants such as the historical accident in Chernobyl 1986 or the more recent disaster in the Fukushima Dai-ichi nuclear power plant in 2011 have drastic impacts on the population and environment. The hazardous consequences reach out on a national and continental scale. Environmental measurements and methods to model the transport and dispersion of the released radionuclides serve as a platform to assess the regional impact of nuclear accidents - both, for research purposes and, more important, to determine the immediate threat to the population. However, the assessments of the regional radionuclide activity concentrations and the individual exposure to radiation dose underlie several uncertainties. For example, the accurate model representation of wet and dry deposition. One of the most significant uncertainty, however, results from the estimation of the source term. That is, the time dependent quantification of the released spectrum of radionuclides during the course of the nuclear accident. The quantification of the source terms of severe nuclear accidents may either remain uncertain (e.g. Chernobyl, Devell et al., 1995) or rely on rather rough estimates of released key radionuclides given by the operators. Precise measurements are mostly missing due to practical limitations during the accident. Inverse modelling can be used to realise a feasible estimation of the source term (Davoine and Bocquet, 2007). Existing point measurements of radionuclide activity concentrations are therefore combined with atmospheric transport models. The release rates of radionuclides at the accident site are then obtained by improving the agreement between the modelled and observed concentrations (Stohl et al., 2012). The accuracy of the method and hence of the resulting source term depends amongst others on the availability, reliability and the resolution in time and space of the observations. Radionuclide activity concentrations are observed on a relatively sparse grid and the temporal resolution of available data may be low within the order of hours or a day. Gamma dose rates on the other hand are observed routinely on a much denser grid and higher temporal resolution. Gamma dose rate measurements contain no explicit information on the observed spectrum of radionuclides and have to be interpreted carefully. Nevertheless, they provide valuable information for the inverse evaluation of the source term due to their availability (Saunier et al., 2013). We present a new inversion approach combining an atmospheric dispersion model and observations of radionuclide activity concentrations and gamma dose rates to obtain the source term of radionuclides. We use the Lagrangian particle dispersion model FLEXPART (Stohl et al., 1998; Stohl et al., 2005) to model the atmospheric transport of the released radionuclides. The gamma dose rates are calculated from the modelled activity concentrations. The inversion method uses a Bayesian formulation considering uncertainties for the a priori source term and the observations (Eckhardt et al., 2008). The a priori information on the source term is a first guess. The gamma dose rate observations will be used with inverse modelling to improve this first guess and to retrieve a reliable source term. The details of this method will be presented at the conference. This work is funded by the Bundesamt für Strahlenschutz BfS, Forschungsvorhaben 3612S60026. References Davoine, X. and Bocquet, M., Atmos. Chem. Phys., 7, 1549-1564, 2007. Devell, L., et al., OCDE/GD(96)12, 1995. Eckhardt, S., et al., Atmos. Chem. Phys., 8, 3881-3897, 2008. Saunier, O., et al., Atmos. Chem. Phys., 13, 11403-11421, 2013. Stohl, A., et al., Atmos. Environ., 32, 4245-4264, 1998. Stohl, A., et al., Atmos. Chem. Phys., 5, 2461-2474, 2005. Stohl, A., et al., Atmos. Chem. Phys., 12, 2313-2343, 2012.

Los Alamos National Laboratory emergency management plan. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramsey, G.F.

1998-07-15

The Laboratory has developed this Emergency Management Plan (EMP) to assist in emergency planning, preparedness, and response to anticipated and actual emergencies. The Plan establishes guidance for ensuring safe Laboratory operation, protection of the environment, and safeguarding Department of Energy (DOE) property. Detailed information and specific instructions required by emergency response personnel to implement the EMP are contained in the Emergency Management Plan Implementing Procedure (EMPIP) document, which consists of individual EMPIPs. The EMP and EMPIPs may be used to assist in resolving emergencies including but not limited to fires, high-energy accidents, hazardous material releases (radioactive and nonradioactive), security incidents,more » transportation accidents, electrical accidents, and natural disasters.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hull, A.P.

From the major accident at Unit 4 of the Chernobyl nuclear power station, a plume of airborne radioactive fission products was initially carried northwesterly toward Poland, thence toward Scandinavia and into Central Europe. Reports of the levels of radioactivity in a variety of media and of external radiation levels were collected in the Department of Energy's Emergency Operations Center and compiled into a data bank. Portions of these and other data which were obtained directly from published and official reports were utilized to make a preliminary assessment of the extent and magnitude of the external dose to individuals downwind frommore » Chernobyl. Radioactive /sup 131/I was the predominant fission product. The time of arrival of the plume and the maximum concentrations of /sup 131/I in air, vegetation and milk and the maximum reported depositions and external radiation levels have been tabulated country by country. A large amount of the total activity in the release was apparently carried to a significant elevation. The data suggest that in areas where rainfall occurred, deposition levels were from ten to one-hundred times those observed in nearby ''dry'' locations. Sufficient spectral data were obtained to establish average release fractions and to establish a reference spectra of the other nuclides in the release. Preliminary calculations indicated that the collective dose equivalent to the population in Scandinavia and Central Europe during the first year after the Chernobyl accident would be about 8 x 10/sup 6/ person-rem. From the Soviet report, it appears that a first year population dose of about 2 x 10/sup 7/ person-rem (2 x 10/sup 5/ Sv) will be received by the population who were downwind of Chernobyl within the U.S.S.R. during the accident and its subsequent releases over the following week. 32 refs., 14 figs., 20 tabs.« less

Development of an Updated Societal-Risk Goal for Nuclear Power Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vicki Bier; Michael Corradini; Robert Youngblood

2014-07-01

This report briefly summarizes work done in FY 2013 on the subject LDRD. The working hypothesis is that societal disruption should be addressed in a safety goal. This is motivated by the point that the Fukushima disaster resulted in very little public dose, but enormous societal disruption; a goal that addressed societal disruption would fill a perceived gap in the US NRC safety goal structure. This year's work entailed analyzing the consequences of postulated accidents at various reactor sites in the US, specifically with a view to quantifying the number of people relocated and the duration of their relocation, tomore » see whether this makes sense as a measure of societal disruption.« less

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

Summary of the Comprehensive Environmental Response, Compensation, and Liability Act (Superfund)

EPA Pesticide Factsheets

CERCLA provides a Federal Superfund to clean up uncontrolled or abandoned hazardous-waste sites as well as accidents, spills, and other emergency releases of pollutants and contaminants into the environment

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

NASA Astrophysics Data System (ADS)

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

2014-05-01

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

Spatial variability of the dose rate from (137)Cs fallout in settlements in Russia and Belarus more than two decades after the Chernobyl accident.

PubMed

Bernhardsson, C; Rääf, C L; Mattsson, S

2015-11-01

Radionuclides from the 1986 Chernobyl accident were released and dispersed during a limited period of time, but under widely varying weather conditions. As a result, there was a high geographical variation in the deposited radioactive fallout per unit area over Europe, depending on the released composition of fission products and the weather during the 10 days of releases. If the plume from Chernobyl coincided with rain, then the radionuclides were unevenly distributed on the ground. However, large variations in the initial fallout also occurred locally or even on a meter scale. Over the ensuing years the initial deposition may have been altered further by different weathering processes or human activities such as agriculture, gardening, and decontamination measures. Using measurements taken more than two decades after the accident, we report on the inhomogeneous distribution of the ground deposition of the fission product (137)Cs and its influence on the dose rate 1 m above ground, on both large and small scales (10ths of km(2) - 1 m(2)), in the Gomel-Bryansk area close to the border between Belarus and Russia. The dose rate from the deposition was observed to vary by one order of magnitude depending on the size of the area considered, whether human processes were applied to the surface or not, and on location specific properties (e.g. radionuclide migration in soil). Copyright © 2015 Elsevier Ltd. All rights reserved.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1988-12-15

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

Development of Northeast Asia Nuclear Power Plant Accident Simulator.

PubMed

Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

2017-06-15

A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Impacts on the marine environment in the case of a hypothetical accident involving the recovery of the dumped Russian submarine K-27, based on dispersion of 137Cs.

PubMed

Hosseini, A; Amundsen, I; Brown, J; Dowdall, M; Karcher, M; Kauker, F; Schnur, R

2017-02-01

There is increasing concern regarding the issue of dumped nuclear waste in the Arctic Seas and in particular dumped objects with Spent Nuclear Fuel (SNF). Amongst dumped objects in the Arctic, the dumped Russian submarine K-27 has received great attention as it contains two reactors with highly enriched fuel and lies at a depth of about 30 m under water. To address these concerns a health and environmental impact assessment has been undertaken. Marine dispersion of potentially released radionuclides as a consequence of different hypothetical accident scenarios was modelled using the model NAOSIM. The outputs from the dispersion modelling have been used as inputs to food-chain transfer and environmental dosimetry models. The annual effective doses for subsistence fishing communities of the Barents-Kara seas region do not exceed 0.6 mSv for hypothetical accidents located at Stepovogo fjord or the Barents Sea. For high rate consumers of fish in Norway, following a potential accident at the Gremikha Bay, annual effects doses would be at around 0.15 mSv. Accumulated doses (over 90 days) for various organisms and for all release scenarios considered were never in excess of 150 μGy. The levels of 137 Cs derived for marine organism in areas close to Norway were not values that would likely cause concern from a regulatory perspective although for subsistence fishing communities close to the considered accident locations, it is not inconceivable that some restrictions on fishing etc. would need to be introduced. Copyright © 2016 Elsevier Ltd. All rights reserved.

Maximal design basis accident of fusion neutron source DEMO-TIN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru

2015-12-15

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

Coastal flooding as a parameter in multi-criteria analysis for industrial site selection

NASA Astrophysics Data System (ADS)

Christina, C.; Memos, C.; Diakoulaki, D.

2014-12-01

Natural hazards can trigger major industrial accidents, which apart from affecting industrial installations may cause a series of accidents with serious impacts on human health and the environment far beyond the site boundary. Such accidents, also called Na-Tech (natural - technical) accidents, deserve particular attention since they can cause release of hazardous substances possibly resulting in severe environmental pollution, explosions and/or fires. There are different kinds of natural events or, in general terms, of natural causes of industrial accidents, such as landslides, hurricanes, high winds, tsunamis, lightning, cold/hot temperature, floods, heavy rains etc that have caused accidents. The scope of this paper is to examine the coastal flooding as a parameter in causing an industrial accident, such as the nuclear disaster in Fukushima, Japan, and the critical role of this parameter in industrial site selection. Land use planning is a complex procedure that requires multi-criteria decision analysis involving economic, environmental and social parameters. In this context the parameter of a natural hazard occurrence, such as coastal flooding, for industrial site selection should be set by the decision makers. In this paper it is evaluated the influence that has in the outcome of a multi-criteria decision analysis for industrial spatial planning the parameter of an accident risk triggered by coastal flooding. The latter is analyzed in the context of both sea-and-inland induced flooding.

The total release of xenon-133 from the Fukushima Dai-ichi nuclear power plant accident.

PubMed

Stohl, Andreas; Seibert, Petra; Wotawa, Gerhard

2012-10-01

The accident at the Fukushima Dai-ichi nuclear power plant (FD-NPP) on 11 March 2011 released large amounts of radioactivity into the atmosphere. We determine the total emission of the noble gas xenon-133 ((133)Xe) using global atmospheric concentration measurements. For estimating the emissions, we used three different methods: (i) using a purely observation-based multi-box model, (ii) comparisons of dispersion model results driven with GFS meteorological data with the observation data, and (iii) such comparisons with the dispersion model driven by ECMWF data. From these three methods, we have obtained total (133)Xe releases from FD-NPP of (i) 16.7 ± 1.9 EBq, (ii) 14.2 ± 0.8 EBq, and (iii) 19.0 ± 3.4 EBq, respectively. These values are substantially larger than the entire (133)Xe inventory of FD-NPP of about 12.2 EBq derived from calculations of nuclear fuel burn-up. Complete release of the entire (133)Xe inventory of FD-NPP and additional release of (133)Xe due to the decay of iodine-133 ((133)I), which can add another 2 EBq to the (133)Xe FD-NPP inventory, is required to explain the atmospheric observations. Two of our three methods indicate even higher emissions, but this may not be a robust finding given the differences between our estimates. Copyright © 2012 Elsevier Ltd. All rights reserved.

Radioactive particles released to the environment from the Fukushima reactors-Confirmation is still needed.

PubMed

Salbu, Brit; Lind, Ole Christian

2016-10-01

After severe nuclear events, a major fraction of refractory radionuclides such as U and Pu are released to the environment in the form of radioactive particles. After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, Pu isotope ratio signals different from that of global fallout have been reported, indicating that spent fuel particles have been released from the reactors or reactor vessels. Radioactive particles containing (37) Cs and other volatile radionuclides, as well as a series of stable refractory metals (Cs, Fe, Zn, U, etc.), have been identified by several authors claiming that these particles originated from the FDNPP fuel. If so, long-lived radioactive isotopes of the refractory metals should have been identified in these particles. It is therefore most probable that volatile radionuclides released as gases during the accidents have deposited on available surfaces such as fly ash, forming condensation particles during release or transport. If spent fuel particles have been deposited in the FDNPP surroundings, information on particle characteristics influencing ecosystem transport, uptake, and effects is essential for assessing environmental impact and risk. More emphasis should therefore be put on the identification of hot spots in the FDNPP environment followed by the characterization of radioactive particles using nanoanalytical-microanalytical techniques to support environmental monitoring, as recommended in the present study. Integr Environ Assess Manag 2016;12:687-689. © 2016 SETAC. © 2016 SETAC.

Environmental accident and its treatment in a developing country: a case study on China.

PubMed

Hou, Yu

2012-08-01

Along with their rapid progress, developing countries have had to deal with more environmental problems, which have been a cause for concern among policy makers and the public in general. This study cites two accidents that happened in China in 2006 that caused serious environmental problems in nearby communities and discusses the problems these accidents created and the resulting disputes among the concerned people. Pollution-causing accidents not only pose threats to the health of the victims but also give rise to environmental disputes that jeopardise national security and social stability. Conflicts normally ensue following a pollution-causing accident, which are more likely to happen within a development zone or industrial park. Few environmental conflicts in the past decades were resolved through litigation. Nevertheless, there are lapses in the regulatory system, which have to be addressed to ensure that the public's rights and interests are protected. Currently, reports on pollution-causing accidents are difficult to obtain and are often released very late. A majority of industrial firms operate without environmental clearance, thus highlighting the government's inefficiency in environmental management. It is about time that the Chinese government takes seriously the use of the Environmental Impact Assessment.

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Knowledge Advancement.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.; Mattie, Patrick D.; Bixler, Nathan E.

2014-02-01

This paper describes the knowledge advancements from the uncertainty analysis for the State-of- the-Art Reactor Consequence Analyses (SOARCA) unmitigated long-term station blackout accident scenario at the Peach Bottom Atomic Power Station. This work assessed key MELCOR and MELCOR Accident Consequence Code System, Version 2 (MACCS2) modeling uncertainties in an integrated fashion to quantify the relative importance of each uncertain input on potential accident progression, radiological releases, and off-site consequences. This quantitative uncertainty analysis provides measures of the effects on consequences, of each of the selected uncertain parameters both individually and in interaction with other parameters. The results measure the modelmore » response (e.g., variance in the output) to uncertainty in the selected input. Investigation into the important uncertain parameters in turn yields insights into important phenomena for accident progression and off-site consequences. This uncertainty analysis confirmed the known importance of some parameters, such as failure rate of the Safety Relief Valve in accident progression modeling and the dry deposition velocity in off-site consequence modeling. The analysis also revealed some new insights, such as dependent effect of cesium chemical form for different accident progressions. (auth)« less

Data reduction and analysis of graphite fiber release experiments

NASA Technical Reports Server (NTRS)

Lieberman, P.; Chovit, A. R.; Sussholz, B.; Korman, H. F.

1979-01-01

The burn and burn/explode effects on aircraft structures were examined in a series of fifteen outdoor tests conducted to verify the results obtained in previous burn and explode tests of carbon/graphite composite samples conducted in a closed chamber, and to simulate aircraft accident scenarios in which carbon/graphite fibers would be released. The primary effects that were to be investigaged in these tests were the amount and size distribution of the conductive fibers released from the composite structures, and how these various sizes of fibers transported downwind. The structures included plates, barrels, aircraft spoilers and a cockpit. The heat sources included a propane gas burner and 20 ft by 20 ft and 40 ft by 60 ft JP-5 pool fires. The larger pool fire was selected to simulate an aircraft accident incident. The passive instrumentation included sticky paper and sticky bridal veil over an area 6000 ft downwind and 3000 ft crosswind. The active instrumentation included instrumented meteorological towers, movies, infrared imaging cameras, LADAR, high voltage ball gages, light emitting diode gages, microwave gages and flame velocimeter.

Guidelines for exposure assessment in health risk studies following a nuclear reactor accident.

PubMed

Bouville, André; Linet, Martha S; Hatch, Maureen; Mabuchi, Kiyohiko; Simon, Steven L

2014-01-01

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. 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. 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. The proposed approach can serve as a basis to optimize the value of radiation dose reconstruction following a nuclear reactor accident.

Newly Reported Hypertension after Military Combat: Deployment in a Large Population-based Study

DTIC Science & Technology

2009-11-01

myocardial infarction and other cardiovascular risk,1 the underlying role of deployment- induced stress or combat-related violence on hypertension is not...well established. Stress is postulated to increase blood pressure through the release of corticoids and inhibition of prostaglandin synthesis, which... exercise in individuals with and without cardiovascular disease: benefits, rationale, safety, and prescription–an advisory from the Committee on Exercise

Retrospection of Chernobyl nuclear accident for decision analysis concerning remedial actions in Ukraine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Georgievskiy, Vladimir

2007-07-01

It is considered the efficacy of decisions concerning remedial actions when of-site radiological monitoring in the early and (or) in the intermediate phases was absent or was not informative. There are examples of such situations in the former Soviet Union where many people have been exposed: releases of radioactive materials from 'Krasnoyarsk-26' into Enisey River, releases of radioactive materials from 'Chelabinsk-65' (the Kishtim accident), nuclear tests at the Semipalatinsk Test Site, the Chernobyl nuclear accident etc. If monitoring in the early and (or) in the intermediate phases is absent the decisions concerning remedial actions are usually developed on the basemore » of permanent monitoring. However decisions of this kind may be essentially erroneous. For these cases it is proposed to make retrospection of radiological data of the early and intermediate phases of nuclear accident and to project decisions concerning remedial actions on the base of both retrospective data and permanent monitoring data. In this Report the indicated problem is considered by the example of the Chernobyl accident for Ukraine. Their of-site radiological monitoring in the early and intermediate phases was unsatisfactory. In particular, the pasture-cow-milk monitoring had not been made. All official decisions concerning dose estimations had been made on the base of measurements of {sup 137}Cs in body (40 measurements in 135 days and 55 measurements in 229 days after the Chernobyl accident). For the retrospection of radiological data of the Chernobyl accident dynamic model has been developed. This model has structure similar to the structure of Pathway model and Farmland model. Parameters of the developed model have been identified for agricultural conditions of Russia and Ukraine. By means of this model dynamics of 20 radionuclides in pathways and dynamics of doses have been estimated for the early, intermediate and late phases of the Chernobyl accident. The main results are following: - During the first year after the Chernobyl accident 75-93% of Commitment Effective Dose had been formed; - During the first year after the Chernobyl accident 85-90% of damage from radiation exposure had been formed. During the next 50 years (the late phase of accident) only 10-15% of damage from radiation exposure will have been formed; - Remedial actions (agricultural remedial actions as most effective) in Ukraine are intended for reduction of the damage from consumption of production which is contaminated in the late phase of accident. I.e. agricultural remedial actions have been intended for minimization only 10 % of the total damage from radiation exposure; - Medical countermeasures can minimize radiation exposure damage by an order of magnitude greater than agricultural countermeasures. - Thus, retrospection of nuclear accident has essentially changed type of remedial actions and has given a chance to increase effectiveness of spending by an order of magnitude. This example illustrates that in order to optimize remedial actions it is required to use data of retrospection of nuclear accidents in all cases when monitoring in the early and (or) intermediate phases is unsatisfactory. (author)« less

10 CFR 72.100 - Defining potential effects of the ISFSI or MRS on the region.

Code of Federal Regulations, 2010 CFR

2010-01-01

... in the region resulting from the release of radioactive materials under normal and accident... with respect to the effects on the regional environment resulting from construction, operation, and...

Advanced risk assessment of the effects of graphite fibers on electronic and electric equipment

NASA Technical Reports Server (NTRS)

Pocinki, L.; Cornell, M.; Kaplan, L.

1980-01-01

An assessment of the risk associated with accidents involving aircraft with carbon fiber composite structural components is examined. The individual fiber segments cause electrical and electronic equipment to fail under certain operating conditions. A Monte Carlo simulation model was used to computer the risk. Aircraft accidents with fire, release of carbon fiber material, entrainment of carbon fibers in a smoke plume transport of fibers downwind, transfer of some fibers/into the the interior of buildings, failures of electrical and electronic equipment, and economic impact of failures are discussed. Risk profiles were prepared for individual airports and the Nation. The vulnerability of electrical transmission equipment to carbon fiber incursion and aircraft accident total costs is investigated.

Radiation Dose Assessments for the Embryo, Fetus, and Nursing Infant during Operation Tomodachi

DTIC Science & Technology

2013-08-01

from the Fukushima Daiichi nuclear power station units’ radiological releases that followed the earthquake and tsunami on March 11, 2011. The...two months following the accident at the Fukushima Daiichi Nuclear Power Station (FDNPS). As discussed in Cassata et al. (2012), this population was...the DOD-affiliated population, who were subject to exposure to radioactive materials released during the Fukushima Daiichi Nuclear Power Station

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su Weinian; Wang, S.-J.; Chiang, S.-C

2005-06-15

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code ismore » chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

Deep Sludge Gas Release Event Analytical Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sams, Terry L.

2013-08-15

Long Abstract. Full Text. The purpose of the Deep Sludge Gas Release Event Analytical Evaluation (DSGRE-AE) is to evaluate the postulated hypothesis that a hydrogen GRE may occur in Hanford tanks containing waste sludges at levels greater than previously experienced. There is a need to understand gas retention and release hazards in sludge beds which are 200 -300 inches deep. These sludge beds are deeper than historical Hanford sludge waste beds, and are created when waste is retrieved from older single-shell tanks (SST) and transferred to newer double-shell tanks (DST).Retrieval of waste from SSTs reduces the risk to the environmentmore » from leakage or potential leakage of waste into the ground from these tanks. However, the possibility of an energetic event (flammable gas accident) in the retrieval receiver DST is worse than slow leakage. Lines of inquiry, therefore, are (1) can sludge waste be stored safely in deep beds; (2) can gas release events (GRE) be prevented by periodically degassing the sludge (e.g., mixer pump); or (3) does the retrieval strategy need to be altered to limit sludge bed height by retrieving into additional DSTs? The scope of this effort is to provide expert advice on whether or not to move forward with the generation of deep beds of sludge through retrieval of C-Farm tanks. Evaluation of possible mitigation methods (e.g., using mixer pumps to release gas, retrieving into an additional DST) are being evaluated by a second team and are not discussed in this report. While available data and engineering judgment indicate that increased gas retention (retained gas fraction) in DST sludge at depths resulting from the completion of SST 241-C Tank Farm retrievals is not expected and, even if gas releases were to occur, they would be small and local, a positive USQ was declared (Occurrence Report EM-RP--WRPS-TANKFARM-2012-0014, "Potential Exists for a Large Spontaneous Gas Release Event in Deep Settled Waste Sludge"). The purpose of this technical report is to (1) present and discuss current understandings of gas retention and release mechanisms for deep sludge in U.S. Department of Energy (DOE) complex waste storage tanks; and (2) to identify viable methods/criteria for demonstrating safety relative to deep sludge gas release events (DSGRE) in the near term to support the Hanford C-Farm retrieval mission. A secondary purpose is to identify viable methods/criteria for demonstrating safety relative to DSGREs in the longer term to support the mission to retrieve waste from the Hanford Tank Farms and deliver it to the Waste Treatment and Immobilization Plant (WTP). The potential DSGRE issue resulted in the declaration of a positive Unreviewed Safety Question (USQ). C-Farm retrievals are currently proceeding under a Justification for Continued Operation (JCO) that only allows tanks 241-AN-101 and 241-AN-106 sludge levels of 192 inches and 195 inches, respectively. C-Farm retrievals need deeper sludge levels (approximately 310 inches in 241-AN-101 and approximately 250 inches in 241-AN-106). This effort is to provide analytical data and justification to continue retrievals in a safe and efficient manner.« less

Of Disasters and Dragon Kings: A Statistical Analysis of Nuclear Power Incidents and Accidents.

PubMed

Wheatley, Spencer; Sovacool, Benjamin; Sornette, Didier

2017-01-01

We perform a statistical study of risk in nuclear energy systems. This study provides and analyzes a data set that is twice the size of the previous best data set on nuclear incidents and accidents, comparing three measures of severity: the industry standard International Nuclear Event Scale, the Nuclear Accident Magnitude Scale of radiation release, and cost in U.S. dollars. The rate of nuclear accidents with cost above 20 MM 2013 USD, per reactor per year, has decreased from the 1970s until the present time. Along the way, the rate dropped significantly after Chernobyl (April 1986) and is expected to be roughly stable around a level of 0.003, suggesting an average of just over one event per year across the current global fleet. The distribution of costs appears to have changed following the Three Mile Island major accident (March 1979). The median cost became approximately 3.5 times smaller, but an extremely heavy tail emerged, being well described by a Pareto distribution with parameter α = 0.5-0.6. For instance, the cost of the two largest events, Chernobyl and Fukushima (March 2011), is equal to nearly five times the sum of the 173 other events. We also document a significant runaway disaster regime in both radiation release and cost data, which we associate with the "dragon-king" phenomenon. Since the major accident at Fukushima (March 2011) occurred recently, we are unable to quantify an impact of the industry response to this disaster. Excluding such improvements, in terms of costs, our range of models suggests that there is presently a 50% chance that (i) a Fukushima event (or larger) occurs every 60-150 years, and (ii) that a Three Mile Island event (or larger) occurs every 10-20 years. Further-even assuming that it is no longer possible to suffer an event more costly than Chernobyl or Fukushima-the expected annual cost and its standard error bracket the cost of a new plant. This highlights the importance of improvements not only immediately following Fukushima, but also deeper improvements to effectively exclude the possibility of "dragon-king" disasters. Finally, we find that the International Nuclear Event Scale (INES) is inconsistent in terms of both cost and radiation released. To be consistent with cost data, the Chernobyl and Fukushima disasters would need to have between an INES level of 10 and 11, rather than the maximum of 7. © 2016 The Authors Risk Analysis published by Wiley Periodicals, Inc. on behalf of Society for Risk Analysis.

The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guntay, Salih; Dehbi, Abdel; Suckow, Detlef

2002-07-01

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes' models for aerosol retention in the secondary side of a steam generator (SG) have not beenmore » assessed against any experimental data, which means that the uncertainties in the source term following an un-isolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (Aerosol Trapping In a Steam generator), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007. (authors)« less

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.

Simulation of Hydrogen Distribution in Ignalina NPP ALS Compartments During BDBA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babilas, Egidijus; Urbonavicius, Egidijus; Rimkevicius, Sigitas

2006-07-01

Accident Localisation System (ALS) of Ignalina NPP is a 'pressure suppression' type confinement, which protects the population, employees and environment from the radiation hazards. According to the Safety Analysis Report for Ignalina NPP {approx}110 m{sup 3} of hydrogen is released to ALS compartments during the Maximum Design Basis Accident. However in case of beyond design basis accident, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. If the volume concentration of hydrogen in the compartment reaches 4%, there is a possibility for a combustible mixture to appear. To prevent the possible hydrogen accumulation inmore » the ALS of the Ignalina NPP during an accident the H{sub 2} control system is installed. The results of the performed analysis derived the places of the possible H{sub 2} accumulation in the ALS compartments during the transient processes and assessed the mixture combustibility in these places for a beyond design basis accident scenario. Such analysis of H{sub 2} distribution in the ALS of Ignalina NPP in case of BDBA was not performed before. (authors)« less

Improved tank car safety research

DOT National Transportation Integrated Search

2007-09-11

Three recent accidents involving the release of hazardous : material have focused attention on the structural integrity of : railroad tank cars: (1) Minot, ND, on January 18, 2002; (2) : Macdona, TX, on June 28, 2004; and (3) Graniteville, SC, on : J...

Cellular automata-based forecasting of the impact of accidental fire and toxic dispersion in process industries.

PubMed

Sarkar, Chinmoy; Abbasi, S A

2006-09-01

The strategies to prevent accidents from occurring in a process industry, or to minimize the harm if an accident does take place, always revolve around forecasting the likely accidents and their impacts. Based on the likely frequency and severity of the accidents, resources are committed towards preventing the accidents. Nearly all techniques of ranking hazardous units, be it the hazard and operability studies, fault tree analysis, hazard indice, etc.--qualitative as well as quantitative--depend essentially on the assessment of the likely frequency and the likely harm accidents in different units may cause. This fact makes it exceedingly important that the forecasting the accidents and their likely impact is done as accurately as possible. In the present study we introduce a new approach to accident forecasting based on the discrete modeling paradigm of cellular automata. In this treatment an accident is modeled as a self-evolving phenomena, the impact of which is strongly influenced by the size, nature, and position of the environmental components which lie in the vicinity of the accident site. The outward propagation of the mass, energy and momentum from the accident epicenter is modeled as a fast diffusion process occurring in discrete space-time coordinates. The quantum of energy and material that would flow into each discrete space element (cell) due to the accidental release is evaluated and the degree of vulnerability posed to the receptors if present in the cell is measured at the end of each time element. This approach is able to effectively take into account the modifications in the flux of energy and material which occur as a result of the heterogeneous environment prevailing between the accident epicenter and the receptor. Consequently, more realistic accident scenarios are generated than possible with the prevailing techniques. The efficacy of the approach has been illustrated with case studies.

Tornado wind-loading requirements based on risk assessment techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deobald, T.L.; Coles, G.A.; Smith, G.L.

Regulations require that nuclear power plants be protected from tornado winds. If struck by a tornado, a plant must be capable of safely shutting down and removing decay heat. Probabilistic techniques are used to show that risk to the public from the US Department of Energy (DOE) SP-100 reactor is acceptable without tornado hardening parts of the secondary system. Relaxed requirements for design wind loadings will result in significant cost savings. To demonstrate an acceptable level of risk, this document examines tornado-initiated accidents. The two tornado-initiated accidents examined in detail are loss of cooling resulting in core damage and lossmore » of secondary system boundary integrity leading to sodium release. Loss of core cooling is analyzed using fault/event tree models. Loss of secondary system boundary integrity is analyzed by comparing the consequences to acceptance criteria for the release of radioactive material or alkali metal aerosol. 4 refs., 4 figs.« less

Assessment of the leak tightness integrity of the steam generator tubes affected by ODSCC at the tube support plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuvelliez, Ch.; Roussel, G.

1997-02-01

An EPRI report gives a method for predicting a conservative value of the total primary-to-secondary leak rate which may occur during, a postulated steam generator depressurization accident such as a Main Steam Line Break (MSLB) in a steam generator with axial through-wall ODSCC at the TSP intersections. The Belgian utility defined an alternative method deviating somewhat from the EPRI method. When reviewing this proposed method, the Belgian safety authorities performed some calculations to investigate its conservatism. This led them to recommend some modifications to the EPRI method which should reduce its undue conservatism while maintaining the objective of conservatism inmore » the offsite dose calculations.« less

Parameter study on the influence of prepressurization on PWR fuel rod behavior during normal operation and hypothetical LOCAs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brzoska, B.; Depisch, F.; Fuchs, H.P.

To analyze the influence of prepressurization on fuel rod behavior, a parametric study has been performed that considers the effects of as-fabricated fuel rod internal prepressure on the normal operation and postulated loss-of-coolant accident (LOCA) rod behavior of a 1300-MW(electric) Kraftwerk Union (KWU) standard pressurized water reactor nuclear power plant. A variation of the prepressure in the range from 15 to 35 bars has only a slight influence on normal operation behavior. Considering the LOCA behavior, only a small temperature increase results from prepressure reduction, while the core-wide straining behavior is improved significantly. The KWU prepressurization takes both conditions intomore » account.« less

Isotopic composition of high-activity particles released in the Chernobyl accident.

PubMed

Osuch, S; Dabrowska, M; Jaracz, P; Kaczanowski, J; Le Van Khoi; Mirowski, S; Piasecki, E; Szeflińska, G; Szefliński, Z; Tropiło, J

1989-11-01

Gamma spectra were measured and activities of the detected isotopes were analyzed for 206 high-activity particles (hot particles, HPs) found in northeastern Poland after the Chernobyl accident. The isotopic composition of HPs observed in gamma-activity is compared with that of the general fallout and core inventory calculations. Particle formation and a process of depletion in Ru and Cs isotopes are discussed. On the basis of a search performed a year later, some comments on the behavior of HPs in the soil are made.

Flow reversal power limit for the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, L.Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. Themore » experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.« less

An novel identification method of the environmental risk sources for surface water pollution accidents in chemical industrial parks.

PubMed

Peng, Jianfeng; Song, Yonghui; Yuan, Peng; Xiao, Shuhu; Han, Lu

2013-07-01

The chemical industry is a major source of various pollution accidents. Improving the management level of risk sources for pollution accidents has become an urgent demand for most industrialized countries. In pollution accidents, the released chemicals harm the receptors to some extent depending on their sensitivity or susceptibility. Therefore, identifying the potential risk sources from such a large number of chemical enterprises has become pressingly urgent. Based on the simulation of the whole accident process, a novel and expandable identification method for risk sources causing water pollution accidents is presented. The newly developed approach, by analyzing and stimulating the whole process of a pollution accident between sources and receptors, can be applied to identify risk sources, especially on the nationwide scale. Three major types of losses, such as social, economic and ecological losses, were normalized, analyzed and used for overall consequence modeling. A specific case study area, located in a chemical industry park (CIP) along the Yangtze River in Jiangsu Province, China, was selected to test the potential of the identification method. The results showed that there were four risk sources for pollution accidents in this CIP. Aniline leakage in the HS Chemical Plant would lead to the most serious impact on the surrounding water environment. This potential accident would severely damage the ecosystem up to 3.8 km downstream of Yangtze River, and lead to pollution over a distance stretching to 73.7 km downstream. The proposed method is easily extended to the nationwide identification of potential risk sources.

Analysis of helium purification system capability during water ingress accident in RDE

NASA Astrophysics Data System (ADS)

Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

Using Student Development Theory to Inform Our Curriculum and Pedagogy: A Response to the Secretary of Education's Commission on the Future of Higher Education

ERIC Educational Resources Information Center

Yordy, Eric D.

2008-01-01

In September 2006, the Secretary of Education's Commission on the Future of Higher Education released its final report entitled "A Test of Leadership: Charting the Future of U.S. Higher Education" postulating that graduates today are lacking important skills such as reading, writing, problem solving, and critical thinking. In the field of…

Computational Study of a Functionally Graded Ceramic-Metallic Armor

DTIC Science & Technology

2006-12-15

UNCLAS –Dist. A - Approved for public release Computational Study of a Functionally Graded Ceramic-Metallic Armor Douglas W. Templeton1, Tara J...efficiency of a postulated FGM ceramic-metallic armor system composed of aluminum nitride (AlN) and aluminum. The study had two primary...2006 2. REPORT TYPE N/ A 3. DATES COVERED - 4. TITLE AND SUBTITLE Computational Study of a Functionally Graded Ceramic-Metallic Armor 5a

Environmental consequences of postulate plutonium releases from Atomics International's Nuclear Materials Development Facility (NMDF), Santa Susana, California, as a result of severe natural phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jamison, J.D.; Watson, E.C.

1982-02-01

Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases caused by severe natural phenomena at the Atomics International's Nuclear Materials Development Facility (NMDF), in the Santa Susana site, California. The severe natural phenomena considered are earthquakes, tornadoes, and high straight-line winds. Plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely 50-year committed dose equivalents are given for the maximum-exposed individual and the population within a 50-mile radius of the plant. The maximum plutonium deposition values likely to occur offsite are alsomore » given. The most likely calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays. The most likely maximum residual plutonium contamination estimated to be deposited offsite following the earthquake, and the 150-mph and 170-mph tornadoes are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 ..mu..Ci/m/sup 2/. The deposition values following the 110-mph and the 130-mph tornadoes are below the EPA proposed guideline.« less

Environmental consequences of postulated plutonium releases from General Electric Company Vallecitos Nuclear Center, Vallecitos, California, as a result of severe natural phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jamison, J.D.; Watson, E.C.

1980-11-01

Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases caused by severe natural phenomena at the General Electric Company Vallecitos Nuclear Center, Vallecitos, California. The severe natural phenomena considered are earthquakes, tornadoes, and high straight-line winds. Maximum plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely 50-year committed dose equivalents are given for the maximum-exposed individual and the population within a 50-mile radius of the plant. The maximum plutonium deposition values likely to occur offsite are also given. The most likelymore » calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays. The most likely maximum residual plutonium contamination estimated to be deposited offsite following the earthquakes, and the 180-mph and 230-mph tornadoes are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 ..mu..Ci/m/sup 2/. The deposition values following the 135-mph tornado are below the EPA proposed guidelines.« less

Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

1995-04-01

This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developedmore » that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.« less

Interpretation and modelling of fission product Ba and Mo releases from fuel

NASA Astrophysics Data System (ADS)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumptionmore » of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. Themore » assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

40 CFR 68.125 - Exemptions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Exemptions. 68.125 Section 68.125 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.125 Exemptions...

40 CFR 68.126 - Exclusion.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Exclusion. 68.126 Section 68.126 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.126 Exclusion...

40 CFR 68.100 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Purpose. 68.100 Section 68.100 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.100 Purpose...

National Oil and Hazardous Substances Pollution Contingency Plan (NCP) Overview

EPA Pesticide Factsheets

First developed and published in 1968, this has become the federal government's blueprint for responding to spills and releases. It provides a system of accident reporting, spill containment and cleanup, and coordination among the hierarchy of responders.

Environmental risk management for radiological accidents: integrating risk assessment and decision analysis for remediation at different spatial scales.

PubMed

Yatsalo, Boris; Sullivan, Terrence; Didenko, Vladimir; Linkov, Igor

2011-07-01

The consequences of the Tohuku earthquake and subsequent tsunami in March 2011 caused a loss of power at the Fukushima Daiichi nuclear power plant, in Japan, and led to the release of radioactive materials into the environment. Although the full extent of the contamination is not currently known, the highly complex nature of the environmental contamination (radionuclides in water, soil, and agricultural produce) typical of nuclear accidents requires a detailed geospatial analysis of information with the ability to extrapolate across different scales with applications to risk assessment models and decision making support. This article briefly summarizes the approach used to inform risk-based land management and remediation decision making after the Chernobyl, Soviet Ukraine, accident in 1986. Copyright © 2011 SETAC.

NASA's Space Shuttle Columbia: Synopsis of the Report of the Columbia Accident Investigation Board

NASA Technical Reports Server (NTRS)

Smith, Marcia S.

2003-01-01

NASA's space shuttle Columbia broke apart on February 1, 2003 as it returned to Earth from a 16-day science mission. All seven astronauts aboard were killed. NASA created the Columbia Accident Investigation Board (CAIB), chaired by Adm. (Ret.) Harold Gehman, to investigate the accident. The Board released its report (available at [http://www.caib.us]) on August 26, 2003, concluding that the tragedy was caused by technical and organizational failures. The CAIB report included 29 recommendations, 15 of which the Board specified must be completed before the shuttle returns to flight status. This report provides a brief synopsis of the Board's conclusions, recommendations, and observations. Further information on Columbia and issues for Congress are available in CRS Report RS21408. This report will not be updated.

Radioactivity impacts of the Fukushima Nuclear Accident on the atmosphere

NASA Astrophysics Data System (ADS)

Lin, W.; Chen, L.; Yu, W.; Ma, H.; Zeng, Z.; Lin, J.; Zeng, S.

2015-02-01

The Fukushima Nuclear Accident (FNA) resulted in a large amount of radionuclides released into the atmosphere and dispersed globally, which has greatly raised public concerns. The state of the art for source terms of 19 kinds of radionuclides derived from the FNA was comprehensively collected and compared with levels of the global fallout and the Chernobyl Nuclear Accident (CNA). The atmospheric impacts of the FNA were evaluated from three aspects including radioactive baseline of the atmosphere, the concentration limits in standards and radiological protection. The FNA should not impose significant radiological risk on the public members in the countries excluding Japan. A conceptual scheme of Fukushima-derived radionuclides with physical and physicochemical insights on different temporal-spatial timescales was discussed and illustrated to understand their fates in the atmosphere.

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.

Observed Barium Emission Rates

NASA Technical Reports Server (NTRS)

Stenbaek-Nielsen, H. C.; Wescott, E. M.; Hallinan, T. J.

1993-01-01

The barium releases from the CRRES satellite have provided an opportunity for verifying theoretically calculated barium ion and neutral emission rates. Spectra of the five Caribbean releases in the summer of 1991 were taken with a spectrograph on board a U.S. Air Force jet aircraft. Because the line of sight release densities are not known, only relative rates could be obtained. The observed relative rates agree well with the theoretically calculated rates and, together with other observations, confirm the earlier detailed theoretical emission rates. The calculated emission rates can thus with good accuracy be used with photometric observations. It has been postulated that charge exchange between neutral barium and oxygen ions represents a significant source for ionization. If so. it should be associated with emissions at 4957.15 A and 5013.00 A, but these emissions were not detected.

STUDIES ON TUBERCULIN FEVER. 3. MECHANISMS INVOLVED IN THE RELEASE OF ENDOGENOUS PYROGEN IN VITRO.

PubMed

ATKINS, E; HEIJN, C

1965-08-01

In a search for the source of the circulating endogenous pyrogen (EP) that mediates tuberculin-induced fever, tuberculin was incubated in vitro with various tissues of rabbits sensitized by intravenous infection with BCG. Evidence was obtained that tuberculin specifically stimulates cells in the blood of sensitized rabbits to generate pyrogen in vitro, whereas both lymph node and spleen cells from the same donors were inactive. Since normal blood cells, incubated in plasma of sensitized donors, were similarly activated, it is postulated that circulating antibodies play a role in sensitizing cells (presumably granulocytes) to release pyrogen on contact with tuberculin) both in vitro and in vivo. Release of endogenous pyrogen in vitro may be a sensitive means of detecting immunologic reactions between antigen and specifically sensitized blood cells-in other allergic states accompanied by fever.

The effect of the NMDA receptor blocker, dextromethorphan, on cribbing in horses.

PubMed

Rendon, R A; Shuster, L; Dodman, N H

2001-01-01

Stereotypic cribbing in horses is thought to involve excess dopaminergic activity within the striatum. Various models of stress-induced stereotypies including cribbing in horses postulate that stress stimulates the release of endorphins, triggering the release of striatal dopamine. Dopamine in turn activates basal ganglia motor programs, reinforcing behavior via a reward mechanism. Furthermore, the release of dopamine by endorphins has been shown to depend on activation of NMDA receptors. In the present study, horses identified as cribbers and volunteered by their owners were treated with the NMDA receptor antagonist dextromethorphan (DM). When DM was administered via jugular injection (1 mg/kg), eight of nine horses responded with reductions in cribbing rate (CR) compared to baseline, and cribbing was suppressed completely for a period of time in almost half of the horses tested.

Simulation Technology Laboratory Building 970 hazards assessment document

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, C.L.; Starr, M.D.

1994-11-01

The Department of Energy Order 5500.3A requires facility-specific hazards assessments be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Simulation Technology Laboratory, Building 970. The entire inventory was screened according to the potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distances at which a postulated facility event will producemore » consequences exceeding the ERPG-2 and Early Severe Health Effects thresholds are 78 and 46 meters, respectively. The highest emergency classification is a Site Area Emergency. The Emergency Planning Zone is 100 meters.« less

Radioactive contamination of nest materials of the Eurasian Tree Sparrow Passer montanus due to the Fukushima nuclear accident: The significance in the first year.

PubMed

Matsui, Shin; Kasahara, Satoe; Morimoto, Gen; Mikami, Osamu K; Watanabe, Mamoru; Ueda, Keisuke

2015-11-01

The 2011 Fukushima nuclear accident contaminated large areas of eastern and northeastern Japan, releasing vast amounts of radiation. Here we investigated radioactive contamination of the nest materials of the Eurasian Tree Sparrow Passer montanus from the breeding season of 2011 directly after the accident to the next breeding season of 2012 at two sites. In Tokyo (222 km southwest of the plant), ambient dose rates in the nestboxes were lower than those in Ibaraki (175 km southwest of the plant), where the levels of 2011 were higher than those of 2012. Further, the amount of radioactive Cs in each nest increased with the increase in nest weight, with a higher increment at Ibaraki than at Tokyo. These data suggested higher nest contamination levels in the breeding season directly after a nuclear accident than in later seasons, and an increment of nest contamination levels via nest materials of birds. Copyright © 2015 Elsevier Ltd. All rights reserved.

The Viareggio LPG railway accident: event reconstruction and modeling.

PubMed

Brambilla, Sara; Manca, Davide

2010-10-15

This manuscript describes in detail the LPG accident occurred in Viareggio on June 2009 and its modeling. The accident investigation highlighted the uncertainty and complexity of assessing and modeling what happened in the congested environment close to the Viareggio railway station. Nonetheless, the analysis allowed comprehending the sequence of events, the way they influenced each other, and the different possible paths/evolutions. The paper describes suitable models for the quantitative assessment of the consequences of the most probable accidental dynamics and its outcomes. The main finding is that after about 80 s from the beginning of the release the dense-gas cloud reached the surrounding houses that were destroyed successively by internal explosions. This fact has two main implications. First, it shows that the adopted modeling framework can give a correct picture of what happened in Viareggio. Second, it confirms the need to develop effective mitigation measures because, in case of this kind of accidents, there is no time to apply any protective emergency plans/actions. 2010 Elsevier B.V. All rights reserved.

Determination of 137Cs activity in soil from Qatar using high-resolution gamma-ray spectrometry

NASA Astrophysics Data System (ADS)

Al-Sulaiti, Huda; Nasir, Tabassum; Al Mugren, K. S.; Alkhomashi, N.; Al-Dahan, N.; Al-Dosari, M.; Bradley, D. A.; Bukhari, S.; Matthews, M.; Regan, P. H.; Santawamaitre, T.; Malain, D.; Habib, A.; Al-Dosari, Hanan; Al Sadig, Ibrahim; Daar, Eman

2016-10-01

With interest in establishing baseline concentrations of 137Cs in soil from the Qatarian peninsula, we focus on determination of the activity concentrations in 129 soil samples collected across the State of Qatar prior to the 2011 Fukushima Dai-ichi nuclear power plant accident. As such, the data provides the basis of a reference map for the detection of releases of this fission product. The activity concentrations were measured via high-resolution gamma-ray spectrometry using a hyper-pure germanium detector enclosed in a copper-lined passive lead shield that was situated in a low-background environment. The activity concentrations ranged from 0.21 to 15.41 Bq/kg, with a median value of 1 Bq/kg, the greatest activity concentration being observed in a sample obtained from northern Qatar. Although it cannot be confirmed, it is expected that this contamination is mainly due to releases from the Chernobyl accident of 26 April 1986, there being a lack of data from Qatar before the accident. The values are typically within but are sometimes lower than the range indicated by data from other countries in the region. The lower values than those of others is suggested to be due to variation in soil characteristics as well as metrological factors at the time of deposition.

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Estimation of the time-dependent radioactive source-term from the Fukushima nuclear power plant accident using atmospheric transport modelling

NASA Astrophysics Data System (ADS)

Schoeppner, M.; Plastino, W.; Budano, A.; De Vincenzi, M.; Ruggieri, F.

2012-04-01

Several nuclear reactors at the Fukushima Dai-ichi power plant have been severely damaged from the Tōhoku earthquake and the subsequent tsunami in March 2011. Due to the extremely difficult on-site situation it has been not been possible to directly determine the emissions of radioactive material. However, during the following days and weeks radionuclides of 137-Caesium and 131-Iodine (amongst others) were detected at monitoring stations throughout the world. Atmospheric transport models are able to simulate the worldwide dispersion of particles accordant to location, time and meteorological conditions following the release. The Lagrangian atmospheric transport model Flexpart is used by many authorities and has been proven to make valid predictions in this regard. The Flexpart software has first has been ported to a local cluster computer at the Grid Lab of INFN and Department of Physics of University of Roma Tre (Rome, Italy) and subsequently also to the European Mediterranean Grid (EUMEDGRID). Due to this computing power being available it has been possible to simulate the transport of particles originating from the Fukushima Dai-ichi plant site. Using the time series of the sampled concentration data and the assumption that the Fukushima accident was the only source of these radionuclides, it has been possible to estimate the time-dependent source-term for fourteen days following the accident using the atmospheric transport model. A reasonable agreement has been obtained between the modelling results and the estimated radionuclide release rates from the Fukushima accident.

General RMP Guidance - Chapter 3: Five-Year Accident History

EPA Pesticide Factsheets

This involves the reporting of significant accidental releases of one or more of the regulated toxic or flammable substances from a covered process in the five years prior to the submission of an initial or updated Risk Management Plan.

INTERNAL DOSE AND RESPONSE IN REAL-TIME.

EPA Science Inventory

Abstract: Rapid temporal fluctuations in exposure may occur in a number of situations such as accidents or other unexpected acute releases of airborne substances. Often risk assessments overlook temporal exposure patterns under simplifying assumptions such as the use of time-wei...

Mechanical properties of tank car steels retired from the fleet

DOT National Transportation Integrated Search

2009-03-03

As a consequence of recent accidents involving the release of hazardous materials (hazmat), the structural integrity and crashworthiness of railroad tank cars have come under scrutiny. Particular attention has been given to the older portion of the f...

Chemical Safety Alert: Chemical Accidents from Electric Power Outages

EPA Pesticide Factsheets

Incident data from the National Response Center (NRC) shows that during 2000 there were about 240 chemical releases reported due to an electric power interruption, as well as resumption/restart; only a few were related to planned rolling blackouts.

10 CFR 30.50 - Reporting requirements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... prevents immediate protective actions necessary to avoid exposures to radiation or radioactive materials... license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

10 CFR 70.50 - Reporting requirements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... event that prevents immediate protective actions necessary to avoid exposures to radiation or... licensee condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

10 CFR 70.50 - Reporting requirements.

Code of Federal Regulations, 2013 CFR

2013-01-01

... event that prevents immediate protective actions necessary to avoid exposures to radiation or... licensee condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

10 CFR 30.50 - Reporting requirements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... prevents immediate protective actions necessary to avoid exposures to radiation or radioactive materials... license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

10 CFR 70.50 - Reporting requirements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... event that prevents immediate protective actions necessary to avoid exposures to radiation or... licensee condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

10 CFR 30.50 - Reporting requirements.

Code of Federal Regulations, 2013 CFR

2013-01-01

... prevents immediate protective actions necessary to avoid exposures to radiation or radioactive materials... license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The...

Fate of stable strontium in the sewage treatment process as an analog for radiostrontium released by nuclear accidents.

PubMed

Kamei-Ishikawa, Nao; Ito, Ayumi; Umita, Teruyuki

2013-09-15

Radionuclides were widely released into the environment due to the nuclear accident at the Fukushima Daiichi Nuclear Power Plant. Some of these radionuclides have flowed into municipal sewage treatment plants through sewer systems. We have observed the fate of stable Sr in the sewage treatment process as a means to predict the fate of radiostrontium. Concentrations of stable Sr were determined in sewage influent, effluent, dewatered sludge, and incinerated sewage sludge ash collected from a sewage treatment plant once a month from July to December 2011. In the mass balance of Sr in the sewage treatment plant, 76% of the Sr entering the plant was discharged to the receiving water on average. Additionally, 14% of the Sr flowing through the plant was transferred to the sewage sludge and then concentrated in the sludge ash without being released to the atmosphere. We also investigated Sr sorption by activated sludge in a batch experiment. Measurements at 3 and 6h after the contact showed Sr was sorbed in the activated sludge; however, the measurements indicated Sr desorption from activated sludge occurred 48 h after the contact. Copyright © 2013 Elsevier B.V. All rights reserved.

Analytics of Radioactive Materials Released in the Fukushima Daiichi Nuclear Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Egarievwe, Stephen U.; Nuclear Engineering Department, University of Tennessee, Knoxville, TN; Coble, Jamie B.

The 2011 Fukushima Daiichi nuclear accident in Japan resulted in the release of radioactive materials into the atmosphere, the nearby sea, and the surrounding land. Following the accident, several meteorological models were used to predict the transport of the radioactive materials to other continents such as North America and Europe. Also of high importance is the dispersion of radioactive materials locally and within Japan. Based on the International Atomic Energy Agency (IAEA) Convention on Early Notification of a nuclear accident, several radiological data sets were collected on the accident by the Japanese authorities. Among the radioactive materials monitored, are I-131more » and Cs-137 which form the major contributions to the contamination of drinking water. The radiation dose in the atmosphere was also measured. It is impractical to measure contamination and radiation dose in every place of interest. Therefore, modeling helps to predict contamination and radiation dose. Some modeling studies that have been reported in the literature include the simulation of transport and deposition of I-131 and Cs-137 from the accident, Cs-137 deposition and contamination of Japanese soils, and preliminary estimates of I-131 and Cs-137 discharged from the plant into the atmosphere. In this paper, we present statistical analytics of I-131 and Cs-137 with the goal of predicting gamma dose from the Fukushima Daiichi nuclear accident. The data sets used in our study were collected from the IAEA Fukushima Monitoring Database. As part of this study, we investigated several regression models to find the best algorithm for modeling the gamma dose. The modeling techniques used in our study include linear regression, principal component regression (PCR), partial least square (PLS) regression, and ridge regression. Our preliminary results on the first set of data showed that the linear regression model with one variable was the best with a root mean square error of 0.0133 μSv/h, compared to 0.0210 μSv/h for PCR, 0.231 μSv/h for ridge regression L-curve, 0.0856 μSv/h for PLS, and 0.0860 μSv/h for ridge regression cross validation. Complete results using the full datasets for these models will also be presented. (authors)« less

Safety analysts training

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolton, P.

The purpose of this task was to support ESH-3 in providing Airborne Release Fraction and Respirable Fraction training to safety analysts at LANL who perform accident analysis, hazard analysis, safety analysis, and/or risk assessments at nuclear facilities. The task included preparation of materials for and the conduct of two 3-day training courses covering the following topics: safety analysis process; calculation model; aerosol physic concepts for safety analysis; and overview of empirically derived airborne release fractions and respirable fractions.

ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.

2002-07-01

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enablemore » much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)« less

A review of the Fukushima nuclear reactor accident: radiation effects on the thyroid and strategies for prevention.

PubMed

Nagataki, Shigenobu; Takamura, Noboru

2014-10-01

This is a summary of the nuclear accident at the Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Stations (FDNPS) on 11 March 2011 to be used as a review of the radiation effects to the thyroid and strategies of prevention. The amount of radioiodine released to the environment following the Fukushima accident was 120 Peta Becquerel, which is approximately one-tenth of that in the Chernobyl accident. Residents near the FDNPS were evacuated within a few days and foodstuffs were controlled within 1 or 2 weeks. Therefore, thyroid radiation doses were less than 100 mSv (intervention levels for stable iodine administration) in the majority of children, including less than 1 year olds, living in the evacuation areas. Because the incidence of childhood thyroid cancer increased in those residing near the site following the Chernobyl accident, thyroid screening of all children (0-18 years old) in the Fukushima Prefecture was started. To date, screening of more than 280 000 children has resulted in the diagnosis of thyroid cancer in 90 children (approximate incidence, 313 per million). Thus, although the dose of radiation was much lower, the incidence of thyroid cancer appears to be much higher than that following the Chernobyl accident. A comparison of the thyroidal consequences following the Fukushima and Chernobyl nuclear reactor accidents is discussed. We also summarize the recent increased incidence in thyroid cancer in the Fukushima area following the accident in relation to increased thyroid ultrasound screening and the use of advanced ultrasound techniques. http://links.lww.com/COE/A8.

A methodology to quantify the release of spent nuclear fuel from dry casks during security-related scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Luna, Robert Earl

Assessing the risk to the public and the environment from a release of radioactive material produced by accidental or purposeful forces/environments is an important aspect of the regulatory process in many facets of the nuclear industry. In particular, the transport and storage of radioactive materials is of particular concern to the public, especially with regard to potential sabotage acts that might be undertaken by terror groups to cause injuries, panic, and/or economic consequences to a nation. For many such postulated attacks, no breach in the robust cask or storage module containment is expected to occur. However, there exists evidence thatmore » some hypothetical attack modes can penetrate and cause a release of radioactive material. This report is intended as an unclassified overview of the methodology for release estimation as well as a guide to useful resource data from unclassified sources and relevant analysis methods for the estimation process.« less

Emergency preparedness for the accidental release of radionuclides from the Uljin Nuclear Power Plant in Korea.

PubMed

Park, Soon-Ung; Lee, In-Hye; Joo, Seung Jin; Ju, Jae-Won

2017-12-01

Site specific radionuclide dispersion databases were archived for the emergency response to the hypothetical releases of 137 Cs from the Uljin nuclear power plant in Korea. These databases were obtained with the horizontal resolution of 1.5 km in the local domain centered the power plant site by simulations of the Lagrangian Particle Dispersion Model (LPDM) with the Unified Model (UM)-Local Data Assimilation Prediction System (LDAPS). The Eulerian Dispersion Model-East Asia (EDM-EA) with the UM-Global Data Assimilation Prediction System (UM-GDAPS) meteorological models was used to get dispersion databases in the regional domain. The LPDM model was performed for a year with a 5-day interval yielding 72 synoptic time-scale cases in a year. For each case hourly mean near surface concentrations, hourly mean column integrated concentrations, hourly total depositions for 5 consecutive days were archived by the LPDM model in the local domain and by the EDM-EA model in the regional domain of Asia. Among 72 synoptic cases in a year the worst synoptic case that showed the highest mean surface concentration averaged for 5 days in the LPDM model domain was chosen to illustrate the emergency preparedness to the hypothetical accident at the site. The simulated results by the LPDM model with the 137 Cs emission rate of the Fukushima nuclear power plant accident for the first 5-day period were found to be able to provide prerequisite information for the emergency response to the early phase of the accident whereas those of the EDM-EA model could provide information required for the environmental impact assessment of the accident in the regional domain. The archived site-specific database of 72 synoptic cases in a year could have a great potential to be used as a prognostic information on the emergency preparedness for the early phase of accident. Copyright © 2017 Elsevier Ltd. All rights reserved.

Cancer near the Three Mile Island nuclear plant: radiation emissions.

PubMed

Hatch, M C; Beyea, J; Nieves, J W; Susser, M

1990-09-01

As a public charge, cancers among the 159,684 residents living within a 10-mile (16-km) radius of the Three Mile Island nuclear plant were studied relative to releases of radiation during the March 28, 1979, accident as well as to routine plant emissions. The principal cancers considered were leukemia and childhood malignancies. Estimates of the emissions delivered to small geographic study tracts were derived from mathematical dispersion models which accounted for modifying factors such as wind and terrain; the model of accident emissions was validated by readings from off-site dosimeters. Incident cancers among area residents for the period 1975-1985 (n = 5,493) were identified by a review of the records at all local and regional hospitals; preaccident and postaccident trends in cancer rates were examined. For accident emissions, the authors failed to find definite effects of exposure on the cancer types and population subgroups thought to be most susceptible to radiation. No associations were seen for leukemia in adults or for childhood cancers as a group. For leukemia in children, the odds ratio was raised, but cases were few (n = 4), and the estimate was highly variable. Moreover, rates of childhood leukemia in the Three Mile Island area are low compared with national and regional rates. For exposure to routine emissions, the odds ratios were raised for childhood cancers as a whole and for childhood leukemia, but confidence intervals were wide and included 1.0. For leukemia in adults, there was a negative trend. Trends for two types of cancer ran counter to expectation. Non-Hodgkin's lymphoma showed raised risks relative to both accident and routine emissions; lung cancer (adjusted only indirectly for smoking) showed raised risks relative to accident emissions, routine emissions, and background gamma radiation. Overall, the pattern of results does not provide convincing evidence that radiation releases from the Three Mile Island nuclear facility influenced cancer risk during the limited period of follow-up.

Chemical Accident Prevention - Chemicals in your Community

EPA Pesticide Factsheets

This pamphlet summarizes the information you can obtain under Emergency Planning and Community Right-to-Know and Clean Air Act, how you can use these sources to build a snapshot of chemicals stored and released, and how SERCs and LEPCs can improve safety.

10 CFR 40.60 - Reporting requirements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or... releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The equipment is required to be...

10 CFR 40.60 - Reporting requirements.

Code of Federal Regulations, 2013 CFR

2013-01-01

... to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or... releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The equipment is required to be...

10 CFR 40.60 - Reporting requirements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or... releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident; (ii) The equipment is required to be...

An Analysis of Accidents Involving Towboat-Barge Combinations on Selected Inland Waterways of the United States.

DTIC Science & Technology

1980-06-01

AD-A091 054 NAVAL POSTGRADUATE SCHOOL MONTEREY CA F/G 13/12 AN ANALYSIS OF ACCIDENTS INVOLVING TOWBOAT-BARGE COMBINATIONS O--ETC(U) JUN 80 W J GAMBLE...STATES*, William John/Gamble JuneV-980 -a- Thesis Advisor: D . E. Neil C) L Approved for public release; distribution unlimited LA. 80 10 21 008...I.mfeno Aa Monterey, California 93940 8 MS MONITOR1ING AGEICV 86NM4 A *DORESVIO ifenne fts Ca"IE O00146. j ISCUNITY CLASS. (0# this Me")w Naval

NASA Standard for Models and Simulations (M and S): Development Process and Rationale

NASA Technical Reports Server (NTRS)

Zang, Thomas A.; Blattnig, Steve R.; Green, Lawrence L.; Hemsch, Michael J.; Luckring, James M.; Morison, Joseph H.; Tripathi, Ram K.

2009-01-01

After the Columbia Accident Investigation Board (CAIB) report. the NASA Administrator at that time chartered an executive team (known as the Diaz Team) to identify the CAIB report elements with Agency-wide applicability, and to develop corrective measures to address each element. This report documents the chronological development and release of an Agency-wide Standard for Models and Simulations (M&S) (NASA Standard 7009) in response to Action #4 from the report, "A Renewed Commitment to Excellence: An Assessment of the NASA Agency-wide Applicability of the Columbia Accident Investigation Board Report, January 30, 2004".

Severe accident skyshine radiation analysis by MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eurajoki, T.

1994-12-31

If a severe accident with a considerable core damage occurs at a nuclear power plant whose containment top is remarkably thin compared with the walls, the radiation transported through the top and scattered in air may cause high dose rates at the power plant area. Noble gases and other fission products released to the containment act as sources. The dose rates caused by skyshine have been calculated by MCNP3A for the Loviisa nuclear power plant (two-unit, 445-MW VVER) for the outside area and inside some buildings, taking the attenuation in the roofs of the buildings into account.

On-line fission products measurements during a PWR severe accident: the French DECA-PF project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ducros, G.; Allinei, P.G.; Roure, C.

Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, fundedmore » in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied: - Gamma spectrometry, with an industrial objective to build a prototype aimed at improving the capacity of the present radiation monitoring system, - Gas chromatography, for the quantification of the fission gases (Xe, Kr) as well as potential carbon oxides produced in case of Molten Corium Concrete Interaction, - Optical absorption spectroscopy, the objective of this most innovative technique being to quantify the tetra-oxide of ruthenium, which could be produced in case of lower head failure, and the gaseous forms of iodine (molecular and organic) released in the environment. A global description and the present status of this project is presented, focusing on the Source Term establishment at the outlet stack of the sand bed filters and on the perspectives of implementation of the on-line gamma spectrometry equipment. (authors)« less

Use of questionnaires and an expert panel to judge the environmental consequences of chemical spills for the development of an environment-accident index.

PubMed

Andersson, Asa Scott; Stjernström, Olof; Fängmark, Ingrid

2005-05-01

Assessing the environmental consequences of a chemical accident is a complex task. To date, the methods used to evaluate the environmental effects of an acute release of a chemical have often been based on measurements of chemical and physical variables deemed to be important, such as the concentration of the chemical. However, a broader strategy is needed to predict the environmental consequences of potential accidents during the planning process. An Environment-Accident Index (EAI), a simple tool based on such a strategy, has been developed to facilitate the consideration of a multitude of influential variables. The objectives of this study were to evaluate whether questionnaire-based expert panel's judgements could provide useful data on the environmental consequences of chemical spills, and an effective basis for further development of the EAI. As expected, the judgements did not agree perfectly, but they do give rough indications of the environmental effects, and highlight consistent trends that should be useful inputs for planning, prevention and decontamination processes. The different accidents were also judged to have caused everything from minor to very major effects in the environment, implying that a wide range of accident scenarios were represented in the material and covered by the EAI. Therefore, questionnaires and expert panel judgements can be used to collect useful data for estimating the likely environmental consequences of chemical accidents and for further development of the EAI.

Surveillance of Strontium-90 in Foods after the Fukushima Daiichi Nuclear Power Plant Accident.

PubMed

Nabeshi, Hiromi; Tsutsumi, Tomoaki; Uekusa, Yoshinori; Hachisuka, Akiko; Matsuda, Rieko; Teshima, Reiko

2015-01-01

As a result of the Fukushima Daiichi nuclear power plant (NPP) accident, various radionuclides were released into the environment. In this study, we surveyed strontium-90 ((90)Sr) concentrations in several foodstuffs. Strontium-90 is thought to be the third most important residual radionuclide in food collected after the Fukushima Daiichi, NPP accident after following cesium-137 ((137)Cs) and cesium-134 ((134)Cs). Results of (90)Sr analyses indicated that (90)Sr was detect in 25 of the 40 radioactive cesium (r-Cs) positive samples collected in areas around the Fukushima Daiichi NPP, ranging in distance from 50 to 250 km. R-Cs positive samples were defined as containing both (134)Cs and (137)Cs which are considered to be indicators of the after-effects of the Fukushima Daiichi NPP accident. We also detected (90)Sr in 8 of 13 r-Cs negative samples, in which (134)Cs was not detected. Strontium-90 concentrations in the r-Cs positive samples did not significantly exceed the (90)Sr concentrations in r-Cs negative samples or the (90)Sr concentration ranges in comparable food groups found in previous surveys before the Fukushima Daiichi NPP accident. Thus, (90)Sr concentrations in r-Cs positive samples were indistinguishable from the background (90)Sr concentrations arising from global fallout prior to the Fukushima accident, suggesting that no marked increase of (90)Sr concentrations has occurred in r-Cs positive samples as a result of the Fukushima Daiichi NPP accident.

Equivalent Safety Basis for Evaluation of On-Site Packages for US DOE Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, A.C.

Packages for transport of radioactive material within the boundaries of a Department of Energy facility (on-site) must conform to the requirements for packages shipped in normal commerce, or must provide equivalent safety. Equivalence is achieved if the frequency of severe on-site accidents, which could result in a release of radioactive material, is less than or equal to the frequency of Beyond-HAC accidents for packages in commerce. This is shown to be achieved it the rate of on-site accident is 22 per 100 MVM or lower. For equivalence to Normal Conditions of Transport, for on-site packages, appropriate, defensible Design Basis Conditionsmore » can be established and the ability of the package to meet the reduced requirements shown in the On-site Safety Assessment.« less

Three Mile Island epidemiologic radiation dose assessment revisited: 25 years after the accident.

PubMed

Field, R William

2005-01-01

Over the past 25 years, public health concerns following the Three Mile Island (TMI) accident prompted several epidemiologic investigations in the vicinity of TMI. One of these studies is ongoing. This commentary suggests that the major source of radiation exposure to the population has been ignored as a potential confounding factor or effect modifying factor in previous and ongoing TMI epidemiologic studies that explore whether or not TMI accidental plant radiation releases caused an increase in lung cancer in the community around TMI. The commentary also documents the observation that the counties around TMI have the highest regional radon potential in the United States and concludes that radon progeny exposure should be included as part of the overall radiation dose assessment in future studies of radiation-induced lung cancer resulting from the TMI accident.

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, P.; Umminger, K.J.; Schoen, B.

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where themore » decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).« less

Calculational note for the radiological and toxicological effects of a UO{sub 3} release from the T-hopper storage pad

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldberg, H.J.

1998-06-18

UO{sub 3} powder is stored at the T-hopper storage area associated with the 2714-U building in the 200 west area. The T-hopper containers and 13 drums containing this material are used to store the powder on pads immediately north of the building. An interim safety basis document (WHC,1996) was issued in 1996 for the UO{sub 3} powder storage area. In this document the isotope {sup 99}Tc was not included in the source term used to calculate the radiological consequences of a postulated release of the powder. A calculations note (HNF, 1998) was issued to remedy that deficiency. The present documentmore » is a revision to that document to reflect updated data concerning the solubility of UO{sub 3} in simulated lung fluid and to utilize more realistic powder release fractions.« less

Development of the table of initial isolation distances and protective action distances for the 2004 emergency response guidebook.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, D. F.; Freeman, W. A.; Carhart, R. A.

2005-09-23

This report provides technical documentation for values in the Table of Initial Isolation and Protective Action Distances (PADs) in the 2004 Emergency Response Guidebook (ERG2004). The objective for choosing the PADs specified in the ERG2004 is to balance the need to adequately protect the public from exposure to potentially harmful substances against the risks and expenses that could result from overreacting to a spill. To quantify this balance, a statistical approach is adopted, whereby the best available information is used to conduct an accident scenario analysis and develop a set of up to 1,000,000 hypothetical incidents. The set accounts formore » differences in containers types, incident types, accident severity (i.e., amounts released), locations, times of day, times of year, and meteorological conditions. Each scenario is analyzed using detailed emission rate and atmospheric dispersion models to calculate the downwind chemical concentrations from which a 'safe distance' is determined. The safe distance is defined as the distance downwind from the source at which the chemical concentration falls below health protection criteria. The American Industrial Hygiene Association's Emergency Response Planning Guideline Level 2 (ERPG-2) or equivalent is the health criteria used. The statistical sample of safe distance values for all incidents considered in the analysis are separated into four categories: small spill/daytime release, small spill/nighttime release, large spill/daytime release, and large spill/nighttime release. The 90th-percentile safe distance values for each of these groups became the PADs that appear in the ERG2004.« less

Advanced risk assessment of the effects of graphite fibers on electronic and electric equipment, phase 1. [simulating vulnerability to airports and communities from fibers released during aircraft fires

NASA Technical Reports Server (NTRS)

Pocinki, L. S.; Kaplan, L. D.; Cornell, M. E.; Greenstone, R.

1979-01-01

A model was developed to generate quantitative estimates of the risk associated with the release of graphite fibers during fires involving commercial aircraft constructed with graphite fiber composite materials. The model was used to estimate the risk associated with accidents at several U.S. airports. These results were then combined to provide an estimate of the total risk to the nation.

Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

NASA Astrophysics Data System (ADS)

Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

40 CFR 68.115 - Threshold determination.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 16 2013-07-01 2013-07-01 false Threshold determination. 68.115 Section 68.115 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.115 Threshold determination. (a) A threshold...

40 CFR 68.115 - Threshold determination.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 16 2012-07-01 2012-07-01 false Threshold determination. 68.115 Section 68.115 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.115 Threshold determination. (a) A threshold...

40 CFR 68.115 - Threshold determination.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Threshold determination. 68.115 Section 68.115 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.115 Threshold determination. (a) A threshold...

40 CFR 68.115 - Threshold determination.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 16 2014-07-01 2014-07-01 false Threshold determination. 68.115 Section 68.115 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Regulated Substances for Accidental Release Prevention § 68.115 Threshold determination. (a) A threshold...

Preliminary Investigation of Time Remaining Display on the Computer-based Emergency Operating Procedure

NASA Astrophysics Data System (ADS)

Suryono, T. J.; Gofuku, A.

2018-02-01

One of the important thing in the mitigation of accidents in nuclear power plant accidents is time management. The accidents should be resolved as soon as possible in order to prevent the core melting and the release of radioactive material to the environment. In this case, operators should follow the emergency operating procedure related with the accident, in step by step order and in allowable time. Nowadays, the advanced main control rooms are equipped with computer-based procedures (CBPs) which is make it easier for operators to do their tasks of monitoring and controlling the reactor. However, most of the CBPs do not include the time remaining display feature which informs operators of time available for them to execute procedure steps and warns them if the they reach the time limit. Furthermore, the feature will increase the awareness of operators about their current situation in the procedure. This paper investigates this issue. The simplified of emergency operating procedure (EOP) of steam generator tube rupture (SGTR) accident of PWR plant is applied. In addition, the sequence of actions on each step of the procedure is modelled using multilevel flow modelling (MFM) and influenced propagation rule. The prediction of action time on each step is acquired based on similar case accidents and the Support Vector Regression. The derived time will be processed and then displayed on a CBP user interface.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walston, S; Rowland, M; Campbell, K

It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting inmore » a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.« less

The Causal Relationship between DNA Damage Induction in Bovine Lymphocytes and the Fukushima Nuclear Power Plant Accident

PubMed Central

Nakamura, Asako J.; Suzuki, Masatoshi; Redon, Christophe E.; Kuwahara, Yoshikazu; Yamashiro, Hideaki; Abe, Yasuyuki; Takahashi, Shintaro; Fukuda, Tomokazu; Isogai, Emiko; Bonner, William M.; Fukumoto, Manabu

2017-01-01

The Fukushima Daiichi Nuclear Power Plant (FNPP) accident, the largest nuclear incident since the 1986 Chernobyl disaster, occurred when the plant was hit by a tsunami triggered by the Great East Japan Earthquake on March 11, 2011. The subsequent uncontrolled release of radioactive substances resulted in massive evacuations in a 20-km zone. To better understand the biological consequences of the FNPP accident, we have been measuring DNA damage levels in cattle in the evacuation zone. DNA damage was evaluated by assessing the levels of DNA double-strand breaks in peripheral blood lymphocytes by immunocyto-fluorescence-based quantification of γ-H2AX foci. A greater than two-fold increase in the fraction of damaged lymphocytes was observed in all animal cohorts within the evacuation zone, and the levels of DNA damage decreased slightly over the 700-day sample collection period. While the extent of damage appeared to be independent of the distance from the accident site and the estimated radiation dose from radiocesium, we observed age-dependent accumulation of DNA damage. Thus, this study, which was the first to evaluate the biological impact of the FNPP accident utilizing the γ-H2AX assays, indicated the causal relation between high levels of DNA damage in animals living in the evacuation zone and the FNPP accident. PMID:28240558

The Causal Relationship between DNA Damage Induction in Bovine Lymphocytes and the Fukushima Nuclear Power Plant Accident.

PubMed

Nakamura, Asako J; Suzuki, Masatoshi; Redon, Christophe E; Kuwahara, Yoshikazu; Yamashiro, Hideaki; Abe, Yasuyuki; Takahashi, Shintaro; Fukuda, Tomokazu; Isogai, Emiko; Bonner, William M; Fukumoto, Manabu

2017-05-01

The Fukushima Daiichi Nuclear Power Plant (FNPP) accident, the largest nuclear incident since the 1986 Chernobyl disaster, occurred when the plant was hit by a tsunami triggered by the Great East Japan Earthquake on March 11, 2011. The subsequent uncontrolled release of radioactive substances resulted in massive evacuations in a 20-km zone. To better understand the biological consequences of the FNPP accident, we have been measuring DNA damage levels in cattle in the evacuation zone. DNA damage was evaluated by assessing the levels of DNA double-strand breaks in peripheral blood lymphocytes by immunocytofluorescence-based quantification of γ-H2AX foci. A greater than two-fold increase in the fraction of damaged lymphocytes was observed in all animal cohorts within the evacuation zone, and the levels of DNA damage decreased slightly over the 700-day sample collection period. While the extent of damage appeared to be independent of the distance from the accident site and the estimated radiation dose from radiocesium, we observed age-dependent accumulation of DNA damage. Thus, this study, which was the first to evaluate the biological impact of the FNPP accident utilizing the γ-H2AX assays, indicated the causal relation between high levels of DNA damage in animals living in the evacuation zone and the FNPP accident.

Worldwide isotope ratios of the Fukushima release and early-phase external dose reconstruction

PubMed Central

Chaisan, Kittisak; Smith, Jim T.; Bossew, Peter; Kirchner, Gerald; Laptev, Gennady V.

2013-01-01

Measurements of radionuclides (RNs) in air made worldwide following the Fukushima accident are quantitatively compared with air and soil measurements made in Japan. Isotopic ratios RN:137Cs of 131I, 132Te, 134,136Cs, are correlated with distance from release. It is shown, for the first time, that both within Japan and globally, ratios RN:137Cs in air were relatively constant for primarily particle associated radionuclides (134,136Cs; 132Te) but that 131I shows much lower local (<80 km) isotope ratios in soils relative to 137Cs. Derived isotope ratios are used to reconstruct external dose rate during the early phase post-accident. Model “blind” tests show more than 95% of predictions within a factor of two of measurements from 15 sites to the north, northwest and west of the power station. It is demonstrated that generic isotope ratios provide a sound basis for reconstruction of early-phase external dose rates in these most contaminated areas. PMID:24018776

Determination of the 240Pu/ 239Pu atomic ratio in soils from Palomares (Spain) by low-energy accelerator mass spectrometry

NASA Astrophysics Data System (ADS)

Chamizo, E.; García-León, M.; Synal, H.-A.; Suter, M.; Wacker, L.

2006-08-01

In 1966, the nuclear fuel of two thermonuclear bombs was released over the Spanish region of Palomares, due to a B52 bomber accident during a refuelling operation. Since then, much effort has been made to assess its impact to the different environmental compartments of this area in South-East Spain, mostly by measuring the 239+240Pu activity concentration and the 238Pu/239+240Pu activity ratio. Nevertheless, these measurements do not give enough information on the problem. In order to recognize unambiguously small traces of the weapon-grade plutonium released in the accident, the ratio of the two major isotopes of plutonium, 240Pu/239Pu, has to be determined. In this work, this ratio has been measured in low- and high-activity samples from Palomares by means of low-energy accelerator mass spectrometry (AMS). That way, we will show the potential of the new generation of compact AMS facilities in terms of plutonium characterization at ultra-trace levels.

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.

The role of hazardousness and regulatory practice in the accidental release of chemicals at U.S. industrial facilities.

PubMed

Elliott, Michael R; Keindorfer, Paul R; Lowe, Robert A

2003-10-01

This article presents the results of an analysis of the accident history data reported under section 112(r) of the Clean Air Act Amendments. These data provide a fairly complete record of the consequences of reportable accidental releases occurring during the time frame 1995-1999 in the U.S. chemical industry and covering 77 toxic and 63 flammable substances subject to the provisions of section 112(r). As such, these results are of fundamental interest to the affected communities, regulators, and insurers, as well as to owners and managers in the chemical industry. The results show the statistical associations between accident frequency and severity and a number of characteristics of reporting facilities, including their size, the hazardousness of the processes and chemicals inventoried, and the regulatory programs (in addition to section 112(r)) to which these facilities are subject. The results are interpreted in light of economic drivers of protective activity and regulatory priorities for monitoring and enforcement.

Newton’s Rotating Water Bucket: A Simple Model

DTIC Science & Technology

2013-01-01

surface of the water must be an equipotential relative to the sum of the gravitational and centrifugal potential energies [9].) The value of z0 can be...twisted and the bucket is then released, it begins to spin and the surface of the water acquires a paraboloidal shape. In this paper, the parabolic...adopts a curved surface . Ernst Mach, for example, postulated that the parabolic shape must be due to the existence of matter in the universe

Salmon migration patterns revealed the temporal and spatial fluctuations of the radiocesium levels in terrestrial and ocean environments.

PubMed

Arai, Takaomi

2014-01-01

The disabling of the Fukushima Daiichi Nuclear Power Plant (F1NPP) resulted in the release of radionuclides, including 134Cs and 137Cs, into the air and the ocean. The unpredicted nuclear accident is of global concern for human health and the ecosystem. Although investigations of radionuclides in environments were performed shortly after the accident started, the temporal and spatial impacts and fluctuations on the releasing radionuclides to natural environment remain unclear. I focused on salmon, which migrate from inland to the open ocean globally, to reveal the three-year (May 2011 to February 2014) fluctuations and accumulations of 134Cs and 137Cs from terrestrial to open ocean environments after the F1NPP accident. The 134Cs and 137Cs concentrations in six salmonids exhibited lower temporal variations for three years after the F1NPP accident, suggesting that these radionuclides are widely distributed and these radionuclides remain in the natural environment globally with less convergence. The accumulation patterns were significantly different among the different salmon species. Fluvial (freshwater residence) type salmons exhibited significantly higher accumulation in 134Cs (25.3-40.2 Bq kg(-1) in mean) and 137Cs (41.4-51.7 Bq kg(-1) in mean) than did the anadromous (sea-run) type salmons (0.64-8.03 Bq kg(-1) in mean 134Cs and 0.42-10.2 Bq kg(-1) in mean 137Cs) suggesting widespread contamination in terrestrial environments versus the coastal and open ocean environments. Salmonids are the most highly migratory animals and are characterised by their strong tendency to return home to their natal site for reproduction. Salmonids have a potential to be a good indicator as an effective monitoring animal.

Fukushima-derived fission nuclides monitored around Taiwan: Free tropospheric versus boundary layer transport

NASA Astrophysics Data System (ADS)

Huh, Chih-An; Hsu, Shih-Chieh; Lin, Chuan-Yao

2012-02-01

The 2011 Fukushima nuclear accident in Japan was the worst nuclear disaster following the 1986 Chernobyl accident. Fission products (nuclides) released from the Fukushima plant site since March 12, 2011 had been detected around the northern hemisphere in about two weeks and also in the southern hemisphere about one month later. We report here detailed time series of radioiodine and radiocesium isotopes monitored in a regional network around Taiwan, including one high-mountain and three ground-level sites. Our results show several pulses of emission from a sequence of accidents in the Fukushima facility, with the more volatile 131I released preferentially over 134Cs and 137Cs at the beginning. In the middle of the time series, there was a pronounced peak of radiocesium observed in northern Taiwan, with activity concentrations of 134Cs and 137Cs far exceeding that of 131I during that episode. From the first arrival time of these fission nuclides and their spatial and temporal variations at our sampling sites and elsewhere, we suggest that Fukushima-derived radioactive nuclides were transported to Taiwan and its vicinity via two pathways at different altitudes. One was transported in the free troposphere by the prevailing westerly winds around the globe; the other was transported in the planetary boundary layer by the northeast monsoon wind directly toward Taiwan.

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Consequences of the nuclear power plant accident at Chernobyl.

PubMed Central

Ginzburg, H M; Reis, E

1991-01-01

The Chernobyl Nuclear Power Plant accident, in the Ukrainian Soviet Socialist Republic (SSR), on April 26, 1986, was the first major nuclear power plant accident that resulted in a large-scale fire and subsequent explosions, immediate and delayed deaths of plant operators and emergency service workers, and the radioactive contamination of a significant land area. The release of radioactive material, over a 10-day period, resulted in millions of Soviets, and other Europeans, being exposed to measurable levels of radioactive fallout. Because of the effects of wind and rain, the radioactive nuclide fallout distribution patterns are not well defined, though they appear to be focused in three contiguous Soviet Republics: the Ukrainian SSR, the Byelorussian SSR, and the Russian Soviet Federated Socialist Republic. Further, because of the many radioactive nuclides (krypton, xenon, cesium, iodine, strontium, plutonium) released by the prolonged fires at Chernobyl, the long-term medical, psychological, social, and economic effects will require careful and prolonged study. Specifically, studies on the medical (leukemia, cancers, thyroid disease) and psychological (reactive depressions, post-traumatic stress disorders, family disorganization) consequences of continued low dose radiation exposure in the affected villages and towns need to be conducted so that a coherent, comprehensive, community-oriented plan may evolve that will not cause those already affected any additional harm and confusion. Images p38-a p38-b PMID:1899937

Concentration of strontium-90 at selected hot spots in Japan.

PubMed

Steinhauser, Georg; Schauer, Viktoria; Shozugawa, Katsumi

2013-01-01

This study is dedicated to the environmental monitoring of radionuclides released in the course of the Fukushima nuclear accident. The activity concentrations of β(-)-emitting (90)Sr and β(-)/γ-emitting (134)Cs and (137)Cs from several hot spots in Japan were determined in soil and vegetation samples. The (90)Sr contamination levels of the samples were relatively low and did not exceed the Bq⋅g(-1) range. They were up four orders of magnitude lower than the respective (137)Cs levels. This study, therefore, experimentally confirms previous predictions indicating a low release of (90)Sr from the Fukushima reactors, due to its low volatility. The radiocesium contamination could be clearly attributed to the Fukushima nuclear accident via its activity ratio fingerprint ((134)Cs/(137)Cs). Although the correlation between (90)Sr and (137)Cs is relatively weak, the data set suggests an intrinsic coexistence of both radionuclides in the contaminations caused by the Fukushima nuclear accident. This observation is of great importance not only for remediation campaigns but also for the current food monitoring campaigns, which currently rely on the assumption that the activity concentrations of β(-)-emitting (90)Sr (which is relatively laborious to determine) is not higher than 10% of the level of γ-emitting (137)Cs (which can be measured quickly). This assumption could be confirmed for the samples investigated herein.

Differential calcium dependence in basal and forskolin-potentiated spontaneous transmitter release in basolateral amygdala neurons.

PubMed

Miura, Yuki; Naka, Masamitsu; Matsuki, Norio; Nomura, Hiroshi

2012-10-31

Action potential-independent transmitter release, or spontaneous release, is postulated to produce multiple postsynaptic effects (e.g., maintenance of dendritic spines and suppression of local dendritic protein synthesis). Potentiation of spontaneous release may contribute to the precise modulation of synaptic function. However, the expression mechanism underlying potentiated spontaneous release remains unclear. In this study, we investigated the involvement of extracellular and intracellular calcium in basal and potentiated spontaneous release. Miniature excitatory postsynaptic currents (mEPSCs) of the basolateral amygdala neurons in acute brain slices were recorded. Forskolin, an adenylate cyclase activator, increased mEPSC frequency, and the increase lasted at least 25 min after washout. Removal of the extracellular calcium decreased mEPSC frequency in both naïve and forskolin-treated slices. On the other hand, chelation of intracellular calcium by BAPTA-AM decreased mEPSC frequency in naïve, but not in forskolin-treated slices. A blockade of the calcium-sensing receptor (CaSR) resulted in an increase in mEPSC frequency in forskolin-treated, but not in naïve slices. These findings indicate that forskolin-induced potentiation is accompanied by changes in the mechanisms underlying Ca(2+)-dependent spontaneous release. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.

Influence of Organic Matter - Mineral Interfacial Reactions on Metal(loid) Speciation and Bioaccessibility

NASA Astrophysics Data System (ADS)

Chorover, J.; Kong, S.; Root, R. A.; Thomas, A.

2015-12-01

Bioaccessibility of contaminant metals in geomedia is often measured on the basis of kinetic release to solution during in vitro reaction with biofluid simulants. We postulate that development of a predictive-mechanistic understanding of bioaccessibility requires knowledge of metal(loid) molecular speciation upon sample introduction, as well as its change over the course of the in vitro reaction. Our results - including data from batch, column, mesocosm and field studies pertaining to arsenic, lead, and zinc contaminated materials - indicate the strong influence of organic matter and associated biological activity on metal(loid) speciation in mine tailings and related model systems. Furthermore, presence/absence of organic matter during bioassays affects the kinetics of metal(loid) release into biofluid simulants through multiple mechanisms.

Historical analysis of US pipeline accidents triggered by natural hazards

NASA Astrophysics Data System (ADS)

Girgin, Serkan; Krausmann, Elisabeth

2015-04-01

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

Non-accident release of hazmat from railroad tank cars : training issues

DOT National Transportation Integrated Search

1999-07-01

This report presents the results of a study to determine the extent to which written materials used in training railroad tank car HAZMAT loaders are consistent with the reading skills of the trainees. Data from four case study sites was used to condu...

Identifying Patterns in the Weather of Europe for Source Term Estimation

NASA Astrophysics Data System (ADS)

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

2017-04-01

During emergencies that involve the release of hazardous substances into the atmosphere the potential health effects on the human population and the environment are of primary concern. Such events have occurred in the past, most notably involving radioactive and toxic substances. Examples of radioactive release events include the Chernobyl accident in 1986, as well as the more recent Fukushima Daiichi accident in 2011. Often, the release of dangerous substances in the atmosphere is detected at locations different from the release origin. The objective of this work is the rapid estimation of such unknown sources shortly after the detection of dangerous substances in the atmosphere, with an initial focus on nuclear or radiological releases. Typically, after the detection of a radioactive substance in the atmosphere indicating the occurrence of an unknown release, the source location is estimated via inverse modelling. However, depending on factors such as the spatial resolution desired, traditional inverse modelling can be computationally time-consuming. This is especially true for cases where complex topography and weather conditions are involved and can therefore be problematic when timing is critical. Making use of machine learning techniques and the Big Data Europe platform1, our approach moves the bulk of the computation before any such event taking place, therefore allowing for rapid initial, albeit rougher, estimations regarding the source location. Our proposed approach is based on the automatic identification of weather patterns within the European continent. Identifying weather patterns has long been an active research field. Our case is differentiated by the fact that it focuses on plume dispersion patterns and these meteorological variables that affect dispersion the most. For a small set of recurrent weather patterns, we simulate hypothetical radioactive releases from a pre-known set of nuclear reactor locations and for different substance and temporal parameters, using the Java flavour of the Euratom-supported funded RODOS (Real-time On-line DecisiOn Support) system2 for off-site emergency management after nuclear accidents. Once dispersions have been pre-computed, and immediately after a detected release, the currently observed weather can be matched to the derived weather classes. Since each weather class corresponds to a different plume dispersion pattern, the closest classes to an unseen weather sample, say the current weather, are the most likely to lead us to the release origin. In addressing the above problem, we make use of multiple years of weather reanalysis data from NCAR's version3 of ECMWF's ERA-Interim4. To derive useful weather classes, we evaluate several algorithms, ranging from straightforward unsupervised clustering to more complex methods, including relevant neural-network algorithms, on multiple variables. Variables and feature sets, clustering algorithms and evaluation approaches are all dealt with and presented experimentally. The Big Data Europe platform allows for the implementation and execution of the above tasks in the cloud, in a scalable, robust and efficient way.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Sleep Deprivation Decreases [11C]Raclopride’s Binding to Dopamine D2/D3 Receptors in the Human Brain

PubMed Central

Volkow, Nora D.; Wang, Gene-Jack; Telang, Frank; Fowler, Joanna S.; Logan, Jean; Wong, Christopher; Ma, Jim; Pradhan, Kith; Tomasi, Dardo; Thanos, Peter K.; Ferré, Sergi; Jayne, Millard

2009-01-01

Sleep deprivation can markedly impair human performance contributing to accidents and poor productivity. The mechanisms underlying this impairment are not well understood but brain dopamine systems have been implicated. Here we test whether one night of sleep deprivation changes dopamine brain activity. We studied fifteen healthy subjects using positron emission tomography and [11C]raclopride (dopamine D2/3 receptor radioligand) and [11C]cocaine (dopamine transporter radioligand). Subjects were tested twice; after one night of rested sleep and after on night of sleep deprivation. [11C]Raclopride’s specific binding in striatum and thalamus were significantly reduced after sleep deprivation and the magnitude of this reduction correlated with increases in fatigue (tiredness and sleepiness) and with deterioration in cognitive performance (visual attention and working memory). In contrast sleep deprivation did not affect the specific binding of [11C]cocaine in striatum. Since [11C]raclopride competes with endogenous dopamine for binding to D2/D3 receptors, we interpret the decreases in binding to reflect dopamine increases with sleep deprivation. However, we can not rule out the possibility that decreased [11C]raclopride binding reflects decreases in receptor levels or affinity. Sleep deprivation did not affect dopamine transporters (target for most wake-promoting medications) and thus dopamine increases are likely to reflect increases in dopamine cell firing and/or release rather than decreases in dopamine reuptake. Inasmuch as dopamine-enhancing drugs increase wakefulness we postulate that dopamine increases after sleep deprivation is a mechanism by which the brain maintains arousal as the drive to sleep increases but one that is insufficient to counteract behavioral and cognitive impairment. PMID:18716203

Fukushima Daiichi Nuclear Power Plant accident: facts, environmental contamination, possible biological effects, and countermeasures.

PubMed

Anzai, Kazunori; Ban, Nobuhiko; Ozawa, Toshihiko; Tokonami, Shinji

2012-01-01

On March 11, 2011, an earthquake led to major problems at the Fukushima Daiichi Nuclear Power Plant. A 14-m high tsunami triggered by the earthquake disabled all AC power to Units 1, 2, and 3 of the Power Plant, and carried off fuel tanks for emergency diesel generators. Despite many efforts, cooling systems did not work and hydrogen explosions damaged the facilities, releasing a large amount of radioactive material into the environment. In this review, we describe the environmental impact of the nuclear accident, and the fundamental biological effects, acute and late, of the radiation. Possible medical countermeasures to radiation exposure are also discussed.

[Assessment and control of health risk caused by the radiological accident at the TEPCO Fukushima Daiichi nuclear power plant].

PubMed

Matsuda, Naoki; Morita, Naoko; Miura, Miwa

2014-01-01

The accident at the Tokyo Electric Power Company (TEPCO) Fukushima Daiichi nuclear power plant on March 11, 2011, released a large amount of radioactive materials resulting in the radioactive contamination of a wide area of eastern Japan. Residents of the Fukushima prefecture experienced various unavoidable damages and fear of radiation effects on their health. A reliable communication of accurate risk assessment for residents is required as a countermeasure aimed at the reconstruction of Fukushima. Here, the current status of individual dose estimation and the issues relating to the radiation risk perception are discussed.

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Health Physics Code System for Evaluating Accidents Involving Radioactive Materials.

DOE Office of Scientific and Technical Information (OSTI.GOV)

2014-10-01

Version 03 The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculational tool for evaluating accidents involving radioactive materials. HOTSPOT codes provide a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. The developer's website is: http://www.llnl.gov/nhi/hotspot/. Four general programs, PLUME, EXPLOSION, FIRE, and RESUSPENSION, calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Additional programs deal specifically with the release of plutonium, uranium, and tritium to expedite an initial assessmentmore » of accidents involving nuclear weapons. The FIDLER program can calibrate radiation survey instruments for ground survey measurements and initial screening of personnel for possible plutonium uptake in the lung. The HOTSPOT codes are fast, portable, easy to use, and fully documented in electronic help files. HOTSPOT supports color high resolution monitors and printers for concentration plots and contours. The codes have been extensively used by the DOS community since 1985. Tables and graphical output can be directed to the computer screen, printer, or a disk file. The graphical output consists of dose and ground contamination as a function of plume centerline downwind distance, and radiation dose and ground contamination contours. Users have the option of displaying scenario text on the plots. HOTSPOT 3.0.1 fixes three significant Windows 7 issues: Executable installed properly under "Program Files/HotSpot 3.0". Installation package now smaller: removed dependency on older Windows DLL files which previously needed to; Forms now properly scale based on DPI instead of font for users who change their screen resolution to something other than 100%. This is a more common feature in Windows 7; Windows installer was starting everytime most users started the program, even after HotSpot was already installed. Now, after the program is installed the installer may come up once for each new user but only the first time they run HotSpot on a particular machine. So no user should see the installer come up more than once over many uses; and GPS capability updated to directly use a serial port through a USB connection. Non-USB connections should still work. Fixed table output inconsistencies for fire scenarios.« less

Stochastic Plume Simulations for the Fukushima Accident and the Deep Water Horizon Oil Spill

NASA Astrophysics Data System (ADS)

Coelho, E.; Peggion, G.; Rowley, C.; Hogan, P.

2012-04-01

The Fukushima Dai-ichi power plant suffered damage leading to radioactive contamination of coastal waters. Major issues in characterizing the extent of the affected waters were a poor knowledge of the radiation released to the coastal waters and the rather complex coastal dynamics of the region, not deterministically captured by the available prediction systems. Equivalently, during the Gulf of Mexico Deep Water Horizon oil platform accident in April 2010, significant amounts of oil and gas were released from the ocean floor. For this case, issues in mapping and predicting the extent of the affected waters in real-time were a poor knowledge of the actual amounts of oil reaching the surface and the fact that coastal dynamics over the region were not deterministically captured by the available prediction systems. To assess the ocean regions and times that were most likely affected by these accidents while capturing the above sources of uncertainty, ensembles of the Navy Coastal Ocean Model (NCOM) were configured over the two regions (NE Japan and Northern Gulf of Mexico). For the Fukushima case tracers were released on each ensemble member; their locations at each instant provided reference positions of water volumes where the signature of water released from the plant could be found. For the Deep Water Horizon oil spill case each ensemble member was coupled with a diffusion-advection solution to estimate possible scenarios of oil concentrations using perturbed estimates of the released amounts as the source terms at the surface. Stochastic plumes were then defined using a Risk Assessment Code (RAC) analysis that associates a number from 1 to 5 to each grid point, determined by the likelihood of having tracer particle within short ranges (for the Fukushima case), hence defining the high risk areas and those recommended for monitoring. For the Oil Spill case the RAC codes were determined by the likelihood of reaching oil concentrations as defined in the Bonn Agreement Oil Appearance Code. The likelihoods were taken in both cases from probability distribution functions derived from the ensemble runs. Results were compared with a control-deterministic solution and checked against available reports to assess their skill in capturing the actual observed plumes and other in-situ data, as well as their relevance for planning surveys and reconnaissance flights for both cases.

Validation of Operational Multiscale Environment Model With Grid Adaptivity (OMEGA).

DTIC Science & Technology

1995-12-01

Center for the period of the Chernobyl Nuclear Accident. The physics of the model is tested using National Weather Service Medium Range Forecast data by...Climatology Center for the first three days following the release at the Chernobyl Nuclear Plant. A user-defined source term was developed to simulate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curry, J J; Gallagher, D W; Modarres, M

Appendices are presented concerning isolation condenser makeup; vapor suppression system; station air system; reactor building closed cooling water system; turbine building secondary closed water system; service water system; emergency service water system; fire protection system; emergency ac power; dc power system; event probability estimation; methodology of accident sequence quantification; and assignment of dominant sequences to release categories.

40 CFR 370.42 - What is Tier II inventory information?

Code of Federal Regulations, 2014 CFR

2014-07-01

... numbers assigned under the Toxic Release Inventory (TRI) and Risk Management Program. If your facility has... Accident Prevention Provisions, also known as the Risk Management Program. (m) The name, mailing address... year. (s) For each hazardous chemical that you are required to report, you must: (1) Pure Chemical...

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena andmore » presents major conclusions on the state of the art.« less

Gas Retention, Gas Release, and Fluidization of Spherical Resorcinol-Formaldehyde (sRF) Ion Exchange Resin

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauglitz, Phillip A.; Rassat, Scot D.; Linn, Diana

The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs in ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. When loaded with radioactive Cs, radiolysis of water in the LAW liquid will generate hydrogen gas. In normal operations, the generated hydrogen is expected to remainmore » dissolved in the liquid and be continuously removed by liquid flow. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin bed and below the bottom screen that supports the resin within the column, which creates a hydrogen flammability hazard. Because there is a potential for a large fraction of the retained hydrogen to be released over a short duration as a gas release event, there is a need to quantify the size and rate of potential gas release events. Due to the potential for a large, rapid gas release event, an evaluation of mitigation methods to eliminate the hydrogen hazard is also needed. One method being considered for mitigating the hydrogen hazard during a loss of flow accident is to have a secondary flow system, with two redundant pumps operating in series, that re-circulates liquid upwards through the bed and into a vented break tank where hydrogen gas is released from the liquid and removed by venting the headspace of the break tank. The mechanism for inducing release of gas from the sRF bed is to fluidize the bed, which should allow retained bubbles to rise and be carried to the break tank. The overall conclusion of the testing is that fluidization is an effective method to remove hydrogen gas from a bed of sRF resin, but that a single fluidization velocity that is adequate to release gas in 55 ºC water will over-fluidize sRF resin in most LAW liquids, including both nominal and high-limit LAW simulants used in testing. An upper packed bed can retain hydrogen gas and pose a flammability hazard. Using periodic on:off fluidization, such as 5:55 min. on:off cycles, is effective at releasing gas while not creating an upper packed bed. Note that lengthening the fluidization duration in a one-hour cycle did result in a stable upper packed bed in one case with the nominal LAW simulant, so testing focused on shorter “on” periods which are needed for effective hydrogen release with periodic on:off fluidization« less

GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise Paul

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. •more » The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added« less

Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.

2014-09-01

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparisonmore » of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary.« less

Particle size distribution of radioactive aerosols after the Fukushima and the Chernobyl accidents.

PubMed

Malá, Helena; Rulík, Petr; Bečková, Vera; Mihalík, Ján; Slezáková, Miriam

2013-12-01

Following the Fukushima accident, a series of aerosol samples were taken between 24th March and 13th April 2011 by cascade impactors in the Czech Republic to obtain the size distribution of (131)I, (134)Cs, (137)Cs, and (7)Be aerosols. All distributions could be considered monomodal. The arithmetic means of the activity median aerodynamic diameters (AMADs) for artificial radionuclides and for (7)Be were 0.43 and 0.41 μm with GDSs 3.6 and 3.0, respectively. The time course of the AMADs of (134)Cs, (137)Cs and (7)Be in the sampled period showed a slight decrease at a significance level of 0.05, whereas the AMAD pertaining to (131)I increased at a significance level of 0.1. Results obtained after the Fukushima accident were compared with results obtained after the Chernobyl accident. The radionuclides released during the Chernobyl accident for which we determined the AMAD fell into two categories: refractory radionuclides ((140)Ba, (140)La (141)Ce, (144)Ce, (95)Zr and (95)Nb) and volatile radionuclides ((134)Cs, (137)Cs, (103)Ru, (106)Ru, (131)I, and (132)Te). The AMAD of the refractory radionuclides was approximately 3 times higher than the AMAD of the volatile radionuclides; nevertheless, the size distributions for volatile radionuclides having a mean AMAD value of 0.51 μm were very close to the distributions after the Fukushima accident. Copyright © 2013 Elsevier Ltd. All rights reserved.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harper, F.T.; Young, M.L.; Miller, L.A.

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 formulatedmore » 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.« less

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

NASA Astrophysics Data System (ADS)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

What Happened, and Why: Toward an Understanding of Human Error Based on Automated Analyses of Incident Reports. Volume 1

NASA Technical Reports Server (NTRS)

Maille, Nicolas P.; Statler, Irving C.; Ferryman, Thomas A.; Rosenthal, Loren; Shafto, Michael G.; Statler, Irving C.

2006-01-01

The objective of the Aviation System Monitoring and Modeling (ASMM) project of NASA s Aviation Safety and Security Program was to develop technologies that will enable proactive management of safety risk, which entails identifying the precursor events and conditions that foreshadow most accidents. This presents a particular challenge in the aviation system where people are key components and human error is frequently cited as a major contributing factor or cause of incidents and accidents. In the aviation "world", information about what happened can be extracted from quantitative data sources, but the experiential account of the incident reporter is the best available source of information about why an incident happened. This report describes a conceptual model and an approach to automated analyses of textual data sources for the subjective perspective of the reporter of the incident to aid in understanding why an incident occurred. It explores a first-generation process for routinely searching large databases of textual reports of aviation incident or accidents, and reliably analyzing them for causal factors of human behavior (the why of an incident). We have defined a generic structure of information that is postulated to be a sound basis for defining similarities between aviation incidents. Based on this structure, we have introduced the simplifying structure, which we call the Scenario as a pragmatic guide for identifying similarities of what happened based on the objective parameters that define the Context and the Outcome of a Scenario. We believe that it will be possible to design an automated analysis process guided by the structure of the Scenario that will aid aviation-safety experts to understand the systemic issues that are conducive to human error.

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

Evaluation of micron size carbon fibers released from burning graphite composites

NASA Technical Reports Server (NTRS)

Sussholz, B.

1980-01-01

Quantitative estimates were developed of micron carbon fibers released during the burning of graphite composites. Evidence was found of fibrillated particles which were the predominant source of the micron fiber data obtained from large pool fire tests. The fibrillation phenomena were attributed to fiber oxidation effects caused by the fire environment. Analysis of propane burn test records indicated that wind sources can cause considerable carbon fiber oxidation. Criteria estimates were determined for the number of micron carbon fibers released during an aircraft accident. An extreme case analysis indicated that the upper limit of the micron carbon fiber concentration level was only about half the permissible asbestos ceiling concentration level.

RCRA, Superfund and EPCRA hotline training module. Introduction to: CERCLA and EPCRA release reporting requirements (CERCLA section 103 and EPCRA section 304)

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1998-06-01

The release reporting requirements set out in the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Emergency Planning and Community Right-to-Know Act (EPCRA) enable federal, state, and local authorities to effectively prepare for and respond to chemical accidents. This module reviews the regulations found at 40 CFR Part 302 promulgated pursuant to CERCLA section 103, and the regulations found at 40 CFR section 355.40 promulgated pursuant to EPCRA section 304. The goal of this module is to explain the notification requirements triggered by releases of CERCLA hazardous substances and EPCRA-designated extremely hazardous substances (EHSs).

Control of accidental releases of hydrogen selenide in vented storage cabinets

NASA Astrophysics Data System (ADS)

Fthenakis, V. M.; Moskowitz, P. D.; Sproull, R. D.

1988-07-01

Highly toxic hydrogen selenide and hydrogen sulfide gases are used in the production of copper-indium-diselenide photovoltaic cells by reactive sputtering. In the event of an accident, these gases may be released to the atmosphere and pose hazards to public and occupational safety and health. This paper outlines an approach for designing systems for the control of these releases given the uncertainty in release conditions and lack of data on the chemical systems involved. Accidental releases of these gases in storage cabinets can be controlled by either a venturi and packed-bed scrubber and carbon adsorption bed, or containment scrubbing equipment followed by carbon adsorption. These systems can effectively reduce toxic gas emissions to levels needed to protect public health. The costs of these controls (˜0.012/Wp) are samll in comparison with current (˜6/Wp) and projected (˜I/Wp) production costs.

A Possible Organic Contribution to the Low Temperature CO2 Release Seen in Mars Phoenix Thermal and Evolved Gas Analyzer Data

NASA Technical Reports Server (NTRS)

Archer, P. D. Jr.; Lauer, H. V., Jr.; Sutter, B.; Ming, D. W.; Niles, P. B.; Boynton, W. V.

2012-01-01

Two of the most important discoveries of the Phoenix Mars Lander were the discovery of approx.0.6% perchlorate [1] and 3-5% carbonate [2] in the soils at the landing site in the martian northern plains. The Thermal and Evolved Gas Analyzer (TEGA) instrument was one of the tools that made this discovery. After soil samples were delivered to TEGA and transferred into small ovens, the samples could be heated up to approx.1000 C and the gases that evolved during heating were monitored by a mass spectrometer. A CO2 signal was detected at high temperature (approx.750 C) that has been attributed to calcium carbonate decomposition. In addition to this CO2 release, a lower temperature signal was seen. This lower temperature CO2 release was postulated to be one of three things: 1) desorption of CO2, 2) decomposition of a different carbonate mineral, or 3) CO2 released due to organic combustion. Cannon et al. [3] present another novel hypothesis involving the interaction of decomposition products of a perchlorate salt and calcium carbonate.

Anaerobic phosphate release from activated sludge with enhanced biological phosphorus removal. A possible mechanism of intracellular pH control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bond, P.L.; Keller, J.; Blackall, L.L.

The biochemical mechanisms of the wastewater treatment process known as enhanced biological phosphorus removal (EBPR) are presently described in a metabolic model. The authors investigated details of the EBPR model to determine the nature of the anaerobic phosphate release and how this may be metabolically associated with polyhydroxyalkanoate (PHA) formation. Iodoacetate, an inhibitor of glycolysis, was found to inhibit the anaerobic formation of PHA and phosphate release, supporting the pathways proposed in the EBPR metabolic model. In the metabolic model, it is proposed that polyphosphate degradation provides energy for the microorganisms in anaerobic regions of these treatment systems. Other investigationsmore » have shown that anaerobic phosphate release depends on the extracellular pH. The authors observed that when the intracellular pH of EBPR sludge was raised, substantial anaerobic phosphate release was caused without volatile fatty acid (VFA) uptake. Acidification of the sludge inhibited anaerobic phosphate release even in the presence of VFA. from these observations, the authors postulate that an additional possible role of anaerobic polyphosphate degradation in EBPR is for intracellular pH control. Intracellular pH control may be a metabolic feature of EBPR, not previously considered, that could have some use in the control and optimization of EBPR.« less

Return to Flight Task Group

NASA Technical Reports Server (NTRS)

2005-01-01

It has been 29 months since Columbia was lost over East Texas in February 2003. Seven months after the accident, the Columbia Accident Investigation Board (CAIB) released the first volume of its final report, citing a variety of technical, managerial, and cultural issues within NASA and the Space Shuttle Program. To their credit, NASA offered few excuses, embraced the report, and set about correcting the deficiencies noted by the accident board. Of the 29 recommendations issued by the CAIB, 15 were deemed critical enough that the accident board believed they should be implemented prior to returning the Space Shuttle to flight. Some of these recommendations were relatively easy, most were straightforward, a few bordered on the impossible, and others were largely overcome by events, particularly the decision by the President to retire the Space Shuttle by 2010. The Return to Flight Task Group (RTF TG, or simply, the Task Group) was chartered by the NASA Administrator in July 2003 to provide an independent assessment of the implementation of the 15 CAIB return-to-flight recommendations. An important observation must be stated up-front: neither the CAIB nor the RTF TG believes that all risk can be eliminated from Space Shuttle operations; nor do we believe that the Space Shuttle is inherently unsafe. What the CAIB and RTF TG do believe, however, is that NASA and the American public need to understand the risks associated with space travel, and that NASA must make every reasonable effort to minimize such risk. Since the release of the CAIB report, NASA and the Space Shuttle Program expended enormous effort and resources toward correcting the causes of the accident and preparing to fly again. Relative to the 15 specific recommendations that the CAIB indicated should be implemented prior to returning to flight, NASA has met or exceeded most of them the Task Group believes that NASA met the intent of the CAIB for 12 of these recommendations. The remaining three recommendations were so challenging that NASA could not comply completely with the intent of the CAIB.

Comparison of the Kinetic Rate Law Parameters for the Dissolution of Natural and Synthetic Autunite in the Presence of Aqueous Bicarbonate Ions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gudavalli, Ravi; Katsenovich, Yelena; Wellman, Dawn M.

2013-08-02

Bicarbonate is one of the most significant components within the uranium geochemical cycle. In aqueous solutions, bicarbonate forms strong complexes with uranium. As such, aqueous bicarbonate may significantly increase the rate of uranium release from uranium minerals. Quantifying the relationship of aqueous bicarbonate concentration to the rate of uranium release during dissolution is critical to understanding the long-term fate of uranium within the environment. Single-pass flow-through (SPTF) experiments were conducted to estimate the rate of uranium release from Na meta-autunite as a function of bicarbonate (0.0005-0.003 M) under the pH range of 6-11 and a temperature range of 5-60oC. Consistentmore » with the results of previous investigation, the rate of uranium release exhibited minimal dependency on temperature; but were strongly dependent on pH. Increasing aqueous bicarbonate concentrations afforded comparable increases in the rate of release of uranium. Most notably under low pH conditions the aqueous bicarbonate resulted in up to 370 fold increases in the rate of uranium release in relative to the rate of uranium release in the absence of bicarbonate. However, the effect of aqueous bicarbonate on the release of uranium was significantly less under higher pH conditions. It is postulated that at high pH values, surface sites are saturated with carbonate, thus the addition of more bicarbonate would have less effect on uranium release.« less

Honokiol Dimers and Magnolol Derivatives with New Carbon Skeletons from the Roots of Magnolia officinalis and Their Inhibitory Effects on Superoxide Anion Generation and Elastase Release

PubMed Central

Chen, Hung-Chung; Kuo, Ping-Chung; Lee, E-Jian; Lee, Kuo-Hsiung; Wu, Tian-Shung

2013-01-01

Two honokiol dimers, houpulins A and B (1 and 2), and two magnolol derivatives, houpulins C and D (3 and 4), were isolated and characterized from an ethanol extract obtained from the roots of Magnolia officinalis. The chemical structures were determined based on spectroscopic and physicochemical analyses, which included 1D and 2D NMR, as well as mass spectrometry data. These four oligomers possess new carbon skeletons postulated to be biosynthesized from the coupling of three or four C6-C3 subunits. In addition, the new oligomers were evaluated for inhibition of superoxide anion generation and elastase release, and houpulin B (2) was identified as a new anti-inflammatory lead compound. PMID:23667420

Novel method for the determination of Sildenafil (Viagra) and its metabolite (UK-103,320) in postmortem specimens using LC/MS/MS and LC/MS/MS/MS.

DOT National Transportation Integrated Search

2000-05-01

During the investigation of aviation accidents, postmortem samples from victims are submitted to the FAAs : Civil Aeromedical Institute for drug analysis. Because new drugs are continually being released to the market, it : is our laboratorys r...

RadNet Map Interface for Near-Real-Time Radiation Monitoring Data

EPA Pesticide Factsheets

RadNet is a national network of monitoring stations that regularly collect air, precipitation, drinking water, and milk samples for analysis of radioactivity. The RadNet network, which has stations in each state, has been used to track environmental releases of radioactivity from nuclear weapons tests and nuclear accidents.

75 FR 62567 - Notice of Lodging of Consent Decree Under the Clean Air Act

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-12

...) and the Chemical Accident Prevention Provisions promulgated at 40 CFR part 68. The United States' CAA... also requires BP Products to regularly report to EPA on indicators of process safety at the Texas City... received process safety training, and (3) whether additional accidental releases of regulated substances...

RadNet Air Quality (Fixed Station) Data

EPA Pesticide Factsheets

RadNet is a national network of monitoring stations that regularly collect air for analysis of radioactivity. The RadNet network, which has stations in each State, has been used to track environmental releases of radioactivity from nuclear weapons tests and nuclear accidents. RadNet also documents the status and trends of environmental radioactivity

28 CFR 301.204 - Continuation of lost-time wages.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Continuation of lost-time wages. 301.204... ACCIDENT COMPENSATION Lost-Time Wages § 301.204 Continuation of lost-time wages. (a) Once approved, the inmate shall receive lost-time wages until the inmate: (1) Is released; (2) Is transferred to another...

Carbon fiber counting. [aircraft structures

NASA Technical Reports Server (NTRS)

Pride, R. A.

1980-01-01

A method was developed for characterizing the number and lengths of carbon fibers accidentally released by the burning of composite portions of civil aircraft structure in a jet fuel fire after an accident. Representative samplings of carbon fibers collected on transparent sticky film were counted from photographic enlargements with a computer aided technique which also provided fiber lengths.

Novel method for estimation of the indoor-to-outdoor airborne radioactivity ratio following the Fukushima Daiichi Nuclear Power Plant accident.

PubMed

Tan, Yanliang; Ishikawa, Tetsuo; Janik, Miroslaw; Tokonami, Shinji; Hosoda, Masahiro; Sorimachi, Atsuyuki; Kearfott, Kimberlee

2015-12-01

The accident at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in Japan resulted in significant releases of fission products. While substantial data exist concerning outdoor air radioactivity following the accident, the resulting indoor radioactivity remains pure speculation without a proper method for estimating the ratio of the indoor to outdoor airborne radioactivity, termed the airborne sheltering factor (ASF). Lacking a meaningful value of the ASF, it is difficult to assess the inhalation doses to residents and evacuees even when outdoor radionuclide concentrations are available. A simple model was developed and the key parameters needed to estimate the ASF were obtained through data fitting of selected indoor and outdoor airborne radioactivity measurement data obtained following the accident at a single location. Using the new model with values of the air exchange rate, interior air volume, and the inner surface area of the dwellings, the ASF can be estimated for a variety of dwelling types. Assessment of the inhalation dose to individuals readily follows from the value of the ASF, the person's indoor occupancy factor, and the measured outdoor radioactivity concentration. Copyright © 2015 Elsevier B.V. All rights reserved.

Fukushima Daiichi Nuclear Power Plant accident: facts, environmental contamination, possible biological effects, and countermeasures

PubMed Central

Anzai, Kazunori; Ban, Nobuhiko; Ozawa, Toshihiko; Tokonami, Shinji

2012-01-01

On March 11, 2011, an earthquake led to major problems at the Fukushima Daiichi Nuclear Power Plant. A 14-m high tsunami triggered by the earthquake disabled all AC power to Units 1, 2, and 3 of the Power Plant, and carried off fuel tanks for emergency diesel generators. Despite many efforts, cooling systems did not work and hydrogen explosions damaged the facilities, releasing a large amount of radioactive material into the environment. In this review, we describe the environmental impact of the nuclear accident, and the fundamental biological effects, acute and late, of the radiation. Possible medical countermeasures to radiation exposure are also discussed. PMID:22247595

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schwantes, Jon M.

Kelly Fitzgerald Kelly Fitzgerald assisted with laboratory testing for an ongoing R&D project known as Electrochemically Modulated Separation (EMS) for on-line rapid preseparations of actinides prior to mass spectrometry analysis. Ryne Burgess Ryne Burgess used SCALE 5.1 ORIGEN-ARP to predict isotope libraries for the Units 1, 2 and 3 reactors and Unit 4 spent fuel pool for comparing against measurements of environmental sampled collected at the site in order to identify the source terms of the accident. Comparison of the cesium 134/137 and cesium 136/137 ratios observed in environmental samples and ORIGEN-ARP predictions indicated that the Unit 4 Spent Fuelmore » Pool did not significantly contribute to radionuclide release during the Fukushima Daiichi accident.« less

Station blackout calculations for Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.; Weber, C.F.; Hyman, C.R.

1985-01-01

This paper presents the results of calculations performed with the ORNL SASA code suite for the Station Blackout Severe Accident Sequence at Browns Ferry. The accident is initiated by a loss of offsite power combined with failure of all onsite emergency diesel generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion (at six hours) and without DC power, cooling water could no longer be injected into the reactor vessel. Calculations are continued through the period of core degradation and melting, reactor vessel failure, and the subsequent containment failure. An estimate ofmore » the magnitude and timing of the concomitant fission product releases is also provided.« less

Inverse modelling for real-time estimation of radiological consequences in the early stage of an accidental radioactivity release.

PubMed

Pecha, Petr; Šmídl, Václav

2016-11-01

A stepwise sequential assimilation algorithm is proposed based on an optimisation approach for recursive parameter estimation and tracking of radioactive plume propagation in the early stage of a radiation accident. Predictions of the radiological situation in each time step of the plume propagation are driven by an existing short-term meteorological forecast and the assimilation procedure manipulates the model parameters to match the observations incoming concurrently from the terrain. Mathematically, the task is a typical ill-posed inverse problem of estimating the parameters of the release. The proposed method is designated as a stepwise re-estimation of the source term release dynamics and an improvement of several input model parameters. It results in a more precise determination of the adversely affected areas in the terrain. The nonlinear least-squares regression methodology is applied for estimation of the unknowns. The fast and adequately accurate segmented Gaussian plume model (SGPM) is used in the first stage of direct (forward) modelling. The subsequent inverse procedure infers (re-estimates) the values of important model parameters from the actual observations. Accuracy and sensitivity of the proposed method for real-time forecasting of the accident propagation is studied. First, a twin experiment generating noiseless simulated "artificial" observations is studied to verify the minimisation algorithm. Second, the impact of the measurement noise on the re-estimated source release rate is examined. In addition, the presented method can be used as a proposal for more advanced statistical techniques using, e.g., importance sampling. Copyright © 2016 Elsevier Ltd. All rights reserved.

Explosion impacts during transport of hazardous cargo: GIS-based characterization of overpressure impacts and delineation of flammable zones for ammonia.

PubMed

Inanloo, Bahareh; Tansel, Berrin

2015-06-01

The aim of this research was to investigate accidental releases of ammonia followed by an en-route incident in an attempt to further predict the consequences of hazardous cargo accidents. The air dispersion model Areal Locations of Hazardous Atmospheres (ALOHA) was employed to track the probable outcomes of a hazardous material release of a tanker truck under different explosion scenarios. The significance of identification of the flammable zones was taken into consideration; in case the flammable vapor causes an explosion. The impacted areas and the severity of the probable destructions were evaluated for an explosion by considering the overpressure waves. ALOHA in conjunction with ArcGIS was used to delineate the flammable and overpressure impact zones for different scenarios. Based on the results, flammable fumes were formed in oval shapes having a chief axis along the wind direction at the time of release. The expansions of the impact areas under the overpressure value which can lead to property damage for 2 and 20 tons releases, under very stable and unstable atmospheric conditions were estimated to be around 1708, 1206; 3742, 3527 feet, respectively, toward the wind direction. A sensitivity analysis was done to assess the significance of wind speed on the impact zones. The insight provided by this study can be utilized by decision makers in transportation of hazardous materials as a guide for possible rerouting, rescheduling, or limiting the quantity of hazardous cargo to reduce the possible impacts after hazardous cargo accidents during transport. Copyright © 2015 Elsevier Ltd. All rights reserved.

Dispersion of Fukushima radionuclides in the global atmosphere and the ocean.

PubMed

Povinec, P P; Gera, M; Holý, K; Hirose, K; Lujaniené, G; Nakano, M; Plastino, W; Sýkora, I; Bartok, J; Gažák, M

2013-11-01

Large quantities of radionuclides were released in March-April 2011 during the accident of the Fukushima Dai-ichi Nuclear Power Plant to the atmosphere and the ocean. Atmospheric and marine modeling has been carried out to predict the dispersion of radionuclides worldwide, to compare the predicted and measured radionuclide concentrations, and to assess the impact of the accident on the environment. Atmospheric Lagrangian dispersion modeling was used to simulate the dispersion of (137)Cs over America and Europe. Global ocean circulation model was applied to predict the dispersion of (137)Cs in the Pacific Ocean. The measured and simulated (137)Cs concentrations in atmospheric aerosols and in seawater are compared with global fallout and the Chernobyl accident, which represent the main sources of the pre-Fukushima radionuclide background in the environment. The radionuclide concentrations in the atmosphere have been negligible when compared with the Chernobyl levels. The maximum (137)Cs concentration in surface waters of the open Pacific Ocean will be around 20 Bq/m(3). The plume will reach the US coast 4-5 y after the accident, however, the levels will be below 3 Bq/m(3). All the North Pacific Ocean will be labeled with Fukushima (137)Cs 10 y after the accident with concentration bellow 1 Bq/m(3). Copyright © 2013 Elsevier Ltd. All rights reserved.

"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. © 2012 Society for Risk Analysis.

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 bymore » 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.« less

Industrial Accidents Triggered by Natural Hazards: an Emerging Risk Issue

NASA Astrophysics Data System (ADS)

Renni, Elisabetta; Krausmann, Elisabeth; Basco, Anna; Salzano, Ernesto; Cozzani, Valerio

2010-05-01

Natural disasters such as earthquakes, tsunamis, flooding or hurricanes have recently and dramatically hit several countries worldwide. Both direct and indirect consequences involved the population, causing on the one hand a high number of fatalities and on the other hand so relevant economical losses that the national gross product may be affected for many years. Loss of critical industrial infrastructures (electricity generation and distribution, gas pipelines, oil refineries, etc.) also occurred, causing further indirect damage to the population. In several cases, accident scenarios with large releases of hazardous materials were triggered by these natural events, causing so-called "Natech events", in which the overall damage resulted from the simultaneous consequences of the natural event and of the release of hazardous substances. Toxic releases, large fires and explosions, as well as possible long-term environmental pollution, economical losses, and overloading of emergency systems were recognised by post-event studies as the main issues of these Natech scenarios. In recent years the increasing frequency and severity of some natural hazards due to climate change has slowly increased the awareness of Natech risk as an emerging risk among the stakeholders. Indeed, the iNTeg-Risk project, co-funded by the European Commission within the 7th Framework Program specifically addresses these scenarios among new technological issues on public safety. The present study, in part carried out within the iNTeg-Risk project, was aimed at the analysis and further development of methods and tools for the assessment and mitigation of Natech accidents. Available tools and knowledge gaps in the assessment of Natech scenarios were highlighted. The analysis mainly addressed the potential impact of flood, lightning and earthquake events on industrial installations where hazardous substances are present. Preliminary screening methodologies and more detailed methods based on quantitative risk analysis were developed. Strategies based on the use of multiple information layers aiming at the identification of mitigation and early warning systems were also explored. A case-study in the Emilia-Romagna region is presented.

Recent plant studies using Victoria 2.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

BIXLER,NATHAN E.; GASSER,RONALD D.

2000-03-08

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised andmore » expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment.« less

MK-801-induced behavioural sensitisation alters dopamine release and turnover in rat prefrontal cortex.

PubMed

Cui, Xiaoying; Lefevre, Emilia; Turner, Karly M; Coelho, Carlos M; Alexander, Suzy; Burne, Thomas H J; Eyles, Darryl W

2015-02-01

Repeated exposure to psychostimulants that either increase dopamine (DA) release or target N-methyl-D-aspartate (NMDA) receptors can induce behavioural sensitisation, a phenomenon that may be important for the processes of addiction and even psychosis. A critical component of behavioural sensitisation is an increase in DA release within mesocorticolimbic circuits. In particular, sensitisation to amphetamine leads to increased DA release within well-known sub-cortical brain regions and also regulatory regions such as prefrontal cortex (PFC). However, it is unknown how DA release within the PFC of animals is altered by sensitisation to NMDA receptor antagonists. The aims of the present study were twofold, firstly to examine whether a single dose of dizocilpine maleate (MK-801) could induce long-term behavioural sensitisation and secondly to examine DA release in the PFC of sensitised rats. Behavioural sensitisation was assessed by measuring locomotion after drug exposure. DA release in the PFC was measured using freely moving microdialysis. We show that a single dose of MK-801 can induce sensitisation to subsequent MK-801 exposure in a high percentage of rats (66 %). Furthermore, rats sensitised to MK-801 have altered DA release and turnover in the PFC compared with non-sensitised rats. Schizophrenia patients have been postulated to have 'endogenous sensitisation' to psychostimulants. MK-801-induced sensitised rats, in particular when compared with non-sensitised rats, provide a useful model for studying PFC dysfunction in schizophrenia.

Aerosol Sample Inhomogeneity with Debris from the Fukushima Daiichi Nuclear Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gomez, Reynaido; Biegalski, Steven R.; Woods, Vincent T.

2014-09-01

Radionuclide aerosol sampling is a vital component in the detection of nuclear explosions, nuclear accidents, and other radiation releases. This was proven by the detection and tracking of emissions from the Fukushima Daiichi incident across the globe by IMS stations. Two separate aerosol samplers were operated in Richland, WA following the event and debris from the accident were measured at levels well above detection limits. While the atmospheric activity concentration of radionuclides generally compared well between the two stations, they did not agree within uncertainties. This paper includes a detailed study of the aerosol sample homogeneity of 134Cs and 137Cs,more » then relates it to the overall uncertainty of the original measurement. Our results show that sample inhomogeneity adds an additional 5–10% uncertainty to each aerosol measurement and that this uncertainty is in the same range as the discrepancies between the two aerosol sample measurements from Richland, WA.« less

Southward spreading of the Fukushima-derived radiocesium across the Kuroshio Extension in the North Pacific

PubMed Central

Kumamoto, Yuichiro; Aoyama, Michio; Hamajima, Yasunori; Aono, Tatsuo; Kouketsu, Shinya; Murata, Akihiko; Kawano, Takeshi

2014-01-01

The accident of the Fukushima Dai-ichi nuclear power plant in March 2011 released a large amount of radiocesium into the North Pacific Ocean. Vertical distributions of Fukushima-derived radiocesium were measured at stations along the 149°E meridian in the western North Pacific during the winter of 2012. In the subtropical region, to the south of the Kuroshio Extension, we found a subsurface radiocesium maximum at a depth of about 300 m. It is concluded that atmospheric-deposited radiocesium south of the Kuroshio Extension just after the accident had been transported not only eastward along with surface currents but also southward due to formation/subduction of subtropical mode waters within about 10 months after the accident. The total amount of decay-corrected 134Cs in the mode water was an estimated about 6 PBq corresponding to 10–60% of the total inventory of Fukushima-derived 134Cs in the North Pacific Ocean. PMID:24589762

Thyroid doses for evacuees from the Fukushima nuclear accident.

PubMed

Tokonami, Shinji; Hosoda, Masahiro; Akiba, Suminori; Sorimachi, Atsuyuki; Kashiwakura, Ikuo; Balonov, Mikhail

2012-01-01

A primary health concern among residents and evacuees in affected areas immediately after a nuclear accident is the internal exposure of the thyroid to radioiodine, particularly I-131, and subsequent thyroid cancer risk. In Japan, the natural disasters of the earthquake and tsunami in March 2011 destroyed an important function of the Fukushima Daiichi Nuclear Power Plant (F1-NPP) and a large amount of radioactive material was released to the environment. Here we report for the first time extensive measurements of the exposure to I-131 revealing I-131 activity in the thyroid of 46 out of the 62 residents and evacuees measured. The median thyroid equivalent dose was estimated to be 4.2 mSv and 3.5 mSv for children and adults, respectively, much smaller than the mean thyroid dose in the Chernobyl accident (490 mSv in evacuees). Maximum thyroid doses for children and adults were 23 mSv and 33 mSv, respectively.

The Chornobyl accident revisited, part II: The state of the nuclear fuel located within the Chornobyl Sarcophagus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borovol, A.A.; Sich, A.R.

1995-10-01

Approximately 135 tonnes of the 190.3-tonne initial core fuel load ({approx}71%) at Chornobyl Unit 4 melted and flowed into the lower regions of the reactor building to form various kinds of the now-solidified lava-like fuel-containing materials (LFCMs) or corium. The results of radiochemical analyses reveal that only 5% of the LFCM inventory of Ru-106 remains, whereas, surprisingly, 35% of the LFCM inventory of Cs-137 remains. Moreover, the results of these analyses support the fact that little if any of the 5020 tonnes of various materials (dropped from helicopters during the active phase of the accident in an attempt to smothermore » the burning graphite) ever made it into the core shaft, where the bulk of the core was located. The results appear to support earlier Western source-term estimates that significantly more volatile radionuclides may have been released as a result of the accident.« less

The Chornobyl accident revisited, Part II: The state of the nuclear fuel located within the Chornobyl sarcophagus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borovoi, A.A.; Sich, A.R.

1995-01-01

Approximately 135 tonnes of the 190.3-tonne initial core fuel load ({approx}71%) at Chernobyl Unit 4 melted and flowed into the lower regions of the reactor building to form various kinds of the now-solidified lava-like fuel-containing materials (LFCMs) or corium. The results of radiochemical analyses reveal that only 5% of the LFCM inventory of Ru-106 remains. whereas, surprisingly, 35% of the LFCM inventory of Cs-137 remains. Moreover, the results of these analyses support the fact that little if any of the 5020 tonnes of various materials (dropped from helicopters during the active phase of the accident in an attempt to smothermore » the burning graphite) ever made it into the core shaft, where the bulk of the core was located. The results appear to support earlier Western source-term estimates that significantly more volatile radionuclides may have been released as a result of the accident. 37 refs., 13 figs., 7 tabs.« less

ESTIMATION OF EFFECTIVE DOSE FROM EXTERNAL EXPOSURE DUE TO SHORT-LIVED NUCLIDES IN THE PREFECTURES SURROUNDING FUKUSHIMA.

PubMed

Miyatake, Hirokazu; Yoshizawa, Nobuaki; Suzuki, Gen

2018-05-11

The Fukushima Daiichi Nuclear Power Plant (FDNPP) accident resulted in a release of radionuclides into the environment. Since the accident, measurements of radiation in the environment such as air dose rate and deposition density of radionuclides have been performed by various organizations and universities. In particular, Japan Atomic Energy Agency (JAEA) has been performing observations of air dose rate using a car-borne survey system continuously over widespread areas. Based on the data measured by JAEA, we estimated effective dose from external exposure in the prefectures surrounding Fukushima. Since car-borne survey started a few months after the accident, the main contribution to measured data comes from 137Cs and 134Cs whose half-lives are relatively long. Using air dose rate of 137Cs and 134Cs and the ratio of deposition density of short-lived nuclides to that of 137Cs and 134Cs, we also estimated contributions to the effective dose from other short-lived nuclides.

The Fukushima Daiichi Nuclear Power Plant accident and school bullying of affected children and adolescents: the need for continuous radiation education.

PubMed

Sawano, Toyoaki; Nishikawa, Yoshitaka; Ozaki, Akihiko; Leppold, Claire; Tsubokura, Masaharu

2018-05-01

The health threats of radiation-release incidents are diverse and long term. In addition to direct radiation effects, it is imperative to manage the indirect effects of radiation such as stigma, prejudice and broader mental health impacts. Six years after the Fukushima Daiichi Nuclear Power Plant accident of March 2011, bullying caused by stigma and prejudice toward evacuees, including children, has become a social problem in Japan. This phenomenon may be associated with the fact that knowledge about radiation has still not reached the general public, and to a potential lack of motivation among Japanese citizens to learn about radiation and bullying. Continuous and sustained education regarding radiation is warranted in order to enhance the general knowledge level about the effects of radiation in Japan after the Fukushima Daiichi Nuclear Power Plant accident, and this education will become an important reference for education after future nuclear disasters.

Thyroid doses for evacuees from the Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Tokonami, Shinji; Hosoda, Masahiro; Akiba, Suminori; Sorimachi, Atsuyuki; Kashiwakura, Ikuo; Balonov, Mikhail

2012-07-01

A primary health concern among residents and evacuees in affected areas immediately after a nuclear accident is the internal exposure of the thyroid to radioiodine, particularly I-131, and subsequent thyroid cancer risk. In Japan, the natural disasters of the earthquake and tsunami in March 2011 destroyed an important function of the Fukushima Daiichi Nuclear Power Plant (F1-NPP) and a large amount of radioactive material was released to the environment. Here we report for the first time extensive measurements of the exposure to I-131 revealing I-131 activity in the thyroid of 46 out of the 62 residents and evacuees measured. The median thyroid equivalent dose was estimated to be 4.2 mSv and 3.5 mSv for children and adults, respectively, much smaller than the mean thyroid dose in the Chernobyl accident (490 mSv in evacuees). Maximum thyroid doses for children and adults were 23 mSv and 33 mSv, respectively.

Experiences in methods to involve key players in planning protective actions in the case of a nuclear accident.

PubMed

Sinkko, K; Hämäläinen, R P; Hänninen, R

2004-01-01

A widely used method in the planning of protective actions is to establish a stakeholder network to generate a comprehensive set of generic protective actions. The aim is to increase competence and build links for communication and coordination. The approach of this work was to systematically evaluate protective action strategies in the case of a nuclear accident. This was done in a way that the concerns and issues of all key players could be transparently and equally included in the decision taken. An approach called Facilitated Decision Analysis Workshop has been developed and tested. The work builds on case studies in which it was assumed that a hypothetical accident had led to a release of considerable amounts of radionuclides and, therefore, various types of countermeasures had to be considered. Six workshops were organised in the Nordic countries where the key players were represented, i.e. authorities, expert organisations, industry and agricultural producers. Copyright 2004 Oxford University Press

The biological impacts of the Fukushima nuclear accident on the pale grass blue butterfly

PubMed Central

Hiyama, Atsuki; Nohara, Chiyo; Kinjo, Seira; Taira, Wataru; Gima, Shinichi; Tanahara, Akira; Otaki, Joji M.

2012-01-01

The collapse of the Fukushima Dai-ichi Nuclear Power Plant caused a massive release of radioactive materials to the environment. A prompt and reliable system for evaluating the biological impacts of this accident on animals has not been available. Here we show that the accident caused physiological and genetic damage to the pale grass blue Zizeeria maha, a common lycaenid butterfly in Japan. We collected the first-voltine adults in the Fukushima area in May 2011, some of which showed relatively mild abnormalities. The F1 offspring from the first-voltine females showed more severe abnormalities, which were inherited by the F2 generation. Adult butterflies collected in September 2011 showed more severe abnormalities than those collected in May. Similar abnormalities were experimentally reproduced in individuals from a non-contaminated area by external and internal low-dose exposures. We conclude that artificial radionuclides from the Fukushima Nuclear Power Plant caused physiological and genetic damage to this species. PMID:22880161

Operational atmospheric modeling system CARIS for effective emergency response associated with hazardous chemical releases in Korea.

PubMed

Kim, Cheol-Hee; Park, Jin-Ho; Park, Cheol-Jin; Na, Jin-Gyun

2004-03-01

The Chemical Accidents Response Information System (CARIS) was developed at the Center for Chemical Safety Management in South Korea in order to track and predict the dispersion of hazardous chemicals in the case of an accident or terrorist attack involving chemical companies. The main objective of CARIS is to facilitate an efficient emergency response to hazardous chemical accidents by rapidly providing key information in the decision-making process. In particular, the atmospheric modeling system implemented in CARIS, which is composed of a real-time numerical weather forecasting model and an air pollution dispersion model, can be used as a tool to forecast concentrations and to provide a wide range of assessments associated with various hazardous chemicals in real time. This article introduces the components of CARIS and describes its operational modeling system. Some examples of the operational modeling system and its use for emergency preparedness are presented and discussed. Finally, this article evaluates the current numerical weather prediction model for Korea.

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.

The biological impacts of the Fukushima nuclear accident on the pale grass blue butterfly.

PubMed

Hiyama, Atsuki; Nohara, Chiyo; Kinjo, Seira; Taira, Wataru; Gima, Shinichi; Tanahara, Akira; Otaki, Joji M

2012-01-01

The collapse of the Fukushima Dai-ichi Nuclear Power Plant caused a massive release of radioactive materials to the environment. A prompt and reliable system for evaluating the biological impacts of this accident on animals has not been available. Here we show that the accident caused physiological and genetic damage to the pale grass blue Zizeeria maha, a common lycaenid butterfly in Japan. We collected the first-voltine adults in the Fukushima area in May 2011, some of which showed relatively mild abnormalities. The F₁ offspring from the first-voltine females showed more severe abnormalities, which were inherited by the F₂ generation. Adult butterflies collected in September 2011 showed more severe abnormalities than those collected in May. Similar abnormalities were experimentally reproduced in individuals from a non-contaminated area by external and internal low-dose exposures. We conclude that artificial radionuclides from the Fukushima Nuclear Power Plant caused physiological and genetic damage to this species.

The Fukushima Daiichi Nuclear Power Plant accident and school bullying of affected children and adolescents: the need for continuous radiation education

PubMed Central

Sawano, Toyoaki; Nishikawa, Yoshitaka; Ozaki, Akihiko; Leppold, Claire; Tsubokura, Masaharu

2018-01-01

Abstract The health threats of radiation-release incidents are diverse and long term. In addition to direct radiation effects, it is imperative to manage the indirect effects of radiation such as stigma, prejudice and broader mental health impacts. Six years after the Fukushima Daiichi Nuclear Power Plant accident of March 2011, bullying caused by stigma and prejudice toward evacuees, including children, has become a social problem in Japan. This phenomenon may be associated with the fact that knowledge about radiation has still not reached the general public, and to a potential lack of motivation among Japanese citizens to learn about radiation and bullying. Continuous and sustained education regarding radiation is warranted in order to enhance the general knowledge level about the effects of radiation in Japan after the Fukushima Daiichi Nuclear Power Plant accident, and this education will become an important reference for education after future nuclear disasters. PMID:29635385

Control of fixed-wing UAV at levelling phase using artificial intelligence

NASA Astrophysics Data System (ADS)

Sayfeddine, Daher

2018-03-01

The increase in the share of fly-by-wire and software controlled UAV is explained by the need to release the human-operator and the desire to reduce the degree of influence of the human factor errors that account for 26% of aircraft accidents. An important reason for the introduction of new control algorithms is also the high level of UAV failures due loss of communication channels and possible hacking. This accounts for 17% of the total number of accidents. The comparison with manned flights shows that the frequency of accidents of unmanned flights is 27,000 times higher. This means that the UAV has 1611 failures per million flight hours and only 0.06 failures at the same time for the manned flight. In view of that, this paper studies the flight autonomy of fixed-wing UAV at the levelling phase. Landing parameters of the UAV are described. They will be used to setup a control scheme for an autopilot based on fuzzy logic algorithm.

Large Scale Gas Mixing and Stratification Triggered by a Buoyant Plume With and Without Occurrence of Condensation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, Domenico; Auban, Olivier; Zboray, Robert

The benefits of using codes with 3-D capabilities to address safety issues of LWRs will be applicable to both the current generation of nuclear reactors as well to future ALWRs. The phenomena governing the containment response in case of some postulated severe accident scenarios include gas (air, hydrogen, steam) stratification in the containment, gas distribution between containment compartments, wall condensation, etc. These phenomena are driven by buoyant high momentum injection (jets) and/or low momentum injection (plumes). For instance, mixing in the immediate vicinity of the postulated line break is mainly dominated by very high velocity efflux, while low-momentum flows aremore » responsible for most of the transport processes within the containment. A project named SETH is currently in progress under the auspices of 15 OECD countries, with the aim of creating an experimental database suitable to assess the 3-D code capabilities in analyzing key-physical phenomena relevant for LWR safety analysis. This paper describes some results of two SETH tests, performed in the PANDA facility (located at PSI in Switzerland), focusing on plumes flowing near a containment wall. The plumes are generated by injecting a constant amount of steam in one of two interconnected vessels initially filled with air. In one of the two tests the temperature of the injected steam and the initial containment wall and fluid temperatures allowed for condensation during the test. (authors)« less

[Comparative analysis of the radionuclide composition in fallout after the Chernobyl and the Fukushima accidents].

PubMed

Kotenko, K V; Shinkarev, S M; Abramov, Iu V; Granovskaia, E O; Iatsenko, V N; Gavrilin, Iu I; Margulis, U Ia; Garetskaia, O S; Imanaka, T; Khoshi, M

2012-01-01

The nuclear accident occurred at Fukushima Dai-ichi Nuclear Power Plant (NPP) (March 11, 2011) similarly to the accident at the Chernobyl NPP (April 26, 1986) is related to the level 7 of the INES. It is of interest to make an analysis of the radionuclide composition of the fallout following the both accidents. The results of the spectrometric measurements were used in that comparative analysis. Two areas following the Chernobyl accident were considered: (1) the near zone of the fallout - the Belarusian part of the central spot extended up to 60 km around the Chernobyl NPS and (2) the far zone of the fallout--the "Gomel-Mogilev" spot centered 200 km to the north-northeast of the damaged reactor. In the case of Fukushima accident the near zone up to about 60 km considered. The comparative analysis has been done with respect to refractory radionuclides (95Zr, 95Nb, 141Ce, 144Ce), as well as to the intermediate and volatile radionuclides 103Ru, 106Ru, 131I, 134Cs, 137Cs, 140La, 140Ba and the results of such a comparison have been discussed. With respect to exposure to the public the most important radionuclides are 131I and 137Cs. For the both accidents the ratios of 131I/137Cs in the considered soil samples are in the similar ranges: (3-50) for the Chernobyl samples and (5-70) for the Fukushima samples. Similarly to the Chernobyl accident a clear tendency that the ratio of 131I/137Cs in the fallout decreases with the increase of the ground deposition density of 137Cs within the trace related to a radioactive cloud has been identified for the Fukushima accident. It looks like this is a universal tendency for the ratio of 131I/137Cs versus the 137Cs ground deposition density in the fallout along the trace of a radioactive cloud as a result of a heavy accident at the NPP with radionuclides releases into the environment. This tendency is important for an objective reconstruction of 131I fallout based on the results of 137Cs measurements of soil samples carried out at late dates after the Fukushima accident.

Mortality among the residents of the Three Mile Island accident area: 1979-1992.

PubMed

Talbott, E O; Youk, A O; McHugh, K P; Shire, J D; Zhang, A; Murphy, B P; Engberg, R A

2000-06-01

The largest U.S. population exposed to low-level radioactivity released by an accident at a nuclear power plant is composed of residents near the Three Mile Island (TMI) Plant on 28 March 1979. This paper (a collaboration of The University of Pittsburgh and the Pennsylvania Department of Health) reports on the mortality experience of the 32,135 members in this cohort for 1979-1992. We analyzed standardized mortality ratios (SMRs) using a local comparison population and performed relative risk regression modeling to assess overall mortality and specific cancer risks by confounding factors and radiation-related exposure variables. Total mortality was significantly elevated for both men and women (SMRs = 109 and 118, respectively). All heart disease accounted for 43.3% of total deaths and demonstrated elevated SMRs for heart disease of 113 and 130 for men and women, respectively; however, when controlling for confounders and natural background radiation, these elevations in heart disease were no longer evident. Overall cancer mortality was similar in this cohort as compared to the local population (male SMR = 100; female SMR = 101). In the relative risk modeling, there was a significant effect for all lymphatic and hematopoietic tissue in males in relation to natural background exposure (p = 0.04). However, no trend was noted. We found a significant linear trend for female breast cancer risk in relation to increasing levels of TMI-related likely [gamma]-exposure (p = 0.02). Although such a relationship has been noted in other investigations, emissions from the TMI incident were significantly lower than in other documented studies. Therefore, it is unlikely that this observed increase is related to radiation exposure on the day of the accident. The mortality surveillance of this cohort does not provide consistent evidence that radioactivity released during the TMI accident has a significant impact on the mortality experience of this cohort to date. However, continued follow-up of these individuals will provide a more comprehensive description of the morbidity and mortality experience of the cohort.

Durability and degradation of HT9 based alloy waste forms with variable Ni and Cr content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, L.

2016-12-31

Short-term electrochemical and long-term hybrid electrochemical corrosion tests were performed on alloy waste forms in reference aqueous solutions that bound postulated repository conditions. The alloy waste forms investigated represent candidate formulations that can be produced with advanced electrochemical treatment of used nuclear fuel. The studies helped to better understand the alloy waste form durability with differing concentrations of nickel and chromium, species that can be added to alloy waste forms to potentially increase their durability and decrease radionuclide release into the environment.

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, Douglas M.; Nesbitt, Loyd B.

1997-01-01

A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through cladmore » melting at 1370/sup 0/C.« less

Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kervinen, T.; Riikonen, V.; Ritonummi, T.

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing themore » heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.« less

ATWS analysis for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R.J.; Jouse, W.C.

1985-01-01

Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of themore » primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4F/min.« less

ASK Talks with Al Diaz

NASA Technical Reports Server (NTRS)

Diaz, Alphonso; OKeefe, Sean

2004-01-01

Following the release of the Columbia Accident Investigation Board (CAIB) Report, Alphonso (Al) Diaz, Goddard Center director, was asked by NASA Administrator Sean O'Keefe to head up the Agency's response. The Diaz team, as it came to be known, was changed with making sure the CAIB Report did not become another dusty volume on a shelf of old Agency reviews.

Cesium-137 Fallout in Indiana Soil

ERIC Educational Resources Information Center

Whitman, Richard T.

2017-01-01

Atomic weapons testing during the Cold War and accidents at nuclear power plants have resulted in the release of radioactive fallout over great distances. Little is known about levels of fallout deposited in Indiana. The reported study sampled soil in all 92 Indiana counties to determine the present level of cesium-137 from the 2 to 12 centimeter…

A novel method for the determination of sildenafil (Viagra(R)) and its metabolite (UK-103,320) in postmortem specimens using LC/MS/MS and LC/MS/MS/MS : final report.

DOT National Transportation Integrated Search

2000-05-01

During the investigation of aviation accidents, postmortem samples from victims are submitted to the FAAs Civil Aeromedical Institute for drug analysis. Because new drugs are continually being released to the market, it is our laboratorys respo...

RADIOACTIVE CONTAMINATION OF FOODS. PROBLEMS IN THE FOOD CONSUMPTION OF THE ITALIAN POPULATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferro-Luzzi, A.; Mariani, A.

The aspects of health physics that are basically applications of physics are reviewed. Units of radiation measurement, RBE, permissible doses, personnel monitoring, applications of radiation spectrometry, and measurement of body activity are considered, as well as the release, dispersion, and deposition of radioactive material in reactor accidents. 140 references. (D.C.W.)

75 FR 922 - Notification and Reporting of Aircraft Accidents or Incidents and Overdue Aircraft, and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-07

... pose to crews, passengers, and bystanders. However, the NTSB notes that propeller blades are designed... intact and in place during normal operation. Propeller blades are not designed or expected to continue to... release of all or a portion of a propeller blade from an aircraft, inconsistent with its design parameters...

Variability in platelet dense granule adenosine triphosphate release findings amongst patients tested multiple times as part of an assessment for a bleeding disorder.

PubMed

Badin, M S; Graf, L; Iyer, J K; Moffat, K A; Seecharan, J L; Hayward, C P M

2016-12-01

Lumi-aggregometry quantification of platelet dense granule adenosine triphosphate (ATP) release is commonly used for diagnosing platelet function disorders. As the test findings show considerable variability for healthy controls, we postulated that patient findings might also be variable and investigated patients who were assessed for dense granule ATP release defects more than once. Analyses were performed on prospectively collected data for first and second tests for subjects tested for dense granule ATP release defects more than once by the Hamilton Regional Laboratory Program (HRLMP) between January 2007 and June 2013 (cohort I). Similar analyses were performed for subjects who were recruited to a platelet disorder study (cohort II) and were assessed for ATP release defects more than once before October 2015. A total of 150 unique subjects had multiple ATP release tests. Results with individual agonists were variable for many subjects. While normal findings with all tested agonists were often confirmed by the second test (cohort I: 83%; cohort II: 100%), impaired release with multiple agonists was confirmed in only some subjects (cohort I: 34%; cohort II: 54%). Inconsistent findings were common (cohort I: 36%; cohort II: 39%). ISTH bleeding scores showed no relationship to the test findings. The finding of impaired ATP release with 2 or more agonists on both tests was not associated with an increased likelihood of a definite bleeding disorder. The variability in platelet dense granule ATP release findings amongst patients assessed for diagnostic purposes suggests that the test has limited value for diagnosing platelet disorders. © 2016 John Wiley & Sons Ltd.

Consequences of severe nuclear accidents in Europe

NASA Astrophysics Data System (ADS)

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

2013-04-01

A first part of the presentation is devoted to the consequences of the severe accident in the 1986 Chernobyl NPP. It lead to a substantial radioactive contaminated of large parts of Europe and thus raised the awareness for off-site nuclear accident consequences. Spatial patterns of the (transient) contamination of the air and (persistent) contamination of the ground were studied by both measurements and model simulations. For a variety of reasons, ground contamination measurements have variability at a range of spatial scales. Results will be reviewed and discussed. Model simulations, including inverse modelling, have shown that the standard source term as defined in the ATMES study (1990) needs to be updated. Sensitive measurements of airborne activities still reveal the presence of low levels of airborne radiocaesium over the northern hemisphere which stems from resuspension. Over time scales of months and years, the distribution of radionuclides in the Earth system is constantly changing, for example relocated within plants, between plants and soil, in the soil, and into water bodies. Motivated by the permanent risk of transboundary impacts from potential major nuclear accidents, the multidisciplinary project flexRISK (see http://flexRISK.boku.ac.at) has been carried out from 2009 to 2012 in Austria to quantify such risks and hazards. An overview of methods and results of flexRISK is given as a second part of the presentation. For each of the 228 NPPs, severe accidents were identified together with relevant inventories, release fractions, and release frequencies. Then, Europe-wide dispersion and dose calculations were performed for 2788 cases, using the Lagrangian particle model FLEXPART. Maps of single-case results as well as various aggregated risk parameters were produced. It was found that substantial consequences (intervention measures) are possible for distances up to 500-1000 km, and occur more frequently for a distance range up to 100-300 km, which is in agreement with Chernobyl experiences. However, emergency planning presently is still often focussing on too small areas. In reality, almost all of Europe should be prepared for nuclear disaster. The project investigated also the effect of a simple phase-out scenario. A regional phase-out policy is effective for reducing or even eliminating high damage in the respective regions. It should also be mentioned that risk distribution depends strongly on accident frequency, but this parameter is highly uncertain. The work in flexRISK was funded by the Austrian Climate and Energy Fund (KLI.EN).

Status of VICTORIA: NRC peer review and recent code applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Schaperow, J.H.

1997-12-01

VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A summary of the results and recommendations of an independent peer review of VICTORIA by the US Nuclear Regulatory Commission (NRC) is presented, along with recent applications of the code. The latter include analyses of a temperature-induced steam generator tube rupture sequence and post-test analyses of the Phebus FPT-1 test. Themore » next planned Phebus test, FTP-4, will focus on fission product releases from a rubble bed, especially those of the less-volatile elements, and on the speciation of the released elements. Pretest analyses using VICTORIA to estimate the magnitude and timing of releases are presented. The predicted release of uranium is a matter of particular importance because of concern about filter plugging during the test.« less

Data summary report for fission product release test VI-5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected inmore » three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.« less

Source term estimation of radioxenon released from the Fukushima Dai-ichi nuclear reactors using measured air concentrations and atmospheric transport modeling.

PubMed

Eslinger, P W; Biegalski, S R; Bowyer, T W; Cooper, M W; Haas, D A; Hayes, J C; Hoffman, I; Korpach, E; Yi, J; Miley, H S; Rishel, J P; Ungar, K; White, B; Woods, V T

2014-01-01

Systems designed to monitor airborne radionuclides released from underground nuclear explosions detected radioactive fallout across the northern hemisphere resulting from the Fukushima Dai-ichi Nuclear Power Plant accident in March 2011. Sampling data from multiple International Modeling System locations are combined with atmospheric transport modeling to estimate the magnitude and time sequence of releases of (133)Xe. Modeled dilution factors at five different detection locations were combined with 57 atmospheric concentration measurements of (133)Xe taken from March 18 to March 23 to estimate the source term. This analysis suggests that 92% of the 1.24 × 10(19) Bq of (133)Xe present in the three operating reactors at the time of the earthquake was released to the atmosphere over a 3 d period. An uncertainty analysis bounds the release estimates to 54-129% of available (133)Xe inventory. Copyright © 2013 Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

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

2014-06-01

Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates associated with reactor pressure changes in Units 2 and 3. The modified WSPEEDI-II simulation using the new source term reproduced local and regional patterns of cumulative surface deposition of total 131I and 137Cs and air dose rate obtained by airborne surveys. The new source term was also tested using three atmospheric dispersion models (MLDP0, HYSPLIT, and NAME) for regional and global calculations and showed good agreement between calculated and observed air concentration and surface deposition of 137Cs in East Japan. Moreover, HYSPLIT model using the new source term also reproduced the plume arrivals at several countries abroad showing a good correlation with measured air concentration data. A large part of deposition pattern of total 131I and 137Cs in East Japan was explained by in-cloud particulate scavenging. However, for the regional scale contaminated areas, there were large uncertainties due to the overestimation of rainfall amounts and the underestimation of fogwater and drizzle depositions. The computations showed that approximately 27% of 137Cs discharged from FNPS1 deposited to the land in East Japan, mostly in forest areas.

High-frequency stimulation-induced peptide release synchronizes arcuate kisspeptin neurons and excites GnRH neurons

PubMed Central

Qiu, Jian; Nestor, Casey C; Zhang, Chunguang; Padilla, Stephanie L; Palmiter, Richard D

2016-01-01

Kisspeptin (Kiss1) and neurokinin B (NKB) neurocircuits are essential for pubertal development and fertility. Kisspeptin neurons in the hypothalamic arcuate nucleus (Kiss1ARH) co-express Kiss1, NKB, dynorphin and glutamate and are postulated to provide an episodic, excitatory drive to gonadotropin-releasing hormone 1 (GnRH) neurons, the synaptic mechanisms of which are unknown. We characterized the cellular basis for synchronized Kiss1ARH neuronal activity using optogenetics, whole-cell electrophysiology, molecular pharmacology and single cell RT-PCR in mice. High-frequency photostimulation of Kiss1ARH neurons evoked local release of excitatory (NKB) and inhibitory (dynorphin) neuropeptides, which were found to synchronize the Kiss1ARH neuronal firing. The light-evoked synchronous activity caused robust excitation of GnRH neurons by a synaptic mechanism that also involved glutamatergic input to preoptic Kiss1 neurons from Kiss1ARH neurons. We propose that Kiss1ARH neurons play a dual role of driving episodic secretion of GnRH through the differential release of peptide and amino acid neurotransmitters to coordinate reproductive function. DOI: http://dx.doi.org/10.7554/eLife.16246.001 PMID:27549338

Chornobyl Nuclear Power Plant - An Overview of the Current Efforts to Stabilize the Chornobyl Shelter and Establish an Environmentally Safe Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Couch, Daniel P.; Gronier, Serge; Heriot, Ian D.

2005-08-08

Chornobyl Nuclear Power Plant ? An Overview of the Current Efforts to Stabilize the Chornobyl Shelter and Establish an Environmentally Safe Site Abstract?The 1986 accident at the Chornobyl Nuclear Power Plant in Ukraine resulted in the destruction of the reactor core and most of the reactor building. The Chornobyl accident released an enormous quantity of radionuclides into the environment, significantly contaminating a large region around the plant. Within seven months of the accident, the damaged Unit 4 was encased in a massive concrete and steel enclosure known as the Shelter. Deterioration of the Shelter over time poses increasing risks. Themore » Shelter is subject to structural damage or collapse due to wind, snow loading, or seismic activity. Collapse could lead to the release of radioactive fallout. Leakage of rainwater into the Shelter has caused the accumulation of a large quantity of highly radioactive liquid, corrosion of extremely contaminated nuclear fuel debris, and creation of hazardous radioactive dust. To address these concerns, the government of Ukraine, the G7 nations, and additional donor countries adopted the Shelter Implementation Plan (SIP) in 1997. The SIP's objectives are to reduce the risk and potential consequences of accidental collapse of the Shelter; improve nuclear, industrial and environmental safety; and develop a long-term strategy for conversion to an environmentally safe site. Implementation of the SIP has made significant progress that will lead to the construction of a new confinement facility by 2009. (Full paper available by contacting lead author, Dan Couch)« less

Qualification of CASMO5 / SIMULATE-3K against the SPERT-III E-core cold start-up experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grandi, G.; Moberg, L.

SIMULATE-3K is a three-dimensional kinetic code applicable to LWR Reactivity Initiated Accidents. S3K has been used to calculate several international recognized benchmarks. However, the feedback models in the benchmark exercises are different from the feedback models that SIMULATE-3K uses for LWR reactors. For this reason, it is worth comparing the SIMULATE-3K capabilities for Reactivity Initiated Accidents against kinetic experiments. The Special Power Excursion Reactor Test III was a pressurized-water, nuclear-research facility constructed to analyze the reactor kinetic behavior under initial conditions similar to those of commercial LWRs. The SPERT III E-core resembles a PWR in terms of fuel type, moderator,more » coolant flow rate, and system pressure. The initial test conditions (power, core flow, system pressure, core inlet temperature) are representative of cold start-up, hot start-up, hot standby, and hot full power. The qualification of S3K against the SPERT III E-core measurements is an ongoing work at Studsvik. In this paper, the results for the 30 cold start-up tests are presented. The results show good agreement with the experiments for the reactivity initiated accident main parameters: peak power, energy release and compensated reactivity. Predicted and measured peak powers differ at most by 13%. Measured and predicted reactivity compensations at the time of the peak power differ less than 0.01 $. Predicted and measured energy release differ at most by 13%. All differences are within the experimental uncertainty. (authors)« less

Estimate of radiocaesium derived FNPP1 accident in the North Pacific Ocean

NASA Astrophysics Data System (ADS)

Inomata, Yayoi; Aoyama, Michio; Tsubono, Takaki; Tsumune, Daisuke; Yamada, Masatoshi

2017-04-01

134Cs and 137Cs (radiocaesium) were released to the North Pacific Ocean by direct discharge and atmospheric deposition released from the TEPCO Fukushima Dai-ichi Nuclear Power Plant (FNPP1) accident in 2011. After the FNPP1 accident, measurements of 134Cs and 137Cs were conducted by many researches. However, those results are only snapshots in order to interpret the distribution and transport of the released radiocaesium on a basin scale. It is recognized that estimation of the total amount of released 134Cs and 137Cs is necessary to assess the radioecological impacts of their release on the environment. It was reported that the inventory of 134Cs or 137Cs on the North Pacific Ocean after the FNPP1 accident was 15.2-18.3 PBq based on the observations (Aoyama et al., 2016a), 15.3±1.6 PBq by OI analysis (Inomata et al., 2016), 16.1±1.64 PBq by global ocean model (Tsubono et al., 2016). These suggest that more than 75 % of the atmospheric-released radiocaesium (15.2-20.4 PBq; Aoyama et al., 2016a) were deposited on the North Pacific Ocean. The radiocaesium from the atmospheric fallout and direct discharge were expected to mixing as well as diluting near the coastal region and transported eastward across the North Pacific Ocean in the surface layer. Furthermore, radicaesium were rapidly mixed and penetrated into the subsurface water in the North Pacific Ocean in winter. It was revealed that these radiocaesium existed in the Subtropical Mode Water (STMW, Aoyama et al., 2016b; Kaeriyama et al., 2016) and Central Mode Water (CMW, Aoyama et al., 2016b), suggesting that mode water formation and subduction are efficient pathway for the transport of FNPP1 derived radiocaesium into the ocean interior within 1-year timescale. Kaeriyama et al. (2016) estimated the total amount of FNPP1 derived radiocaesium in the STMW was 4.2 ± 1.1 PBq in October-November 2012. However, there is no estimation of the amount of radiocaesium in the CMW. Therefore, it is impossible to discuss about the mass balance of radiocaesium injected into the North Pacific Ocean. In this study, we conducted the optimum interpolation (OI) analysis to estimate the inventory of radiocaesium in the ocean interior as well as surface sweater by using the measured activities. Furthermore, transport speed of radiocaesium in the surface layer in the North Pacific Ocean were also estimated. The data used in this study were derived from all of the available data reported by such as the Tokyo Electric Power Company (TEPCO), the Japanese Ministry of Education, Culture, Sports, Science and Technology (MEXT), and voluntary cargo ships. The data analysis period was until December 2015 after the FNPP1 accident. It was found that the radiocaesium across the North Pacific Ocean were reached to 180˚ E around 40˚ N latitude at July, 2012 by OI analysis. The transport speed was estimated to 8.5 cm s-1. These were reached to the coastal site of America continent and the activities were increased after the year of 2014. The transport speed across 70˚ W (40˚ N latitude) was decreased to 5.2 cm s-1. We estimated the inventory of radiocaesium in the surface seawater (depth; 0-100m) during the periods from August to December, 2012, based on the OI analysis. Amount of 134Cs inventory was estimated to 4.7 PBq with decay-corrected to 1 October 2012 (7.9 PBq at the time on 11 March 2011). (In the case of 137Cs, the inventory was estimated to 12.5 PBq with decay-corrected to 1 October 2012 and 13 PBq at the time on 11 March 2011 which includes pre-Fukushima 137Cs derived from atmospheric weapons test conducted in late 1950s and eraly 1960s). These correspond to 43-53% of the injected 134Cs in the North Pacific Ocean. It was reported that the 4.2±1.1 PBq of 134Cs were distributed in the STMW (Kaeriyama et al., 2016), which corresponds to 22-28% of the injected 134Cs in the North Pacific Ocean. Taking into account these estimation, FNPP1 derived radiocaesium existed in the CMW in the North Pacific Ocean would be about 3-6 PBq. (References) Aoyama, M., Hamajima, Y., Hult, M., Uematsu, M., Oka, E., . 2016. 134Cs and 137Cs in the North Pacific Ocean derived from the March 2011 TEPCO Fukushima Dai-Ichi Nuclear Power Plant accident, Japan. Part one: surface pathway and vertical distributions. J. Oceanogr. 72:53-65. Aoyama, M., Kajino, M., Tanaka, T. Y., Sekiyama, T.T., Tsumune, D., Tsubono, T., Hamajima, Y., Inomata, Y., Gamo, T., .2016. 134Cs and 137Cs in the North Pacific Ocean derived from the March 2011 TEPCO Fukushima Daiichi Nuclear Power Plant accident, Japan. Part two: estimation of 134Cs and 137Cs inventories in the North Pacific Ocean. J. Oceanogr. 72:53-65. Inomata, Y., Aoyama, M., Tsubono, T., Tsumune, D., Hirose, K. 2016. Spatial and temporal distributions of 134Cs and 137Cs derived from the TEPCOFukushima Daiichi nuclear power plant accident in the North Pacific Ocean by using optimal interpolation analysis. Environ. Sci. Process. Impacts 18:126-36. Tsubono, T., Misumi. K., Tsumune, D., Bryan, F.O., Hirose, K., Aoyama, M. 2016. Evaluation of radioactive cesium impact from atmospheric deposition and direct release fluxes into the North Pacific from the Fukushima Daiichi nuclear power plant. Deep-Sea Res. I. 115:10-21. Kaeriyama, H., Shimizu, Y., Setou, T., Kumamoto, Y., Okazaki, M., Ambe, D., Ono, T. 2016. Intrusion of Fukushima-derived radiocaesium into subsurface water due to formation of mode waters in the North Pacific. Sci. Report., 6:22010 | DOI: 10.1038/srep22010.

Recent MELCOR and VICTORIA Fission Product Research at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Cole, R.K.; Gauntt, R.O.

1999-01-21

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes inmore » the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.« less

Source, dispersion and combustion modelling of an accidental release of hydrogen in an urban environment.

PubMed

Venetsanos, A G; Huld, T; Adams, P; Bartzis, J G

2003-12-12

Hydrogen is likely to be the most important future energy carrier, for many stationary and mobile applications, with the potential to make significant reductions in greenhouse gas emissions especially if renewable primary energy sources are used to produce the hydrogen. A safe transition to the use of hydrogen by members of the general public requires that the safety issues associated with hydrogen applications have to be investigated and fully understood. In order to assess the risks associated with hydrogen applications, its behaviour in realistic accident scenarios has to be predicted, allowing mitigating measures to be developed where necessary. A key factor in this process is predicting the release, dispersion and combustion of hydrogen in appropriate scenarios. This paper illustrates an application of CFD methods to the simulation of an actual hydrogen explosion. The explosion occurred on 3 March 1983 in a built up area of central Stockholm, Sweden, after the accidental release of approximately 13.5 kg of hydrogen from a rack of 18 interconnected 50 l industrial pressure vessels (200 bar working pressure) being transported by a delivery truck. Modelling of the source term, dispersion and combustion were undertaken separately using three different numerical tools, due to the differences in physics and scales between the different phenomena. Results from the dispersion calculations together with the official accident report were used to identify a possible ignition source and estimate the time at which ignition could have occurred. Ignition was estimated to occur 10s after the start of the release, coinciding with the time at which the maximum flammable hydrogen mass and cloud volume were found to occur (4.5 kg and 600 m(3), respectively). The subsequent simulation of the combustion adopts initial conditions for mean flow and turbulence from the dispersion simulations, and calculates the development of a fireball. This provides physical values, e.g. maximum overpressure and far-field overpressure that may be used as a comparison with the known accident details to give an indication of the validity of the models. The simulation results are consistent with both the reported near-field damage to buildings and persons and with the far-field damage to windows. The work was undertaken as part of the European Integrated Hydrogen Project-Phase 2 (EIHP2) with partial funding from the European Commission via the Fifth Framework Programme.

Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

NASA Astrophysics Data System (ADS)

Kobayashi, K.; Isobe, K.; Iwai, Y.; Hayashi, T.; Shu, W.; Nakamura, H.; Kawamura, Y.; Yamada, M.; Suzuki, T.; Miura, H.; Uzawa, M.; Nishikawa, M.; Yamanishi, T.

2007-12-01

Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R&D task: (1) tritium behaviour in concrete; (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected that the electrolysis cell of the ITER-WDS could endure 3 years' operation under the ITER design conditions. Measuring the concentration of the fluorine ions could be a promising technique for monitoring the damage to the electrolysis cell.

Introduction and Mission Response Team (MRT)

NASA Technical Reports Server (NTRS)

Pool, Sam

2005-01-01

On February 1, 2003 the Space Shuttle Columbia, returning to Earth with a crew of seven astronauts, disintegrated along a track extending from California to Louisiana. Observers on the ground filmed breakup of the spacecraft. Debris fell along a 567 statute mile track from Littlefield, Texas to Fort Polk, Louisiana; the largest ever recorded debris field. At the time of the accident the National Aeronautics and Space Administration (NASA) flight surgeon on-duty at the Mission Control Center (MCC) in Houston, Texas initiated the medical contingency response. The DOD surgeon at Patrick Air Force Base was notified, NASA medical personnel were recalled and the services of Armed Forces Institute of Pathology (AFIP) were requested. Subsequent to the accident the NASA flight surgeons that had supported the crew on orbit now provided medical support to the crewmember s families. Federal Emergency Management Agency (FEMA), the National Transportation Safety Board (NTSB), the Federal Bureau of Investigation (FBI) and numerous other federal, state and local agencies along with the citizens of Texas and Louisiana responded to the disaster. Search and recovery was managed from a Disaster Field Office (DFO) established in Lufkin, Texas. Mishap Investigation Team (MIT) medical operations were managed from Barksdale Air Force Base, Louisiana. Accident investigation teams (Columbia Accident Investigation Task Force (CAITF) and Columbia Accident Investigation Board (CAIB)) appointed immediately after the disaster included current and former authorities in space medicine. In August 2003, the CAIB concluded its investigation and released its findings in a report published in February 2004.

Waste management facility accident analysis (WASTE ACC) system: software for analysis of waste management alternatives

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohout, E.F.; Folga, S.; Mueller, C.

1996-03-01

This paper describes the Waste Management Facility Accident Analysis (WASTE{underscore}ACC) software, which was developed at Argonne National Laboratory (ANL) to support the US Department of Energy`s (DOE`s) Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). WASTE{underscore}ACC is a decision support and database system that is compatible with Microsoft{reg_sign} Windows{trademark}. It assesses potential atmospheric releases from accidents at waste management facilities. The software provides the user with an easy-to-use tool to determine the risk-dominant accident sequences for the many possible combinations of process technologies, waste and facility types, and alternative cases described in the WM PEIS. In addition, its structure willmore » allow additional alternative cases and assumptions to be tested as part of the future DOE programmatic decision-making process. The WASTE{underscore}ACC system demonstrates one approach to performing a generic, systemwide evaluation of accident risks at waste management facilities. The advantages of WASTE{underscore}ACC are threefold. First, the software gets waste volume and radiological profile data that were used to perform other WM PEIS-related analyses directly from the WASTE{underscore}MGMT system. Second, the system allows for a consistent analysis across all sites and waste streams, which enables decision makers to understand more fully the trade-offs among various policy options and scenarios. Third, the system is easy to operate; even complex scenario runs are completed within minutes.« less

The Role of Deposition in Limiting the Hazard Extent of Dense-Gas Plumes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dillon, M B

2008-01-29

Accidents involving release of large (multi-ton) quantities of toxic industrial chemicals often yield far fewer fatalities and causalities than standard, widely-used assessment and emergency response models predict. While recent work has suggested that models should incorporate the protection provided by buildings, more refined health effect methodologies, and more detailed consideration of the release process; investigations into the role of deposition onto outdoor surfaces has been lacking. In this paper, we examine the conditions under which dry deposition may significantly reduce the extent of the downwind hazard zone. We provide theoretical arguments that in congested environments (e.g. suburbs, forests), deposition tomore » vertical surfaces (such as building walls) may play a significant role in reducing the hazard zone extent--particularly under low-wind, stable atmospheric conditions which are often considered to be the worst-case scenario for these types of releases. Our analysis suggests that in these urban or suburban environments, the amount of toxic chemicals lost to earth's surface is typically a small fraction of overall depositional losses. For isothermal gases such as chlorine, the degree to which the chemicals stick to (or react with) surfaces (i.e. surface resistance) is demonstrated to be a key parameter controlling hazard extent (the maximum distance from the release at which hazards to human health are expected). This analysis does not consider the depositional effects associated with particulate matter or gases that undergo significant thermal change in the atmosphere. While no controlled experiments were available to validate our hypothesis, our analysis results are qualitatively consistent with the observed downwind extent of vegetation damage in two chlorine accidents.« less

Two-dimensional fuzzy fault tree analysis for chlorine release from a chlor-alkali industry using expert elicitation.

PubMed

Renjith, V R; Madhu, G; Nayagam, V Lakshmana Gomathi; Bhasi, A B

2010-11-15

The hazards associated with major accident hazard (MAH) industries are fire, explosion and toxic gas releases. Of these, toxic gas release is the worst as it has the potential to cause extensive fatalities. Qualitative and quantitative hazard analyses are essential for the identification and quantification of these hazards related to chemical industries. Fault tree analysis (FTA) is an established technique in hazard identification. This technique has the advantage of being both qualitative and quantitative, if the probabilities and frequencies of the basic events are known. This paper outlines the estimation of the probability of release of chlorine from storage and filling facility of chlor-alkali industry using FTA. An attempt has also been made to arrive at the probability of chlorine release using expert elicitation and proven fuzzy logic technique for Indian conditions. Sensitivity analysis has been done to evaluate the percentage contribution of each basic event that could lead to chlorine release. Two-dimensional fuzzy fault tree analysis (TDFFTA) has been proposed for balancing the hesitation factor involved in expert elicitation. Copyright © 2010 Elsevier B.V. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan; Faucett, Christopher; Haskin, Troy Christopher

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recoverymore » of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.« less

On-line measurements of RuO{sub 4} during a PWR severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reymond-Laruinaz, S.; Doizi, D.; Manceron, L.

After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is notmore » trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)« less

INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

GREENE,G.A.; GUPPY,J.G.

1998-09-01

This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install andmore » make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.« less

Simulation of internal contamination screening with dose rate meters

NASA Astrophysics Data System (ADS)

Fonseca, T. C. F.; Mendes, B. M.; Hunt, J. G.

2017-11-01

Assessing the intake of radionuclides after an accident in a nuclear power plant or after the intentional release of radionuclides in public places allows dose calculations and triage actions to be carried out for members of the public and for emergency response teams. Gamma emitters in the lung, thyroid or the whole body may be detected and quantified by making dose rate measurements at the surface of the internally contaminated person. In an accident scenario, quick measurements made with readily available portable equipment are a key factor for success. In this paper, the Monte Carlo program Visual Monte Carlo (VMC) and MCNPx code are used in conjunction with voxel phantoms to calculate the dose rate at the surface of a contaminated person due to internally deposited radionuclides. A whole body contamination with 137Cs and a thyroid contamination with 131I were simulated and the calibration factors in kBq per μSv/h were calculated. The calculated calibration factors were compared with real data obtained from the Goiania accident in the case of 137Cs and the Chernobyl accident in terms of the 131I. The close comparison of the calculated and real measurements indicates that the method may be applied to other radionuclides. Minimum detectable activities are discussed.

The inhibition of cholera toxin-induced 5-HT release by the 5-HT3 receptor antagonist, granisetron, in the rat

PubMed Central

Turvill, J L; Connor, P; Farthing, M J G

2000-01-01

The secretagogue 5-hydroxytryptamine (5-HT) is implicated in the pathophysiology of cholera. 5-HT released from enterochromaffin cells after cholera toxin exposure is thought to activate non-neuronally (5-HT2 dependent) and neuronally (5-HT3 dependent) mediated water and electrolyte secretion. CT-secretion can be reduced by preventing the release of 5-HT. Enterochromaffin cells possess numerous receptors that, under basal conditions, modulate 5-HT release. These include basolateral 5-HT3 receptors, the activation of which is known to enhance 5-HT release. Until now, 5-HT3 receptor antagonists (e.g. granisetron) have been thought to inhibit cholera toxin-induced fluid secretion by blockading 5-HT3 receptors on secretory enteric neurones. Instead we postulated that they act by inhibiting cholera toxin-induced enterochromaffin cell degranulation. Isolated intestinal segments in anaesthetized male Wistar rats, pre-treated with granisetron 75 μg kg−1, lidoocaine 6 mg kg−1 or saline, were instilled with a supramaximal dose of cholera toxin or saline. Net fluid movement was determined by small intestinal perfusion or gravimetry and small intestinal and luminal fluid 5-HT levels were determined by HPLC with fluorimetric detection. Intraluminal 5-HT release was proportional to the reduction in tissue 5-HT levels and to the onset of water and electrolyte secretion, suggesting that luminal 5-HT levels reflect enterochromaffin cell activity. Both lidocaine and granisetron inhibited fluid secretion. However, granisetron alone, and proportionately, reduced 5-HT release. The simultaneous inhibition of 5-HT release and fluid secretion by granisetron suggests that 5-HT release from enterochromaffin cells is potentiated by endogenous 5-HT3 receptors. The accentuated 5-HT release promotes cholera toxin-induced fluid secretion. PMID:10882387

Real-time noble gas release signaling rock deformation

NASA Astrophysics Data System (ADS)

Bauer, S. J.; Gardner, W. P.; Lee, H.

2016-12-01

We present empirical results/relationships of rock strain, microfracture density, acoustic emissions, and noble gas release from laboratory triaxial experiments for a granite and basalt. Noble gases are contained in most crustal rock at inter/intra granular sites, their release during natural and manmade stress and strain changes represents a signal of brittle/semi brittle deformation. The gas composition depends on lithology, geologic history and age, fluids present, and uranium, thorium and potassium-40 concentrations in the rocks that affect radiogenic noble gases (helium, argon) production. Noble gas emission and its relationship to crustal processes have been studied, including correlations to tectonic velocities and qualitative estimates of deep permeability from surface measurements, finger prints of nuclear weapon detonation, and as potential precursory signals to earthquakes attributed to gas release due to pre-seismic stress, dilatancy and/or rock fracturing. Helium emission has been shown as a precursor of volcanic activity. Real-time noble gas release is observed using an experimental system utilizing mass spectrometers to measure gases released during triaxial rock deformation. Noble gas release is shown to represent a sensitive precursor signal of rock deformation by relating real-time noble gas release to stress-strain state changes and acoustic emissions. We propose using noble gas release to also signal rock deformation in boreholes, mines and nuclear waste repositories. We postulate each rock exhibits a gas release signature which is microstructure, stress/strain state, and or permanent deformation dependent. Such relationships, when calibrated, may be used to sense rock deformation and then develop predictive models. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corp., for the US Dept. of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND2016-7468 A

Carbon/graphite composite material study, appendix A and appendix B

NASA Technical Reports Server (NTRS)

1981-01-01

A comprehensive assessment of the possible damage to electrical and electronic equipment caused by accidental release of carbon fibers from burning civil aircraft with carbon composite parts was completed. The study concluded that the amount of fiber likely to be released is much lower than initially predicted. Carbon fiber released from an aircraft crash fire was found (from atmospheric dissemination models) to disperse over a much larger area than originally estimated, with correspondingly lower fiber concentrations. Long term redissemination of fiber was shown to be insignificant if reasonable care is exercised in accident cleanup. The vulnerability of electrical equipment to structural fibers in current use was low. Consumer appliances, industrial electronics, and avionics were essentially invulnerable to carbon fibers. Shock hazards (and thus potential injury or death) were found to be extremely unlikely.

A scheme for the classification of explosions in the chemical process industry.

PubMed

Abbasi, Tasneem; Pasman, H J; Abbasi, S A

2010-02-15

All process industry accidents fall under three broad categories-fire, explosion, and toxic release. Of these fire is the most common, followed by explosions. Within these broad categories occur a large number of sub-categories, each depicting a specific sub-type of a fire/explosion/toxic release. But whereas clear and self-consistent sub-classifications exist for fires and toxic releases, the situation is not as clear vis a vis explosions. In this paper the inconsistencies and/or shortcomings associated with the classification of different types of explosions, which are seen even in otherwise highly authentic and useful reference books on process safety, are reviewed. In its context a new classification is attempted which may, hopefully, provide a frame-of-reference for the future.

Chernobyl and Fukushima nuclear accidents: what has changed in the use of atmospheric dispersion modeling?

PubMed

Benamrane, Y; Wybo, J-L; Armand, P

2013-12-01

The threat of a major accidental or deliberate event that would lead to hazardous materials emission in the atmosphere is a great cause of concern to societies. This is due to the potential large scale of casualties and damages that could result from the release of explosive, flammable or toxic gases from industrial plants or transport accidents, radioactive material from nuclear power plants (NPPs), and chemical, biological, radiological or nuclear (CBRN) terrorist attacks. In order to respond efficiently to such events, emergency services and authorities resort to appropriate planning and organizational patterns. This paper focuses on the use of atmospheric dispersion modeling (ADM) as a support tool for emergency planning and response, to assess the propagation of the hazardous cloud and thereby, take adequate counter measures. This paper intends to illustrate the noticeable evolution in the operational use of ADM tools over 25 y and especially in emergency situations. This study is based on data available in scientific publications and exemplified using the two most severe nuclear accidents: Chernobyl (1986) and Fukushima (2011). It appears that during the Chernobyl accident, ADM were used few days after the beginning of the accident mainly in a diagnosis approach trying to reconstruct what happened, whereas 25 y later, ADM was also used during the first days and weeks of the Fukushima accident to anticipate the potentially threatened areas. We argue that the recent developments in ADM tools play an increasing role in emergencies and crises management, by supporting stakeholders in anticipating, monitoring and assessing post-event damages. However, despite technological evolutions, its prognostic and diagnostic use in emergency situations still arise many issues. Copyright © 2013 Elsevier Ltd. All rights reserved.

Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

NASA Astrophysics Data System (ADS)

Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

2015-10-01

In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

Vegetation fires, smoke emissions, and dispersion of radionuclides in the chernobyl exclusion zone

Treesearch

Wei Min Hao; Oleg O. Bondarenko; Sergiy Zibtsev; Diane Hutton

2009-01-01

The accident of the Chernobyl nuclear power plant (ChNPP) in 1986 was probably the worst environmental disaster in the past 30 years. The fallout and accumulation of radionuclides in the soil and vegetation could have long-term impacts on the environment. Radionuclides released during large, catastrophic vegetation fires could spread to continental Europe, Scandinavia...

Health Aspects of a Nuclear or Radiological Attack

DTIC Science & Technology

2010-07-01

treatment capabilities, population monitoring, decontamination, and so on. The requirements should be developed by a process that engages stakeholders...IND or reactor accident would release radioiodine that could result in hypothyroidism or late thyroid nodules or cancer. External contamination... Treatment and Long-Term Monitoring Victims of acute radiation-related events will require prompt diagnosis and treatment of emergency medical and

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

Progress Is No Accident: Why ESEA Can't Backtrack on High School Graduation Rates

ERIC Educational Resources Information Center

Alliance for Excellent Education, 2015

2015-01-01

In conjunction with the White House Next Gen High School Summit on November 10, a new report finds that the number of high school dropouts decreased from 1 million in 2008 to approximately 750,000 in 2012. The report, released by the Alliance for Excellent Education, America's Promise Alliance, Civic Enterprises, and Everyone Graduates Center,…