Energy Technology Data Exchange (ETDEWEB)

The personal neutron dosimetry becomes more important with the increasing use of nuclear and accelerator facilities. The solid state track detector, film badge and thermo-luminescent dosimeter have widely been used as passive-type personal neutron dosimeters, but the real-time neutron dosimeter is strongly needed. A real time personal dosimeter which could give neutron dose equivalent over wide energy range from thermal to 10 odd MeV by using two neutron sensors has been developed by our group. For practical commercial product, some changes from these prototype sensors have recently been done by Fuji Electric Co. Ltd. The purpose of this work is the final performance test of the dosimeter just before sale. We checked again about the sensitivity in the monoenergetic neutron field from 8 keV to 15 MeV and in the neutron fields around a few accelerator facilities. ...

Energy Technology Data Exchange (ETDEWEB)

DOE installations possessing sufficient quantities of fissile material to potentially constitute a critical mass, such that the excessive exposure of personnel to radiation from a nuclear accident is possible, are required to provide nuclear accident dosimetry services. This document describes the personal nuclear accident dosimeter (PNAD) used by SNL and prescribes methodologies to initially screen, and to process PNAD results. In addition, this report describes PNAD dosimetry results obtained during the Nuclear Accident Dosimeter Intercomparison Study (NAD23), held during 12-16 June 1995, at Los Alamos National Laboratories. Biases for reported neutron doses ranged from -6% to +36% with an average bias of +12%.

Energy Technology Data Exchange (ETDEWEB)

High energy physics facilities must monitor the radiation doses received by their personnel. This monitoring can only be effective if the radiation detection devices can be calibrated with a good degree of accuracy. Radiation fields are usually composed of several types of radiation, including gamma rays, beta radiation, neutrons, etc. The neutron detection instruments respond not only to the neutrons coming directly from the source but also those scattered from the floor, walls, and ceiling. The amount of neutron scattering varies from site to site depending on the construction materials and layout of the building. The purpose of this study was to determine the scattered neutron fraction in the central volume of the calibration mezzanine of the Radiation Physics Calibration Facility (RPCF) at the Fermi National Accelerator Laboratory (Fermilab). At Fermilab, radiation workers dosimeters use CR39 for neutron detection, which are sent to an outside vendor for ...

Energy Technology Data Exchange (ETDEWEB)

This article consists of a series of manufacturer's descriptions outlining specifications of underground monitoring equipment which range in complexity from gas sensitive personal badges to full mine monitoring systems. 8 figs.

Science.gov (United States)

Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents Characterization Contingency Plans National Contingency Plan (NCP), Oil Removal...

Science.gov (United States)

Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents Characterization Contingency Plans National Contingency Plan (NCP), Oil Removal...

International Nuclear Information System (INIS)

The optically stimulated luminescent (OSL) radiation dosimeter technically surveys a wide dynamic measurement range and a high sensitivity. Optical fiber dosimeters provide capability for remote monitoring of the radiation in the locations which are difficult-to-access and hazardous. In addition, optical fiber dosimeters are immune to electrical and radio-frequency interference. In this paper, a novel remote optical fiber radiation dosimeter is described. The optical fiber dosimeter takes advantage of the charge trapping materials CaS:Ce, Sm that exhibit OSL. The measuring range of the dosimeter is from 0.1 to 100 Gy. The equipment is relatively simple and small in size, and has low power consumption. This device is suitable for measuring the space radiation dose and also can be used in high radiation dose condition and other dangerous radiation occasions. ...

International Nuclear Information System (INIS)

It needs accurate estimation of radiation level for verifying machinery and cable in Nuclear Power Plant. Therefore, in this study, we used ESR(Electron Spin Resonance) system for estimate dose of Alanine dosimeter. Alanine/ESR dosimetry, already known as a dosimetric method in medical and industrial field, was applied to estimate dose quantity at cable locations within a nuclear power plant as a part of equipment qualification program. Alanine/ESR dosimetry of absorbed dose range is 1 - 100 KGy. The alanine dosimeter is not significantly affected by temperature and fading is limited to 1% per year. The alanine dosimeters were fixed on the targeted cable or nearest position to measure dose quantity to get accurate value. Alanine dosimeters were scanned by commercially used two different ESR systems, e-scan and EMX series for alanine dosimeters. To estimate more accurate dose, two ...

International Nuclear Information System (INIS)

As part of a general study of personnel neutron dosimetry systems, three British and two German TL albedo dosimeters were compared. They range from simple albedo devices to multi-element devices which detect combinations of incident and albedo neutrons. They were exposed to beams of mono-energetic neutrons, a thermal neutron column and a "2"5"2Cf source. The variation in response with neutron energy above 0.1 MeV was similar for all dosimeters. The resuls show that in general multi-component dosimeters fo not provide enough information to deduce the effective energy of incident neutrons. The choice of single or multi-element dosimeters for particular types of environment is discussed. The Harvey dosimeter has the most acceptable thermal neutron response of the simpler devices. For neutron energies below 0.5 MeV its response varies roughly as the cosine of the angle of incidence but ...

Energy Technology Data Exchange (ETDEWEB)

Neutron spectra measured in the FFTF In-Reactor Thimble (IRT) by proton recoil proportional counters, proton recoil emulsions, and passive dosimeters have been evaluated and compared with each other and with three-dimensionl diffusion theory calculated spectra for the purpose of validating the passive dosimeter spectrum adjustment technique. The least squares data adjustment code, FERRET, was used to combine measured reaction rates, calculated spectra, and dosimeter cross sections, resulting in adjusted spectra and cross sections with uncertainties and correlations that properly account for uncertainties and correlations on the input parameters.

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

Alanine/EPR Dosimetry system is classified as a reference-standard dosimeter which used to calibrate radiation environments and to calibrate routine dosimeters in absorbed dose range of 1 to 105 Gy. Characteristics of Alanine/EPR dosimetry system are better suitable than routinely used personnel dosimeters for the long term radiation measurement in harsh condition of Nuclear power plant : This system is not significantly affected by temperature and humidity and also has low fading rate. About 1 year ago, Alanine and Lithium compound dosimeters were installed at Wolsung unit 1 for the environment monitoring as a part of equipment qualification program. 35 positions which are most severe and representative cable positions were chosen for radiation and temperature monitoring. The recovered alanine dosimeter in the period of maintenance was measured by E-scan alanine analyzer system, ...

Energy Technology Data Exchange (ETDEWEB)

This report shows the probability of a catastrophic accident involving the WIPP waste hoist system. Calculations and mitigation to reduce the probability of an accident and to minimize the impact of such an accident should be included. 10 refs., 8 figs., 4 tabs.

International Nuclear Information System (INIS)

Short communication.

International Nuclear Information System (INIS)

... computer calculations fftf reactor nonlinear problems reactor accidents reactor

International Nuclear Information System (INIS)

The themal neutron sensitivities of SEI and Toshiba FD-P10-7 radiophotoluminescent silver activated metaphosphate glass dosimeters have been measured and studied theoretically. The graphite thermal neutron column attached to the AEEW reactor NESTOR was utilised as a source of thermal neutrons with minimal fast neutron and gamma-ray contamination. Both the neutron removal and gamma correction measurement techniques were applied and these gave consistent results. It was found that the thermal neutron sensitivity of the Toshiba dosimeters is comparable to the most neutron intensitive thermoluminescent materials available. Because of their larger size and higher silver concentration, however, SEI dosimeters are an order of magnitude more neutron sensitive. The neutron sensitivity of FD-7 glass has not been reported previously, whereas in the case of SEI glass, there exists a modest body of previous work on ...

International Nuclear Information System (INIS)

The properties of a new commercial thermoluminescent dosimeter, CaSO4:Dy without a LiF binder (TLD-900) was studied for low-exposure measurements in diagnostic radiology. The former TLD-900 had a LiF binder (herein referred to as ''TLD-900/LiF''). The principle features of this dosimeter are its high sensitivity to low-energy radiation and its relatively low fading which permits measurements down to less than 2.6 X 10(-8) C kg (0.1 mR) with an accuracy better than 20%. The characteristics, annealing procedures, light sensitivity, energy response, reproducibility, and fading, of TLD-900 are discussed and compared with TLD-900/LiF. When the precautions presented in this paper are used, the dosimeters can be used for the measurement of x-ray exposures.

International Nuclear Information System (INIS)

Macroparticle acceleration driven by low current, high voltage cathode spots has been investigated for potential applications ranging from micrometeoroid simulation to nanoparticle deposition/implantation. Acceleration by this process was observed to occur when nanometer and micrometer-sized particles were exposed to a high voltage pulse in the presence of a plasma discharge. The applied negative voltage pulse initiates the formation of multiple, high voltage, low current cathode spots which provide the mechanism of actual acceleration of the charged dust particles. Dust streams generated by this process were detected using laser scattering techniques. Cathode spot behavior was also documented. The particle impact craters observed at the surface of downstream witness badges were documented using scanning electron microscopy and light microscopy. The observed impacts suggest the presence of energetic macroparticles formed during this process.

International Nuclear Information System (INIS)

1982. p. 4.3-4.6. Israel Eisen, Y. Engler, G. Ovadia, E. Shamai, Y. Israel

Energy Technology Data Exchange (ETDEWEB)

The aim of this work is to determine and compare the dosimetric properties of NaCl and KCl samples using infrared-stimulated luminescence (IRSL) technique. For a material to be used as dosimeter, both the IRSL temperature dependence and the radiation dose response have critical importance. In this work the IRSL characteristics from NaCl and KCl samples were experimentally investigated as a function of temperature and laboratory radiation doses. Dosimetric properties of NaCl and KCl samples were found significantly different. The IRSL signals displayed by NaCl were found to be more stable, reliable and agreeable than those of KCl.

International Nuclear Information System (INIS)

To make a quantitative comparison of risks between the SNR 300 and a modern PWR (Biblis B), the consequences of an accident or the extent of damage of a release of radionuclides to the environment due to an accident are estimated by computer programs for accident consequence models. The accident analysis includes an analysis of events for Bethe-Tait accidents with failure of the outer containment. The FGSB release rates are compared with those of the Society for Reactor Safety (GRS). (HP).

UK PubMed Central (United Kingdom)

Objective: To determine the experience of junior doctors cited as witnesses at fatal accident inquiries (FAIs). Design: Retrospective questionnaire study. Setting...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

The report gives general principles of treatment and care of casualties caused by radiation accidents or nuclear explosions.

Science.gov (United States)

... radiation during clinical therapy and exposures due to radiation accidents or attacks, in which the doses are uncontrolled ... only be used off-label in victims of radiation accidents or attacks. The idea...

International Nuclear Information System (INIS)

Full text: Semiconductor detectors coated with boron or lithium compounds have been studied for neutron detection for decades but, until recently, have been limited to thermal neutron detection efficiencies of less than 5%. We reported previously on development and simulation studies of perforated detectors whose perforations are filled with neutron-reactive material in order to produce higher detection efficiencies. Incorporation of bare and cadmium-backed detectors into battery-powered devices with low-power electronics enables us to produce compact personal neutron dosimeters that provide LED readout of counts, which can be related approximately to neutron dose. We report here on experimental studies with such compact devices; devices capable of direct readout in dose units are anticipated. The thermal and epithermal neutron flux densities from the tangential beam tube of the TRIGA Mark II reactor at Kansas State University were measured. Then, thermoluminescent ...

Science.gov (United States)

The biological age (BA) of the majority of the liquidators of the consequences of the radiation accidents in the Navy and of the liquidators of the Chernobyl' APS accident exceeds the medium standard and the DBA (due BA). The index of the BA can be a characteristic of the influence of the social-hygienic factors on the health condition of the Special Risk Subunit--the liquidators of the consequences of the radiation accidents. It was established, that the radiation influence concerns to the factors dramatically increasing the BA and the rate of senescence of the liquidators of the consequences of the radiation accidents. PMID:21809627

International Nuclear Information System (INIS)

The consumption of food products obtained in areas subjected to radioactive contamination as a consequence of a radiation accident appears to be the most significant source of irradiation for the population. At the same time, this route can be regulated very effectively. The regularities of contamination of agricultural production, peculiar features of internal dose formation in the population and the effectiveness of countermeasures in agriculture have been analysed using the experience of two major accidents in the former USSR - in the South Urals (Kyshtym accident) in 1957, and at the Chernobyl NPP in 1986. (Author).

International Nuclear Information System (INIS)

Short communication.

Energy Technology Data Exchange (ETDEWEB)

With the exception of the 1945 Hiroshima and Nagasaki nuclear weapon explosions and the 1986 Tchernobyl reactor accident, most of the radiation accidents concerns the medical and the traditional industrial sectors. The seriousness of the accident is directly function of the absorbed dose. The paper, first, gives the definition of a radiologic accident with its specific criteria and pathological manifestations. Then, some famous historical accidents are reviewed from the discovery of X-rays to recent acute irradiations due to the careless manipulation of radiation sources. From this analysis, three main causes are put forward: the dysfunction of nuclear medicine apparatuses, the victims` lack of training and knowledge of the risks, and the non-identification or the loss of radiation sources. (J.S.). 1 photo.

Energy Technology Data Exchange (ETDEWEB)

Beneath the harsh deserts of Saudi Arabia lie dark chambers and complex mazes filled with strange shapes and wondrous beauty. Radon concentration measurements have been carried out in the desert caves of Al-Somman Plateau in the Eastern Province of Saudi Arabia. Passive radon dosimeters, based on alpha particle etch track detectors with an inlet filter, were used in this study. A total of 59 dosimeters were placed in five caves for a period of six months. Out of 59 dosimeters, 37 could be collected for analysis. Measurements showed significant variations in radon concentrations in caves depending upon their natural ventilation. The results of the study show that the average radon concentration in the different caves ranges from 74 up to 451Bqm{sup -3}. The average radon concentration in four of the caves was low in the range 74-114Bqm{sup -3}. However, one cave showed an average radon concentration of 451Bqm{sup -3}. Radon ...

Energy Technology Data Exchange (ETDEWEB)

The recent definition of a postulated thermal shock accident followed promptly by system repressurization, termed an overcooling or pressurized thermal shock accident, has set a large analysis and research effort into motion. The essential elements are concerned with defining the accident transients, evaluating the instrumentation and controls that cause the postulated accidents, and evaluating the metallurgical and structural mechanics aspects of the reactor vessel with respect to its failure potential. This paper poses the question faced by the Nuclear Regulatory Commission (NRC) for the vessel steel embrittlement, annealing, and surveillance dosimetry facets of this postulated accident and provides information on our plans for study of this problem as well as current status.

International Nuclear Information System (INIS)

The radiolytic reduction of the Ce4+ in a 0.8N sulfuric acid was examined, with the objective of measuring large radiation doses. the ceric sulphate dosimeter was tested in the range from 1 to 20kGy. The 0.01 M sodium oxalate solution was chosen instead of the arsenious oxide solution (used in the conventional method), in order to obtain a less expensive dosimetric system than that normally use for routine chemical dosimetry. The absorption spectrum of the Ce"4"+ solution presents an absorption peak at 370 nm; the obtained molar extinction coefficients of Ce"4"+ and Ce"3"+ were 5670 and 2.7 M"-"1 cm"-"1 respectively. The purpose of this work was to establish a simple, reliable and inexpensive routine chemical dosimetric system for high doses.

Energy Technology Data Exchange (ETDEWEB)

Severe accident analysis has been performed for the Wolsung nuclear power= plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks (LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code) computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanisms of calandria vessel and containment. In addition, some insights for accident management program (AMP) are also given. (Author) 5 refs., 1 tab., 12 figs.

International Nuclear Information System (INIS)

SCDAP/RELAP5 code has been developed in US for best-estimate simulation of light water reactors transients during nuclear accidents. The code models the coupled behaviour of the cooling system, reactor core and fission products release during the accident. It is the result of the coupling between RELAP5, modelling thermal hydraulic, control system, reactor kinetics and the transport of noncondensable gases, and SCDAP code modelling the behaviour of the reactor core during severe accidents. The paper briefly presents the application of SCDAP/RELAP5 code to CANDU severe accident analysis. Also, the paper proposes a summary of the needs for development that could enhance the quality of the severe accidents related predictions in CANDU reactors. (authors)

International Nuclear Information System (INIS)

Containment is considered as ultimate safety system and is designed to enclose whole reactor system and prevent the spread of active air-borne fission products. For Pressure and Temperature calculation, Design Basis Accident (Dba) is double ended break of reactor inlet header or main steam line break but activity release studies are done to access its performance following limiting design basis accident i.e. Loss of Coolant Accident (LOCA) and Emergency Core Cooling System (ECCS). In such accident scenario, the core is severely damaged and results in production of steam and hydrogen along with release of activity to containment environment. Containment functions are maintained in such accident, and radiological consequences are within the prescribed limits. (author)

International Nuclear Information System (INIS)

The National Radiological Protection Board is to carry out a survey of doses received by patients from the diagnostic use of X-rays, beginning in June 1977, and carried out at over 100 National Health Service hospitals. Details are given of the reasons for carrying out the survey and the survey objectives. The survey working methods are discussed, together with methods of calculation of the genetically significant dose. The radiation doses will be measured by a thermoluminescent dosimeter developed specially for the survey. (U.K.).

Energy Technology Data Exchange (ETDEWEB)

The calculations with the broad-group cross-section library Bugle-96, and atom displacement (dpa) cross sections for iron, both derived from ENDF/B-VI data, result in higher calculated fast neutron fluxes, better agreement of calculations with radiometric dosimeter measurements, and significantly slower dpa rate attenuation through pressure vessel walls relative to the results with their predecessors: the Sailor library and ASTM iron dpa cross sections.

Energy Technology Data Exchange (ETDEWEB)

In 1998, research began at Idaho National Engineering and Environmental Laboratory to investigate the application of {sup 6}Li and {sup 7}Li isotopes to the measurement of neutron and gamma radiation. Various size pairs of {sup 6}Li and {sup 7}Li based detectors were exposed to mixed neutron and gamma radiation. Experiments demonstrated that these detectors could be used to measure low level neutron radiation in the presence of high level gamma radiation. (author)

Science.gov (United States)

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

UK PubMed Central (United Kingdom)

This is a retrospective study of the development of the social worker role within the multi-disciplinary team setting of the Accident and Emergency (A&E) Department at Burnley General Hospital...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

... 5 360 p. RADIATION PROTECTION AND DOSIMETRY fire fighting occupational

International Nuclear Information System (INIS)

The early accident Sequence of the Station Blackout accident is simulated for Daya Bay Nuclear Power Plant, using MELCOR code. The radioactivity of main fission products was derived after calculating the source term in containment. The data will be used for Daya Bay NPP PSA analysis

Energy Technology Data Exchange (ETDEWEB)

The information given in the present report about the Chernobyl accident and its radiological consequences indicates a serious crisis of the international radiation community. The following signs of this crises can be discerned: The international radiation community did not recognize the real reasons of the accident for a long time. It could not make a correct assessment of the damage to the thyroid of the affected populations of Belarus, Russia and the Ukraine. Up to present time it rejects the reliable data on hereditary malformations. It is not able to accept reliable data on the increase in the incidence in all categories of people affected by the Chernobyl accident. The international radiation community supported the Soviet authorities in their attempts to play down the radiological consequences of the Chernobyl accident for a long time. (author)

International Nuclear Information System (INIS)

Probabilistic risk assessment is a technique used to systematically analyze complex technical systems, such as nuclear waste management facilities, in order to identify and measure their public health, environmental, and economic risks. Probabilistic techniques have been utilized at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, to evaluate the probability of a catastrophic waste hoist accident. A probability model was developed to represent the hoisting system, and fault trees were constructed to identify potential sequences of events that could result in a hoist accident. Quantification of the fault trees using statistics compiled by the Mine Safety and Health Administration (MSHA) indicated that the annual probability of a catastrophic hoist accident at WIPP is less than one in 60 million. This result allowed classification of a catastrophic hoist accident as ''not credible'' at ...

Energy Technology Data Exchange (ETDEWEB)

The NUclear Power Engineering Corporation (NUPEC) has developed IMPACT-SAMPSON code to analyze integral behavior of light water nuclear power plants under severe accident conditions. IMPACT-SAMPSON's distinguishing features include interconnected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to severe accident events, and high-speed simulation on parallel processing computers. The integral plant behaviors of typical PWR and BWR under severe accident conditions have been analyzed with the IMPACT-SAMPSON code. The PWR plant analyzed was the three-loop, steel-dry containment type with 2,440 MWt. The AE accident scenario was supposed, that is, LOCA by 6-inch hot leg failure followed by accumulated water injection, but no ECCS and containment spray activation. The BWR plant analyzed was the 3,293 MWt BWR-5, Mark-II containment type. ...

International Nuclear Information System (INIS)

This paper describes the activities for improving the Probabilistic Safety Assessment (PSA) quality in the human reliability analysis (HRA) of the pre-accident human errors for the Korea Standard Nuclear Power Plant (KSNP). We evaluate the HRA results of the PSA for the KSNP and identify the items to be improved using the ASME PRA Standard. Evaluation results show that the ratio of items to be improved for pre-accident human errors is relatively high when compared with the ratio of those for post-accident human errors. They also show that more than 50% of the items to be improved for pre-accident human errors are related to the identification and screening analysis for them. In this paper, we develop the modeling guidelines for pre-accident human errors and apply them to the auxiliary feedwater system of the KSNP. Application results show that more than 50% of the items to be ...

Energy Technology Data Exchange (ETDEWEB)

Color formation in aqueous solutions of the ditetrazolium salt blue tetrazolium (BT{sup 2+}) in the absence or presence of oxygen is a complex radiation chemical reaction. The final stable product is the poorly soluble diformazan violet to blue pigment having a broad spectral absorption band ({lambda}{sub max}=552 nm). The reaction of BT{sup 2+} with the hydrated electron proceeds by rapid reduction of BT{sup 2+} followed by protonation at the nitrogen closest to the unsubstituted phenyl group, via the two intermediate tetrazolinyl radicals shared by the ditetrazole ring nitrogens. The effect of solution pH, N{sub 2}O saturation, and the presence of the reducing agent dextrose are examined. The system serves as a radiochromic sensor and a dosimeter of ionizing radiations. Solutions of 5 mmol l{sup -1} BT{sup 2+} at pH 7.3 serve as dosimeters over an absorbed-dose range of approximately 0.2-6 kGy (dearated, with a range of 1-8 mmol l{sup -1} ...

Energy Technology Data Exchange (ETDEWEB)

Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

International Nuclear Information System (INIS)

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to ...

International Nuclear Information System (INIS)

A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key ...

Energy Technology Data Exchange (ETDEWEB)

The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals ...

Science.gov (United States)

The potential public safety impacts from accidents in conceptual fusion power plants were investigated. Fusion was found to have some potential for accidents, as does any energy generating system. Functions of fusion power plants were identified that possess sufficient potential for an accidental release of toxic materials to the environment. An assessment was made of the impact of the potential accidents and recommendations are included for R and D that will allow incorporation of safety concerns in fusion power plant design. This work was based on a review of information available in conceptual design documents of fusion reactor systems.

International Nuclear Information System (INIS)

Death as a criterion (age distribution of occupational death; mean loss of life years due to radiation deaths); accidents at work (incidence of accidents of certain degrees of severity); total loss of working days due to accidents; occupational diseases; somatic and genetic radiation effects; radiation effects during pregnancy (incidence of pregnancies, ristes before implantation, hazards to the embryo, hazards to the foetus, total additional risk due to radiation exposure during pregnancy); age and sex dependence of risk figures; attempted formulation of an index of harm. (HP/orig.).

Energy Technology Data Exchange (ETDEWEB)

The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

UK PubMed Central (United Kingdom)

This paper attempts to clarify the most common measures available for the fire and electrical engineer in the prevention of polychlorinated biphenyl (PCB) hazards. It points out the risks and the potential...Full Text Available

Science.gov (United States)

... Except for isolated radiation accidents over the ensuing years, little practical experience has been gained in the treatment of thermal injuries ...

UK PubMed Central (United Kingdom)

During a high speed road traffic accident, a 26-year-old man suffered multiple fractures of his thoracic vertebrae and bilateral pneumothoraces. The day after admission and commencement of nasogastric...Full Text Available

International Nuclear Information System (INIS)

Korea Atomic Energy Research Institute, using the ANS Low Power /Shutdown (LPSD)PRA Standard, evaluated the LPSD PSA model of the KSNP, Younggwang (YGN) Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the pre-accident human errors in the LPSD PSA model of the KSNP showed that 13 items among 15 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for pre-accident human errors in the LPSD PSA model for the KSNP to improve its quality. We considered potential pre-accident human errors for all manual valves and control/instrumentation equipment of the systems modeled in the KSNP LPSD PSA model except reactor protection system/ engineering safety features actuation system. We reviewed 160 manual valves and 56 control/instrumentation equipment. The number of newly identified ...

Energy Technology Data Exchange (ETDEWEB)

The risk perception of the public consists of two components: the objectively factual component and the subjectively irrational component. The two strategies adopted by the German gas supply industry are the internal and the external communication strategy. Concepts and measures of accident precaution, registration and analysis of accident data (installation and operating errors, defects on flue systems, pipelines and valves, subsequent installation of gas appliances) are discussed. (R.P.)

Energy Technology Data Exchange (ETDEWEB)

The volume contains reports on plant technology, on systems organisation with the aim to minimize the risk (human error), on the problem of seismic risk, on core-disruptive accidents and on accident consequence models with different release categories and a comparison of the potential damage incurred. Mr. Webb; one of the authors, attempts to disprove the objections to his two earliest SNR statements by experts of Karlsruhe Nuclear Research Centre.

Science.gov (United States)

Only a small number of radiation accidents in the United States have been severe enough to result in multi-organ failure (MOF). Medical details of selected medical misadministration and criticality cases are reviewed, with an emphasis on pathophysiology. The four criticality cases are particularly relevant for analysis of MOF, since medical treatment was supportive and did not appreciably alter the clinical evolution of radiation injury. PMID:15975871

International Nuclear Information System (INIS)

Medico-biological aspects of Chernobyl accident among suffered children and adult population in Ukraine are exposed. Health condition of children irradiated in postnatal period and born from irradiated parents are described. Results of the most important organs and systems monitoring in different categories of suffered adults and data about non-stochastic and stochastic effects are given. Special attention is given to neuropsychiatric and endocrinological effects, conditions of visceral systems

International Nuclear Information System (INIS)

The objectives of this paper are to discuss the following aspects of physiopathology of iodine nutrition related to thyroid irradiation by nuclear accidents: (1) The cycle of iodine in nature, the dietary sources of iodine and the recommended dietary allowances for iodine. (2) The anomalies of thyroid metabolism induced by iodine deficiency. The caricatural situation as seen in endemic goitre will be used as mode. (3) The specific paediatric aspects of adaptation to iodine deficiency. (4) The present status of iodine nutrition in Europe. (author).

Energy Technology Data Exchange (ETDEWEB)

This book presents the papers given at a symposium on occupational safety in coal mines. Topics considered at the symposium included human factors, causes and prevention of personal injuries, remote sensing for ground control, respirable dust generation by continuous miners, accident analysis, hazard analysis of mining equipment, coal mine blasting accidents, coal mine respirable dust sampling, and noise in the mining industry.

International Nuclear Information System (INIS)

The Chernobyl reactor accident has pronupted all European countries to rehabilitate their existing measurement and monitoring systems and to design and erect new ones. These systems are meant to ensure a rapid overview on the situation in case of an accident to adopt suitable actions for protection or prevention. 6 papers report on the state of such measurement systems in Europe, inparticular those in France (TELERAY), in Germany (IMIS) and in Switzerland (RADAIR). The IMIS-system is discussed for its extension to Eastern Germany. (HP).

International Nuclear Information System (INIS)

Traditionally, probabilistic risk analyses [PRA] of severe accidents in nuclear power plants have limited themselves to consideration of the set of initiating events occurring during full power operation. However, some analyses of accident initiators during low power, shutdown, and other modes of plant operation other than full power have been performed. These studies as well as the Chernobyl accident and recent operating experience at U.S. pressurized water reactors suggested that risks during low power and shutdown could be significant. As such, the analysis of the frequencies, consequences, and risks of these accidents was identified as one task in the Nuclear Regulatory Commission staff's study of the implications of the Chernobyl accident to U.S. commercial nuclear power plants. The surry PRA project is an ongoing high priority effort at BNL [Brookhaven National Laboratory] ...

International Nuclear Information System (INIS)

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that ...

Energy Technology Data Exchange (ETDEWEB)

The dosimetry for radiotherapy involving neutrons is very complicated, owing to the complexity of secondary radiation components, whose contributions to the total absorbed dose have to be discriminated, owing to the different radiobiological effects. In order to separate thermal neutrons and photons, LiF dosimeters are mostly utilized. containing different percentage of Li, like as TLD-700, TLD-100 and TLD-600, but many problems arise. In the response of TLD-700 exposed to neutron-gamma mixed fields with high neutron flux, the contribution of thermal neutrons to the Tl emission is high. Moreover. TLD-100 and TLD-600 may undergo radiation damage, and great care has to be taken in order to obtain reliable results. Other TLDs showing lower sensitivity to neutrons are proposed and experimented for such high-flux neutron fields. The faced problems and various proposed solutions are here described. (Author)

Energy Technology Data Exchange (ETDEWEB)

Measurements of neutron and gamma-ray penetrations in an iron shield were performed up to a 60-cm depth in a tightly coupled source shield configuration with the fast-neutron reactor YAYOI as a source. Rates of neutron reactions and gamma-ray dose rates in the iron shield were obtained using activation foils and thermoluminescent dosimeters. Analyses of the experiments were made by using the DOT-III code with coupled neutron and gamma-ray cross sections from ENDF/B-IV and POPOP4 libraries. To obtain the source condition for the iron shield analyses, the calculated spectrum was adjusted to the measured reaction rates at the reactor shield boundary. The calculated neutron and gamma-ray distributions in the iron shield show fairly good agreement with the experiments. The effect of difference in Bondarenko-type self-shielding factors on the analyses of the iron shield is also shown.

Science.gov (United States)

The induced radioactivity in the construction materials of a Cockcroft-- Walton type neutron generator was measured. Major activation products (/sup 24/ Na, /sup 28/Al, /sup 56/Mn, /sup 64/Cu, /sup 65/Ni, /sup 69m/Zn, /sup 88/Rb /sup 91/Sr /sup 101/Mo, /sup 187/W/ and resulting doses are tabulated. Results show that the highest gamma activities would be observed in the fluorescent bulbs, copper pipe, aluminum lattice rod, and the aluminum pipe clamp. Thermoluminescent dosimeter readings yield the highest doses for the copper pipe tee, copper pipe, and aluminum lattice rod. Results of measuremerts of the neutron and gamma dose profiles of the facility are shown. However the indication is clearly that the tritium target, compared to other components, is the major source of radiation both during and after shutdown. (UK)

Energy Technology Data Exchange (ETDEWEB)

This study investigated dose to bone tissue in electron beam therapy. Measurements were made using films and thermoluminescent dosimeters in a polystyrene phantom containing bone inhomogeneity for 15-MeV, 12-MeV, and 9-MeV electron beams. An increase in dose of approximately 18%, 12%, and 11%, for the three electron energies respectively, relative to the dose in polystyrene, was found for bone material having an electron density (relative to water) of 1.73. Measurements were also made using films for 15- and 9-Mev electrons in a phantom with a mandibular bone and teeth. A dose enhancement in bone of approximately 10% and 7%, respectively, for the two energies was found in the phantom where the electron density of bone was about 1.60. These results suggest that injury to bone is possible in those clinical situations where high doses of electrons are used for therapy.

Science.gov (United States)

A Pulsed Radiation Dosimetry System designed to measure radiation produced by particle accelerators is described. The problems associated with total-dose measurement of irregular shaped pulses of ionizing radiation have been simplified. The system responds to extremely narrow pulses of charged particles or X-rays using a pin diode as the detector. Direct readout of dose in rads (Si) is displayed on a 3-1/2 digit digital panel meter. The system will operate in either the multiple-pulse or single-pulse mode. The multiple-pulse mode would be useful in monitoring or tuning a linear accelerator. In the single-pulse mode the system will automatically display total dose of a transient event in real time and hold that measurement indefinitely or until the next measurement is made. The system features an automatic reset in either mode.

Energy Technology Data Exchange (ETDEWEB)

The results of a scoping study of onsite radiation exposures which could take place in each of three types of postulated reactor accidents are presented. The accident types are (1) a fuel handling accident at a Mark III BWR; an interfacing system LOCA or V sequence at a PWR; and and Anticipated Transient Without Scram (ATWS) at a Mark I BWR. Both external and internal dose pathways are considered. The results of the study indicate the prohibitively high radiation doses could be received in some plant areas if personnel were to remain there. However, times of the order of a few minutes to a few hours, depending on the type of accident, would be available before life-threatening doses would be accumulated assuming that the provided full face respiratory protection equipment were used promptly. Special attention was given radiation doses possibly received by control room personnel for several control room ...

International Nuclear Information System (INIS)

Stable boiloff of core water during a severe LWR accident, that is, boiloff driven only by the decay power generated below the water level, is tractable analytically and is relatively insensitive to axial power distribution. As might be expected, calculated accident event times are sensitive to the fidelity of the decay power model. During later stages of boiloff, heat transfer or transport of energy from above the water level to the residual water can result in an unstable condition during which the boiloff rate increases greatly. The unstable boiloff phenomenon illustrates the highly nonlinear influence of core heat transfer during meltdown and emphasizes the great accuracy requirements which attend the modeling of the accident during periods of enhanced heat transfer when significant zirconium oxidation is possible.

British Library Electronic Table of Contents (United Kingdom)

Abstract:- Two motorcycle riders lost control of their vehicle, fell, and hit a guardrail, which acted as a blade and led to a rapid, fatal outcome. In one case, the high velocity of the body at the time of the impact resulted in complete detachment of the trunk. Reconstruction of the accident dynamics enabled the guardrail post to be identified as the means of injury in both cases. The two accidents occurred over a short period of time, highlighting a dangerous phenomenon that in less severe cases is presumably associated with different degrees of survivor disability. The accidents deserve mention, because a different design of the impact surface of the guardrail post might have prevented the lethal outcome. There is an urgent need for legislators to pass regulations that modify crash bar...

International Nuclear Information System (INIS)

SAMPSON is an integration of twelve analysis modules under the final development phase (phase-2) and will be capable of simulating hypothesized severe accidents in a nuclear power plant. One of these modules, the Molten Core Relocation Analysis (MCRA) module, simulates the relocation behavior of a molten core during a severe accident. MCRA models severe accident phenomena by using mechanistic formulations for multi-phase, multi-component, and multi-velocity field. As one of the verification studies of SAMPSON in Phase-1, the in-core phenomena of PHEBUS/FPT0 was analyzed with three modules, MCRA, fuel rod heat up analysis (FRHA) module, and the analysis control module (ACM) of SAMPSON. (author)

International Nuclear Information System (INIS)

In view of the persisting and understandable fear of parents with regard to radioactivity in the food of their babies as a consequence of the Chernobyl reactor accident, the Commission on Nutrition of the Deutsche Gesellschaft fuer Kinderheilkunde (German Society of Pediatrics) and the Strahlenschutzkommission have published a statement. According to this statement, the maximum permissible level of radioactivity in commercial baby food has been fixed by the EC to be 370 Bq/kg. The dietetic food industry itself has fixed a maximum for its products which is only a tenth of the radioactivity level permitted by the EC directive. The milk powders for infants tested since the reactor accident contained no measurable radioactivity or only very low amounts of Cs 134 or Cs 137, correspondung to a maximum of 25 Bq/kg in the product. Late damage to health is not to be expected. (orig./ECB).

International Nuclear Information System (INIS)

We have studied in detail the gamma radiation induced changes in the electrical properties of the (TeO_2)_0_._9 (In_2O_3)_0 _._1 thin films of different thicknesses, prepared by thermal evaporation in vacuum. The current-voltage characteristics for the as-deposited and exposed thin films were analysed to obtain current versus dose plots at different applied voltages. These plots clearly show that the current increases quite linearly with the radiation dose over a wide range and that the range of doses is higher for the thicker films. Beyond certain dose (a quantity dependent on the film thickness), however, the current has been observed to decrease. In order to understand the dose dependence of the current, we analysed the optical absorption spectra for the as-deposited and exposed thin films to obtain the dose dependences of the optical bandgap and energy width of band tails of the localized states. The increase of the current with the gamma radiation dose may be attributed partly to ...

UK PubMed Central (United Kingdom)

A patient is presented with a cervical spinal cord transection which occurred after a motor vehicle accident in which the air bag deployed and the seat belt was not in use. The patient had complete...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

This paper describes a functional slide rule that provides a readily usable ?in-hand? method for estimating nuclear criticality accident information from sliding graphs, thereby permitting (1) the rapid estimation of pertinent criticality accident information without laborious or sophisticated calculations in a nuclear criticality emergency situation, (2) the appraisal of potential fission yields and external personnel radiation exposures for facility safety analyses, and (3) a technical basis for emergency preparedness and training programs at nonreactor nuclear facilities. The slide rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, distance from the fission source, and time-after criticality accidents for five different critical systems. Another sliding graph permits the estimation of critical solution fission yields based upon fissile material ...

Energy Technology Data Exchange (ETDEWEB)

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models ...

UK PubMed Central (United Kingdom)

The objectives of this study were to evaluate the current fitness of an area ambulance service based in Belfast and to quantify the physiological demands of accident and emergency work. From a total...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

NUPEC has carried out hydrogen mixing tests to investigate hydrogen distribution behavior within a model containment and to provide a set of experimental data for validation of severe accident analysis codes.

UK PubMed Central (United Kingdom)

Blunt chest-wall trauma is common; however, resultant tricuspid valve rupture is rare and can be subtle in its presentation. Transthoracic echocardiography plays a key role in diagnosis.Herein,...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

This guidebook organizes the profusion of rules and regulations surrounding radiation protection into a single-volume reference. Employee and public protection, accident prevention, and emergency preparedness are included in this comprehensive coverage. Whenever possible, information is presented in convenient checklists, tables, or outlines that enable you to locate information quickly.

Energy Technology Data Exchange (ETDEWEB)

This article, published in issue 80 of 'l'ACROnique du nucleaire', aims to retrace the early steps in the consideration of the possibility of a nuclear accident in France, with the inclusion of 'non-institutional' participants and applying the lessons learned in Belarus in the contaminated territories around the Chernobyl nuclear power plant. After a review of the origin of the involvement of the Association pour le Controle de la Radioactivite dans l'Ouest (ACRO) in addressing post-accident issues alongside the populations living in an environment polluted by radioactivity, it discusses, from the critical viewpoint of an NGO, the context and the working method adopted for this examination. This is followed by some key elements of the programme and unresolved questions about the available body of knowledge which motivates research and about the method adopted for the work. The conclusion, ...

International Nuclear Information System (INIS)

The major feed-water line break accident is re-analyzed, which is based on Guangdong Daya Bay nuclear power station final safety analysis report, to justify the impacts of the decreasing of auxiliary feed-water flow rate on the safety margin in Daya Bay. The results showed that the accident analysis can meet the demands of acceptance criteria with the auxiliary feed-water flowrate decreasing from 45 m"3/h to 41.8 m"3/h, and enough safety margin is still retained

International Nuclear Information System (INIS)

An experimental technique is described which allows for parametric investigations of transient behavior of mobile core materials in a fuel bundle geometry. For the out-of-pile simulation of energy releases resulting from mild TOP- or LOF-accidents the exothermic reaction of an aluminium-oxide-thermite is used. Transient material relocation inside the test section is recorded by X-ray-cinematography. Results of some experiments recently performed close to conditions expected to be achieved during mild TOP-accidents are described in detail.

International Nuclear Information System (INIS)

Within the APROS simulation environment we have built a thermo-hydraulic model of a conceptual fusion power plant which is water cooled and uses lithium-lead for tritium breeding. For the safety assessment of this design we have studied an accident sequence which starts from a loss or coolant flow then leads to first wall breach and pressurisation of the vacuum vessel. Simulations have revealed strong pressure transients which can be alleviated by design changes. One goal is to verify the adequacy of the containment design: it remains intact at least 14 h without any mitigating efforts. Estimates for radioactive releases are obtained. (author)

Energy Technology Data Exchange (ETDEWEB)

This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power analyses performed on Grand Gulf.

International Nuclear Information System (INIS)

A semi-mechanistic model for calculating solid radionuclide release rates from bubbling pools of sodium was developed. The influence of particle spacial and size distributions on the decontamination of the releases was analysed and found significant. Decontamination factors are shown as a function of pool depth, bubbling characteristics and particle size distribution. The calculation of a decontamination factor for estimating the source term of large scale hypothetical core disruptive accidents is presented. The decontamination factor for a large scale accident was found to be two orders of magnitude greater than results obtained from small scale experiments conducted with uniform particle distributions. (orig.).

Energy Technology Data Exchange (ETDEWEB)

As part of a U.S. multi-regional pilot study conducted by the six Centers for Radiological Physics, 12-mm-diameter by 0.4-mm-thick CaSO/sub 4/:Dy Teflon-embedded discs were evaluated and used to measure patient entrance exposure on 60 average patients at 12 clinical centers. The discs were found to have adequate sensitivity, reproducibility and linearity up to 69.7 microC kg-1 (270 mR). The minimum measurable exposure was estimated as 0.4 microC kg-1 (1.5 mR). All responses were corrected for energy dependence, which varied +/- 20% from 1.7 to 6.5 mm Al half-value layer. Patient entrance exposure values ranged from 1.3 to 28 microC kg-1 (5 to 110 mR), with a median value of 5.2 microC kg-1 (20 mR). This value agreed with exposure measurements made on the chest radiography equipment using an ionization chamber and a phantom which simulated an average patient, and with published Nationwide Evaluation of X-Ray Trends (NEXT) data for the same period.

International Nuclear Information System (INIS)

Optically stimulated luminescence (OSL) and thermoluminescence are observed in polycrystalline RbMgF3:Eu2+ after x-ray, ?-ray, or ? irradiation. The main electron traps are F-centers but there are other unidentified traps. The main hole traps at room temperature are probably Eu3+ and thermal or optical stimulation leads to electron-hole recombination at the Eu3+ site and Eu2+ emissions arising from 6PJ to 8S7/2 and 4f5d(Eg) to 8S7/2 transitions. We find that some of the electron traps can be emptied by infrared stimulation and all of the electron traps can be emptied by white light stimulation. The OSL dark decay is long and exceeds 5 days for traps that are emptied by white light stimulation after initial infrared bleaching. Our results show that this compound can be used as a radiation dosimeter for intermediate dose levels where the 87Rb self-dose does not significantly affect the dose reading.

British Library Electronic Table of Contents (United Kingdom)

In this work, thermoluminescent dosimeters have been used to measure entrance surface doses (ESDs) of patients undergoing mammographic screening at the University College Hospital, Ibadan, Oyo state. The mean glandular doses (MGDs) were also calculated using the measured ESDs. The results showed that the ESDs ranged from 0.26mGy to 21.26mGy for the mediolateral oblique (MLO) views and 0.08mGy to 5.36mGy for the craniocaudal (CC) views. The calculated MGD ranged from 0.07mGy to 3.57mGy for the MLO views and 0.02mGy to 0.98mGy for the CC views. The possible reasons for the large variations in the individual ESD values and MGD values for both views are discussed using patients data, equipment specific data and the technical parameters used for the examinations. Comparison showed that the mean...

International Nuclear Information System (INIS)

Tl(InS_2)_1_-_x(FeSe_2)_x single crystals (where x=0; 0.001; 0.005; 0.01 and 0.015) were grown by the Bridgman-Stockbarger method. Obtained single crystals were crystallized in monoclinic structure. The present paper deals with experimental results relative to X-ray dosimetric characteristics of the Tl(InS_2)_1_-_x(FeSe_2)_x solid solutions at 300 K. Installation URS-55a was the source of radiation. X-ray radiation dose (E) falling on the crystals is measured by the crystalline X-ray dosimeter DRGZ-02. The value of X-ray conductivity coefficient K_#sigma# characterising X-ray sensitivity is defined as K_#sigma#= (#sigma#_E-#sigma#_0)/E#sigma#_0. where #sigma#_E is conductivity under the effect of X-ray radiation by intensity E, #sigma#_0 is conductivity in the lack of radiation. Comparing X-ray dosimetric characteristics of TlInS_2 and Tl(InS_2)_1_-_x(FeSe_2)_x crystals notes that because of partial substitution of [InS_2]- anion radical by [FeSe_2] there have been ...

Energy Technology Data Exchange (ETDEWEB)

This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC`s Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube ...

International Nuclear Information System (INIS)

In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, ...

Energy Technology Data Exchange (ETDEWEB)

On March 28, 1979, the Pennsylvania Three Mile Island nuclear power plant Unit 2 (TMI-2) suffered a partial fuel-melt accident. During this accident, over 20,000 lb of molten fuel flowed through holes melted through the baffle plates and through the lower-core support assembly (LCSA). The molten fuel subsequently resolidified in the bottom of the reactor vessel. The lower-core support assembly of the TMI-2 reactor was not structurally damaged during the accident. In order to permit defueling of that region of the core, the LCSA was cut to permit access. A five-axis teleoperator was developed to deliver plasma arc cutting, rotary grinding and abrasive waterjet cutting of end effectors to the LCSA. Complex geometry sectioning was completed in a mock-up facility at chemistry and pressure conditions simulating those of the vessel, prior to actual in-vessel operations. In-vessel activities began in early May 1988 and were ...

Energy Technology Data Exchange (ETDEWEB)

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power Plants.`` The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different ...

International Nuclear Information System (INIS)

During the past 2 yr, significant progress has been made in several areas related to the safety and environmental (S and E) aspects of inertial fusion energy (IFE). An updated methodology has been developed, and accident analyses have been performed for two IFE conceptual power plants and a target fabrication facility. Parallel to the consequence analyses of different accident scenarios, ongoing studies of accident initiating events are being used to support safety assessment and create a basic framework of types of events to consider in future risk characterization of new plant designs. Target designers/fabrication specialists have been provided with ranking information related to the S and E characteristics of candidate target materials. We have revisited waste management options for IFE, introducing the concept of clearance versus the traditional shallow land burial. A brief summary of results in each of these activities ...

Science.gov (United States)

The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel-double-steel-plate structure similar to a hull structure, which contains coolant between the inner and outer walls to absorb the heat transferred from ...

International Nuclear Information System (INIS)

In the VEGA program on radionuclide release from irradiated fuel under severe accident conditions, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed the increase in release rate above 2,800 K or at the fuel liquefaction and the decrease in release rate under elevated pressure, which was a first observation in the world. The data on low-volatility radionuclide release, release from MOX fuel, effect of fuel oxidation, and eutectic reaction with cladding on release were obtained from the tests. The mechanism of pressure effect on release was examined and a new release model with pressure effect was proposed. In addition, the pressure effect on source term evaluation and effectiveness of accident management measures were investigated. This article summarizes the major outcomes described above that have already been published and ...

Energy Technology Data Exchange (ETDEWEB)

This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. Unlike the Facility hazard classification documented in WHC-SD-NR-HC-004, Hazard Classification for 300 Area N Reactor Fuel Fabrication and Storage Facility, (Huang 1993), which is based on unmitigated consequences, credit is taken in an HA for administrative controls or engineered safety features planned or in place. The HA is the foundation for the accident analysis. The significant event scenarios identified by this HA will be further evaluated in a ...

Energy Technology Data Exchange (ETDEWEB)

The Feedwater System supplies feedwater to the steam generator at the required pressure, temperature and flow rate during the plant start-up, normal power operation, shutdown. When the Feedwater System is inoperable or unavailable, the Auxiliary Feedwater System supplies emergency feedwater to the steam generator. If main steam line break occurs, the increase of feedwater flow rate of the faulted steam generator due to decrease of the pressure in the faulted steam generator results in adverse effects in aspect of overcooling the Reactor Coolant System and increased containment pressure/temperature. To optimize the containment mass/energy analysis, this paper evaluates the maximum feedwater and auxiliary feedwater flow rate delivered to the faulted steam generator at each stage of pressure decrease in the faulted steam generator after a main steam line break accident. Calculated Feedwater flows are applied to calculate mass and energy release following MSLB ...

International Nuclear Information System (INIS)

Plasma surface interactions in tokamaks are known to create significant quantities of dust, which settles onto surfaces and accumulates in the vacuum vessel. In ITER, a loss of vacuum accident may result in the release of dust which will be radioactive and/or toxic, and provides increased surface area for chemical reactions or dust explosion. A new method of analysis has been developed for modeling dust resuspension and transport in loss of vacuum accidents. The aerosol dynamic equation is solved via the user defined scalar (UDS) capability in the commercial CFD code Fluent. Fluent solves up to 50 generic transport equations for user defined scalars, and allows customization of terms in these equations through user defined functions (UDF). This allows calculation of diffusion coefficients based on local flow properties, inclusion of body forces such as gravity and thermophoresis in the convection term, and user defined source terms. The code ...

International Nuclear Information System (INIS)

The program package COSYMA for assessing the radiological and economic consequences of nuclear accidents, developed with the support of the European Commission, was applied to investigate the health effects and risks from accidental releases of radioactive material from the Daya Bay nuclear power plant. Population distribution data in the range of 80 km around the site and hourly meteorological data for the year 1985 representative of accident consequence analysis were used. The results showed that early effects are more important at distances closer to the site, while the number of fatal cancers is closely related to the population density and the late effects are still important at distances larger than 50 km from the site. The mean annual expected values for early mortality and late mortality estimated for the population within a circle of 80 km around the Daya Bay nuclear power plant are 4.5x10"-"3 and 0.1 yr"-"1, respectively.

Energy Technology Data Exchange (ETDEWEB)

Contamination of certain areas of Europe with radiocaesium from the Chernobyl accident led to a higher {sup 137}Cs accumulation (i.e. 300-600 Bq kg{sup -1}) in grain and to potential post-accident contamination of broiler chickens. In future, such contamination may require a simple determination of the {sup 137}Cs activity concentration in broiler chicken meat which would lead to measures for preventing the recommended limits of radionuclide contamination of the meat for human consumption from being exceeded. This paper describes the development of a rapid method for the in vivo monitoring of the broiler chicken using a lead-shielded sodium iodide detector. The method enables simply fixed live chicken to be monitored, the results showing a good correlation (R{sup 2}=0.98) with measurements of meat from chicken previously monitored in vivo prior to slaughter.

International Nuclear Information System (INIS)

This standard provides requirements and guidelines for installation, inspection and testing activities that assure the quality of important mechanical parts of a nuclear power plant not covered by the ASME Boiler and Pressure Vessel Code, Section III, during construction. These parts include those mechanical systems and components whose satisfactory performance is required: for the plant to operate reliably; to prevent accidents that could cause undue risk to the health and safety of the public; or to mitigate the consequences of such accidents if they were to occur. The requirements of this standard deal with the protection and control necessary to assure that the requisite quality of those important parts of the plant are preserved from the time items are removed from storage or receiving until they are incorporated into the plant up to but not including fuel loading for PWR plants and the completion of cold functional testing for BWR and ...

Energy Technology Data Exchange (ETDEWEB)

The probability of damage and the potential resulting hazards are analyzed for a representative rail shipping container for three severe rail accident scenarios. The scenarios are: (1) the rupture of closure bolts and resulting opening of closure lid due to a severe impact, (2) the puncture of container by an impacting rail-car coupler, and (3) the yielding of container due to side impact on a rigid uneven surface. The analysis results indicate that scenario 2 is a physically unreasonable event while the probabilities of a significant loss of containment in scenarios 1 and 3 are extremely small. Before assessing the potential risk for the last two scenarios, the uncertainties in predicting complex phenomena for rare, high- consequence hazards needs to be addressed using a rigorous methodology.

International Nuclear Information System (INIS)

Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab.

International Nuclear Information System (INIS)

Dr. Rudolf Schulten of West Germany and expert witnesses from national laboratories, utilities, and the nuclear industry testified on reactor safety issues as they relate to the Chernobyl accident and public concern that modern technology has not paid enough attention to public safety. Each of the witnesses contributed safety-related information based on what has been learned from the Soviet incident. Particular focus went to similarities and differences between the Chernobyl and US reactors in safety design and engineering and to the environmental effects of the accident. The N reactor near Richland and a commercial reactor at Fort St. Vrain, Colorado are the only two operating graphite reactors, but neither is a boiling water reactor.

International Nuclear Information System (INIS)

The present paper deals with the experimental determination of pressure drop across a four-cusped vertical channel. This geometry represents, ideally, the blockage condition in a typical pressurized water reactor with core degraded by accident. Experiments were performed for both single and two-phase flow. Water was utilized for the single-phase measurements whilst simultaneous flow of air and water simulated the steam-water flow. Observation of the prevailing two-phase flow regime was carried out, so that its mechanism could be fully understood. The averaged void fraction was also measured, by the gamma-ray attenuation technique. A wide range of water and air mass flow rates was covered, so that all flow conditions, possible to exist in a reactor with LOCA, could be investigated. New correlations for pressure drop are proposed. (Author).

International Nuclear Information System (INIS)

Integrated Leakage Rate Test (ILRT) of containment system plays a very important role in safety of a Nuclear Power Plant. Containment system constitutes the last physical barrier to release of radioactivity from the core and is called upon to mitigate the consequences of not only accidents within the design basis, but also some of the highly unlikely severe accidents. Hence, leak tightness of containment becomes uttermost priority for the safety of plant personnel and public. The containment and associated ESFs are tested before the first criticality and there after periodically during service. The pre-operational integrated leakage rate is carried out at LOCA based design pressure, at periodic test pressure and at some intermediate pressure points to assess the leakage characteristics. This paper summarizes the various requirements and activities relevant to the ILRT of the Indian Pressurized Heavy Water Reactor (PHWR) containment system. ...

Energy Technology Data Exchange (ETDEWEB)

Possible accident scenarios for a fusion reactor include breaches in the vacuum or cooling system. Intruding air or steam could react with structural or plasma facing materials, possibly mobilizing radioactive isotopes. Safety assessments must consider the early dose at the site boundary from the release of these activated materials. Previous calculations have indicated that cobalt isotopes dominate dose calculations for designs using stainless steel. Values used in these calculations, however, had been largely determined by the measurement limits of the chemical analysis methodology instead of measured releases. The purpose of the current study was to refine the analytical method to reduce the limit for detecting cobalt, and then test PCA steel in air and steam between 973 and 1473 K. Goals were to obtain more accurate measurements of cobalt mobilization in terms of g/m{sup 2}{center_dot}h and insight into the mobilization mechanisms.

Energy Technology Data Exchange (ETDEWEB)

This association participated in different working groups: North Cotentin radioecology group, groups of expertise on the uranium mines of Limousin, executive committee for the management of the post accidental phase of a nuclear accident or a radiological emergency situation, radioactive waste management, radiological surveillance of the territory, radiation protection mission by the Asn, radiological surveillance of the environment of the Chinon nuclear power plant, study of the presence of {sup 235}U around the site of Brennilis, study of the radioactive waste management at the Manche plant, radiological surveillance of the Cyceron cyclotron at Caen, Aurengo commission on the consequences in France of the Chernobylsk accident. Actions of information, regular publications, meeting with public are also a part of the work of this association. (N.C.)

International Nuclear Information System (INIS)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

International Nuclear Information System (INIS)

The possibility of using Direct Flow Control (DFC) to avoid catastrophic accidents due to containment breaches in chemical plant is discussed. Recommendations are made for locating fluidic elements, and the effectiveness of simple DFC protection is analysed. More powerful methods of protection are outlined using spin diversion and the complementary properties of fluidic and conventional valves are exploited. (author).

International Nuclear Information System (INIS)

The petroleum and petroleum products increase offer a real development opportunity to the alternative fuels. In the context of the french energy accounting increase, the energy independence notion incites the government to promote these new fuels. If the LPG seems declining because of the accident risks fear, the fuel cell is not for today. Near these two sectors what is the future of the biofuels and the natural gas vehicle or the electric cars? (A.L.B.)

Energy Technology Data Exchange (ETDEWEB)

This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not ...

International Nuclear Information System (INIS)

'Small radioactive source results in big accident' have occurred repeatedly in China and worldwide alike. Radioactive source management is one of the key activities for a nuclear power plant to maintain its good safety record and image to the public. From aspects of establishing the management system, centralized storage, periodic accounting, performing whole process control to the source usage and experience feedback etc., the author reports the practice and experience of radioactive source management in Daya Bay Nuclear Power Plant

International Nuclear Information System (INIS)

The study of the melting of combustible rods is of great importance for the safety analysis of nuclear reactors. Due to the special characteristics of the problem, a sharp interface between the solid and liquid region does not exist, but appears a 'mushy' region in which the material is partially melted. The Finite Element Method is employed here, together with a regularized enthalpy formulation. Finally, the results obtained are presented and discussed. (Author).

International Nuclear Information System (INIS)

Since 1945, more than 1.8 x 10"2"1 Bq of artificial radionuclides have been released into the atmosphere. Approximately 2.04 x 10"1"8B, i.e. approx. 0.11%, are the result of accidents at nuclear industrial facilities. This percentage is causing increased interest among researchers. This is due to the fact that in the wake of accidental release radionuclides become distributed unevenly across the Earth's surface, and the associated exposures, fluctuating from background level to several grays, an induce both stochastic and deterministic effects in the irradiated population. A comparative analysis of the medical consequences of the twentieth century's most serious nuclear events, namely the authorized dumping of high level radioactive waste into the river Techa in 1950, the explosion of a storage tank containing long lived radioactive waste in the Southern Urals in 1957, the fire at Sellafield in 1957 and the accident at the Chernobyl nuclear ...

International Nuclear Information System (INIS)

In the case of a release of residual power and fragmenting following a hypothetical accident the applied powers are small. The boiling in the fluid in the bed promotes leveling and the angles of repose obtained are very small. For a specific power in water of 3.1 W/cm_3 a limiting angle of repose of less than 2 degrees is obtained after a time interval of between 1 and 3 hours. EDULCOREE-and ETABUL-research programs are carried out. (DG).

Energy Technology Data Exchange (ETDEWEB)

Following a postulated meltdown accident, the integrity of containment building structural material under attack by hot molten core debris and the safeguard of environment against radiological releases constitutes the final line of defense in PAHR safety assessment. Such assessment requires a good knowledge of UO/sub 2//interaction and penetration with different types of concrete. The present study focuses on the phenomena associated with core debris interaction/penetration with substrate basalt concrete.

International Nuclear Information System (INIS)

Objective: To develop a computerized method of internal dose estimation on radiation protection. Methods: Based on MIRD mathematic model of the organs and by means of the programming language of MS Visual Basic 6.0, a computer program of dose estimation in internal radiation was developed for radiation protection. Results: The computerized method of dose estimation for internal radiation was completed. Conclusions: This computerized method is very convenient for internal radiation dose estimation of several aspects. It can also be used in radiation accident. (authors)

UK PubMed Central (United Kingdom)

A 21-year-old male patient was transferred to the emergency room of our hospital after suffering seat belt abdominal injury in a traffic accident. Abdominal computed tomography revealed a massive hematoma...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the 'ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's ...

Energy Technology Data Exchange (ETDEWEB)

Background: In an effort to standardise radiological practices in the Republic of Ireland, current legislation states that 'written protocols for every type of standard radiological practice shall be established'. In order to fulfil this requirement the Irish Medical Council recommends the protocols issued by the Commission of European Communitees (CEC) for adoption in the country. Whilst this document does provide good guidance with regard to various radiographic factors, patient shielding instructions are notably ambiguous. The aim of this study was to remove some of this ambiguity by defining the optimal method of positioning patient shielding in antero-posterior (AP) and lateral lumbar spine radiographic examinations. These projections were chosen on the basis of their area of coverage being in close to and in some cases including the reproductive organs. They also represent the highest source of collective population dose of any conventional radiographic ...

International Nuclear Information System (INIS)

The risks of genetic, leukemia and malignant diseases from medical X-ray diagnostic examinations were estimated using the frequency of radiographic and fluoroscopic exposures per diagnostic examination, child expectancy, leukemia and malignancy significant factors, and using a weighting factor determined on the basis of data concerning the cancer mortality among atomic bomb survivors in Nagasaki and of a recommendation of International Commission of Radiological Protection. The organ or tissue doses with respect to the stochastic risks were determined with ionization chambers and thermoluminescent dosimeters placed at the positions of the organs or tissues in a RANDO woman phantom which was exposed to diagnostic X-rays according to technical factors of typical radiographic and fluoroscopic examinations obtained from a nationwide survey. The resultant risks by age-group and type of radiographic and fluoroscopic examination are tabulated in terms of risk level of ...

Energy Technology Data Exchange (ETDEWEB)

Background: Despite the fact that adverse health effects are not confirmed for exposure to radiofrequency electromagnetic field (RFEMF) levels below the limit values, as defined in the guidelines of the International Commission on Non-Ionizing Radiation Protection, many persons are worried about possible adverse health effects caused by the RF-EMF emitted from mobile phone base stations, or they attribute their unspecific health complaints like headache or sleep disturbances to these fields. Method: In the framework of a cross-sectional study a questionnaire was sent to 4150 persons living in predominantly urban areas. Participants were asked whether base stations affected their health. Health complaints were measured with standardized health questionnaires for sleep disturbances, headache, health complaints and mental and physical health. 3,526 persons responded (85%) to the questionnaire and 1,808 (51%) agreed to dosimetric measurements in their flats. Exposure was measured in 1,500 ...

International Nuclear Information System (INIS)

In this work, thermoluminescent dosimeters have been used to measure entrance surface doses (ESDs) of patients undergoing mammographic screening at the University College Hospital, Ibadan, Oyo state. The mean glandular doses (MGDs) were also calculated using the measured ESDs. The results showed that the ESDs ranged from 0.26 mGy to 21.26 mGy for the mediolateral oblique (MLO) views and 0.08 mGy to 5.36 mGy for the craniocaudal (CC) views. The calculated MGD ranged from 0.07 mGy to 3.57 mGy for the MLO views and 0.02 mGy to 0.98 mGy for the CC views. The possible reasons for the large variations in the individual ESD values and MGD values for both views are discussed using patients' data, equipment specific data and the technical parameters used for the examinations. Comparison showed that the mean ESD values and MGD values reported in this work are below published values. The mean of the calculated MGD values is also found to be lower than the recommended guidance ...

International Nuclear Information System (INIS)

The Radiation Technology Center from IPEN-CNEN/SP, Brazil, developed a revolutionary design and national technology, a small-sized continuous run and multipurpose industrial gamma irradiator, to be used as a demonstration facility for manufacturers and contract service companies, which need economical and logistical in-house irradiation system alternatives. Also, to be useful for supporting the local scientific community on development of products and process using gamma radiation, assisting the traditional and potential users on process validation, training and qualification of operators and radioprotection officers. The developed technology for this facility consists of a continuous tote box transport system, comprising a single concrete vault, where the automated transport system of products inside and outside of the irradiator utilizes a rotating door, integrated with the shielding, avoiding the traditional maze configuration. Covering 76 m"2 of floor area, the irradiator design is ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this paper is to perform an impact analysis of the Interim Storage Cask (ISC) of the Fast Flux Test Facility (FFTF) for a 4-ft end drop. The ISC is a concrete cask used to store spent nuclear fuels. The analysis is to justify the impact force calculated by General Atomics (General Atomics, 1994) using the ILMOD computer code. ILMOD determines the maximum force developed by the concrete crushing which occurs when the drop energy has been absorbed. The maximum force, multiplied by the dynamic load factor (DLF), was used to determine the maximum g-level on the cask during a 4-ft end drop accident onto the heavily reinforced FFTF Reactor Service Building`s concrete surface. For the analysis, this surface was assumed to be unyielding and the cask absorbed all the drop energy. This conservative assumption simplified the modeling used to qualify the cask`s structural integrity for this accident condition.

Energy Technology Data Exchange (ETDEWEB)

Condensing two-phase channel flow occurs in many industrial applications, including heating and refrigeration systems. It can also occur in certain nuclear reactor accidents. For example, during a small-break loss-of-coolant accident in a pressurized water reactor, following the partial depletion of the primary coolant, condensation of steam on the primary side of the steam generator tubes can provide a heat sink for disposal of the decay heat generated in the reactor core. Condensing two-phase flow can also play an important role in the operation of the passive emergency cooling system in the advanced simplified boiling water reactor. Here, steady-state condensation in the presence of a noncondensable in a concurrent two-phase channel flow is analyzed using a two-fluid model. The effect of noncondensables on the combined heat transfer at the liquid-gas mixture interphase is accounted for by using the stagnant film model, and closure relations ...

International Nuclear Information System (INIS)

The Belgian Tihange 2 nuclear power plant went into commercial operation in 1983 producing a thermal power of 2785 MW. Since the commissioning of the plant the steam generators U-tubes have been affected by primary stress corrosion cracking. In order to avoid further degradation of the performance and an increase in repair costs, Electrabel, the owner of the plant, decided in 1997 to replace the 3 steam generators. This decision was supported by the feasibility study performed by Tractebel Energy Engineering which demonstrated that an increase of 10% of the initial power together with a fuel cycle length of 18 months was achieved. Tractebel Energy Engineering was entrusted by Electrabel as the owner's engineer to manage the project. This paper presents the role of Tractebel Energy Engineering in this project and the safety analysis program necessary to justify the new operation point and the fuel cycle extension to 18 months re-analysis of FSAR chapter 15 accidents ...

Energy Technology Data Exchange (ETDEWEB)

The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for ...

International Nuclear Information System (INIS)

A featfinding committee of the German Federal Parliament in July 1980 recommended to perform a ''risk-oriented study'' of the SNR-300, the German 300 MW fast breeder prototype reactor being under construction in Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the risk analysis six groups of accidents have been identified which may initiate core destruction. These groups comprise all conceivable courses, potentially leading to core destruction. By reliability analyses, expected frequency of each group has been calculated. In the accident analysis potential failure modes of the reactor tank have been investigated. Core destruction may be accompanied by the release of significant amounts of mechanical energy. The primary coolant ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station 3 and 4 is a 540 MWe Pressurized Heavy Water Reactor. It uses 37 - element natural Uranium dioxide (UO_2) fuel pellets encapsulated inside the cylindrical sheath and are welded to the end plate at each end. During an postulated accident in which part of the fuel bundle are exposed to very high temperature (no means of heat removal) and other are at lower temperature (coolant temperature) possibility of failure of end plate weld due of thermal stresses developed by these relative temperature cannot be ruled out. In this report an attempt is made to study behaviour of fuel bundle under different temperature loading. Modelling of 37 element fuel bundle was done in ANSYS FEM. System was analysed for various sets of temperature loading. The system was analysed for plasticity and creep as material nonlinearity. The total strain, creep strain and stress increase as the temperature increases in upper portion of fuel bundle and decrease with ...

Energy Technology Data Exchange (ETDEWEB)

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, ...

Energy Technology Data Exchange (ETDEWEB)

This report presents the thermal-hydraulic analysis of four Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the alternative SEAFP reactor design. The LOFAs considered result from a loss of electrical power for the recirculation pump in the primary cooling circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall and blanket. For the LOFA without plasma shutdown, significant loss of heat removal due to dryout occurs at the midplane of the outboard first wall cooling pipes about 41 s after pump trip. For the three LOFA cases with emergency plasma shutdown that have been studied, the temperature increase in the Be-coating at the midplane of the outboard first wall is limited to about 30 K. (orig.).

Energy Technology Data Exchange (ETDEWEB)

Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: GDH impact on an adjacent GDH and its attached piping; and GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results ...

International Nuclear Information System (INIS)

The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded ...

International Nuclear Information System (INIS)

A detailed outline of the application of the MELCOR and RELAP5/MOD3.1 codes to the analysis of the thermohydraulic response and determination of other parameters of the medium on the floor is given for several classes of secondary coolant circuit accidents along with the description of the related facilities. An overview is presented of the maximum values and time behavior of the thermohydraulic parameters, pressure, temperature, relative humidity, and water level on the floor. Transverse rupture of the steam generator, main steam header, or main feedwater header piping during normal operation is considered as the initiating event. Pressure is only 10% higher than the atmospheric pressure. Air temperature attains a value as high as 100 degC. Relative humidity is 100%, persisting as long as the steam source is available. The water level is typically about 8 cm and never exceeds 15 cm. (M.D.). 16 tabs., 37 figs., 32 refs.

Energy Technology Data Exchange (ETDEWEB)

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for evaluation of debris in-vessel retention. The debris coolability analysis module is developed for the accurate prediction of the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with a temperature distribution and failure evaluation of the vessel wall. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis. The calculation for spreading is compared with a water spreading experiment on the floor and the calculation for coolability is compared with a n-octadecane melting experiment in a rectangular vessel. The comparisons show capabilities for predictions of spearhead transportation in the debris spreading process and of melting front transportation ...

Energy Technology Data Exchange (ETDEWEB)

The US Department of Energy, Division of Electric Energy Systems, has undertaken to identify the environmental issues and potential impacts associated with the installation of underground electric power transmission systems. This study reports the results of investigations into the advanced cable technologies being considered for future underground applications, as part of the development oriented research program of the Division of Electric Energy Systems. While the technology involves a high level of sophistication, there are relatively few impacts to the environment that are potentially significant, and of these none are inherently non-mitigable. Route planning, system design, and methods of construction and accident response can be pursued in order to minimize impacts where strict constraints are appropriate.

Energy Technology Data Exchange (ETDEWEB)

FLOWTRAN-TF is a new two-component, two-phase thermal-hydraulics code to capture the detailed assembly behavior associated with loss-of-coolant accident analyses in multichannel assemblies of the SRS reactors. The local interfacial area of the two-phase mixture is computed by summing the interfacial areas contributed by each of three flow regimes. For smooth flow regime transitions, the code uses an interpolation technique in terms of component void fraction for each basic flow regime.

International Nuclear Information System (INIS)

Object: To provide a tank, which can safely and reliably accommodate high temperature sodium containing radioactive substance in case of occurrence of an accident in a sodium system and thus prevent spread of contamination. Structure: A sodium drain duct inserted into a tank from above the tank is provided at the position of its lower end with a buffer means for preventing direct flow-down of sodium to a bottom plate. A means for preventing the discharge of radioactive substance to the cover gas is provided above the lower end of the sodium drain tube so as to surround the sodium drain tube. (Kamimura, M.).

Wastenet

... Thursday: South Brisbane Sailing Club, Orleigh Park Hill End Terrace, West End. Friday: Booroodabin Bowls Club, 126 Breakfast Creek Road, Newstead. Related Photos Six months on: Kerrin Quinn's Story (Emma Sykes - ABC Local) Map Fernvale 4306 Subscribe/RSS Subscribe to ABC Brisbane videos Subscribe to all ABC Local videos Topics: disasters-and-accidents, floods Locations: brisbane-4000, fernvale-4306 Print page ...

International Nuclear Information System (INIS)

A conceptual framework is proposed within which the notion of risk as normally used in risk assessment (RA) could be enlarged in line with the real substance of social issues of technology policy, to help avoid RA's threatened irrelevance to social decision making. It is argued that the frequent organizational incoherence and thus the unviability of modern technology arises from 'social alienation' between the innovation-commitment phase and the implementation of the technology in society. The roles of technical elites and of particular concepts of technology in this alienation are emphasized. One of the case studies deals with 'Nuclear power - myths of scientific and organizational realism' and discusses the UK nuclear 'programme' and the Three Mile Island accident. (author).

International Nuclear Information System (INIS)

The second international conference on containment design and operation included sessions on the following topics: performance and regulatory requirements; radionuclide behaviour; severe accident design and analysis; operation, maintenance, leaking and aging of containment systems; thermal hydraulic behaviour of containment systems; hydrogen mixing and mitigation; design methods and concepts; code validation; structural analysis and response tests; passive safety systems; aerosol behaviour; containment reliability, integrity, and risk assessment; hydrogen deflagration and detonation. Due prominence was given to CANDU and other PHWR reactors. The individual papers have been abstracted separately.

International Nuclear Information System (INIS)

The analysis for priority rankings of the recommendations to reduce the total core damage frequency (CDF) of Wolsong nuclear power plant units 2/3/4 was performed in this paper. In order to derive the recommendations, the sensitivity analysis of CDF on which major contributors effect was performed based on the accident quantification results during Level 1 probabilistic safety assessments (PSA). Priorities were ranked in the way that compares the CDF reduction rate with the efforts required to implement those recommendations using risk matrix.

International Nuclear Information System (INIS)

The validation of the forecast model for early emergency response to nuclear accidents is evaluated by trace tests in atmosphere in Daya Bay nuclear power site. The simulation experiment of the Daya Bay nuclear power site shows that the particle spreading image and the time-integrated concentration distribution given by plume concentration prediction model can perform the variation of pathway of the pollutant transport, as well as the effects of topography on transport and diffusion of pollutants. The simulation of five trace tests in field shows that 59.1% of ratios between predicted results and observed results are within the range of 10, and 41% of ratios are within the range of 5 approximately. (authors)

Energy Technology Data Exchange (ETDEWEB)

The contribution under consideration spans a wide curve from the first deadly accident of motorized aviation in the year 1908 up to newer tendencies of the so-called team resources management on observation points and control stations in order to pursue fundamental questions in man-machine systems. A continuous differentiation in the consideration of the factor human beings in complex technical systems is described. This is illustrated by the example of concepts for leadership, communication and decision making on observation points and control stations.

Energy Technology Data Exchange (ETDEWEB)

The author discusses in detail the following topics: Compensation for domestic nuclear damage and for transfrontier nuclear damage - rule of formal equality of parties which belongs to the basic rule of civil law considering the position of domestic and foreign victims of a grave accident-juridical consequences of the preponderant role played by the state in the promotion, development and supervision of the nuclear industry-rationale for applying the concept of global limitation of liability in the law on nuclear liability and compensation - financial consequences of uncompensated nuclear damage, borne by the victims directly affected or spread over the whole community of the affected state? (HP)

Wastenet

...soy, rights, justice, contamination, water, cargill, port soy, rights, justice, contamination, water, cargill, port Friends of the ... The global food giant Cargill has built its own port on the banks of the River Paraguay with plans to expand. It's ...allows Puerto Union, the port facility belonging to the transnational food giant Cargill, to continue operating. This decree was issued despite the ... The Cargill port facility represents a hazard to the water supply of the entire population of the city, and any accident such ...

International Nuclear Information System (INIS)

In this paper a comparison between calculations of severe accidents occurred from WWER-1000 with ASTEC code specified for an event of full unloading with relief valves stuck opened with no hydroaccumulators intervention is presented. The purpose of the analyses provided is to present the relationship between the improvements of the actual version (ASTEC Vl.2 rev. 1) and ASTEC V1.1 p2 like: code modifications, incoming data improvements. Such discrepancies are to be examined. Case by case suggestions for ASTEC improvements are to be provided.

International Nuclear Information System (INIS)

It has been issued that ROP(Regional Overpower Protection) for Wolsong Unit 1 needed to be reanalyzed due to the aging effect. Thermo-hydraulics and core simulation have to be performed for calculation of the fuel bundle power, channel power and detector signal production. PPV/MULTICELL/RFSP code system was used to calculate the power distribution for the ROP analysis. In this study, 232 cases out of 926 scenarios which include postulated accidents such as Startup after Short Shutdown, Shim Cases, Stepback, Insertion and Withdrawal of Reactivity Control Rods were simulated.

International Nuclear Information System (INIS)

This paper gives a summary of a risk study assuming unprotected loss of flow (ULOF) in the SNR 300. This study was initiated in 1979/80 by the Karlsruhe Nuclear Research Center and performed in close cooperation with Science Applications Inc., Palo Alto, USA, and Interatom Company. Part of the results also was integrated in the 'Risk Related Analysis for the SNR 300' carried out by the Gesellschaft fuer Reactorsicherheit. The character of the study described here is similar to other risk studies like the Reactor Safety Study and the German Risk Study for Nuclear Power Plants. The objectives and the methodology of the analyses are described and its results are discussed. (orig./RW).

Energy Technology Data Exchange (ETDEWEB)

The focus of the Chernobyl Studies Project has now turned to the issue of health effects from the Chernobyl accident. Currently, we are involved in and making progress on the case-control and co-hort studies of thyroid diseases among Belarussian children. Dosimetric aspects are a fundamental part of these studies. We are working to implement similar studies in Ukraine. A major part of the effort of these projects is supporting these studies, both by providing methods and applications of dose reconstruction and by providing support and equipment for the medical teams.

Energy Technology Data Exchange (ETDEWEB)

A novel device called the gas expansion model (GEM) is being developed at the Hanford Engineering Development Laboratory for testing in the 400-MW(th) fast flux test facility (FFTF) reactor. Incorporation of the GEM into liquid-metal reactor designs is intended to measurably contribute to the achievement of inherent safety, by allowing the reactor to passively shut down even in the extremely remote (hypothetical) event of an unprotected (no scram) loss-of-flow accident. The purpose of this paper is to describe the GEM and present predictive analyses of the effectiveness of the device during unprotected loss-of-flow experiments in the FFTF.

International Nuclear Information System (INIS)

When analyzing the loss-of-coolant accidents at VVER reactor NPP the problem of the effect of noncondensable gases on heat transfer in a horizontal steam generator (HSG) is gaining in importance. Based on the RELAP5/MOD3.2 computer code one analyzed the experiments to condense steam-and-gas mixture in a HSG. The calculations are shown to predict satisfactorily duration of steam generator poisoning from noncondensable gas

Energy Technology Data Exchange (ETDEWEB)

During LOCA (Loss of Coolant Accident) in design bases accident (DBA), emergency core coolant supplements form a recirculation sump and cooled core and containment. When the double ended guillotine Break (DEGB) at the hot leg near steam generator, due to the jet impingement discharge flow, the debris could be potentially generated at pipe or wall nearby steam generator and be transported to the recirculation sump. Therefore, the debris, such as insulations and paint chips, could be accumulated and be clogged in the recirculation sump screen. If debris is blocked the sump strainer, the pressure drop is increased at the screen so as to increase the pressure loss of ECCS (Emergency Core Cooling System) pump NPSH (Net positive suction head). It is potentially influenced to decrease the long-term cooling capability of the recirculation sump. The recirculation sump screen clogging accident has happened in BWR of USA and Sweden. ...

Energy Technology Data Exchange (ETDEWEB)

The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled {open_quotes}Health Effects Models for Nuclear Power Plant Consequence Analysis{close_quotes}, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information ...

Energy Technology Data Exchange (ETDEWEB)

Since the integrity of the containment penetrations was confirmed under accident management (AM) conditions in the former test, failure criteria tests and aerosol trapping tests were carried out using low-voltage modules and flange gaskets of an actual plant under severe accident (SA) conditions, without AM. The safety margin for failure temperature of the penetrations and the credit for fission product (FP) aerosol trapping effect along the leakage paths of the degraded penetrations were evaluated in the present tests. Failure temperature of the penetrations ranged from 270 to 300degC for low-voltage modules and 300 to 350degC for flange gaskets under 0.4 to 1.0 MPa conditions. Pressure dependency on failure temperature was small. This means that the safety margin of failure temperature under AM condition is more than 70degC. By introducing a equivalent leak area for the damaged test pieces, total leak area was estimated which was smaller than ...

Energy Technology Data Exchange (ETDEWEB)

Several steam generator tube rupture accidents have occurred at plants in the past. In this paper the Computational Multi-Fluid Dynamics (CMFD) investigation of the horizontal steam generator thermal-hydraulics during the tube rupture accident is performed. A guillotine of a steam generator U-tube is assumed with choked flow from the primary to the secondary side of the steam generator. We have computed water and steam velocity fields, steam volume fraction distribution on the steam generator secondary (shell) side, as well as the swell level increase. The simulation results are a support to the safety analyses of the steam generator tube rupture accident. Numerical simulation is performed with the multidimensional multi-fluid modelling approach. The two-phase flow around steam generator tubes in the bundle is modelled by the porous media approach. Interfacial mass, momentum and energy transfer are modelled with the closure ...

International Nuclear Information System (INIS)

The understanding of fission product deposition in realistic steam generator conditions is needed for release estimates in PSA studies, and for the design of efficient accident management procedures. This is considered very important because primary-to-secondary leakages risk dominant sequences in many plants. Furthermore, the decay heat of the fission product deposits adds to the thermal load to the steam generator tubes also in other sequences, especially in case of cold leg leakages. This brings out the concern of induced steam generator tube ruptures in cases, where the steam generators are initially intact. The experimental data showed that the highest deposited fraction within the tubes were found in cases with lowest flow velocities. The minimum value of the deposited fraction was observed at intermediate flow velocities. With these relatively low Reynolds numbers, the results calculated with deposition models agree well with the experiments. At high ...

Energy Technology Data Exchange (ETDEWEB)

To date, no single passive detector has been found that measures dose equivalent from ionizing radiation exposure in low-Earth orbit. We have developed the I.S.S. Passive Dosimetry System (P.D.S.), utilizing a combination of TLD in the form of the self-contained Pille TLD system and stacks of CR-39 plastic nuclear track detector (P.N.T.D.) oriented in three mutually orthogonal directions, to measure total dose and dose equivalent aboard the International Space Station (I.S.S.). The Pille TLD system, consisting on an on board reader and a large number of Ca{sub 2}SO{sub 4}:Dy TLD cells, is used to measure absorbed dose. The Pille TLD cells are read out and annealed by the I.S.S. crew on orbit, such that dose information for any time period or condition, e.g. for E.V.A. or following a solar particle event, is immediately available. Near-tissue equivalent CR-39 P.N.T.D. provides Let spectrum, dose, and dose equivalent from charged particles of LET{sub {infinity}}H{sub 2}O {>=} 10 ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this project was to compile a bibliography of references containing environmental transuranic radionuclide data. The authors intent was to identify those parameters affecting transuranic radionuclide transport that may be generic and those that may be dependent on chemical form and/or environmental conditions. An understanding of the unique characteristics and similarities between source terms and environmental conditions relative to transuranic radionuclide transport and cycling will provide the ability to assess and predict the long term impact on man and the environment. An additional goal of the literature review, was to extract the ranges of environmental transuranic radionuclide data from the identified references for inclusion in a data base. Related to source term, these ranges of data can be used to calculate the dose to man from the radionuclides, and to perform uncertainty analyses on these dose assessments.

British Library Electronic Table of Contents (United Kingdom)

Using longitudinal data from Ukraine we examine the extent of any long-lasting effects of exposure to the Chernobyl disaster on the health and labour market performance of the adult workforce. Variation in the local area level of radiation fallout from the Chernobyl accident is considered as a random exogenous shock with which to try to establish its causal impact on poor health, labour force participation, hours worked and wages. There appears to be a significant positive association between local area-level radiation dosage and perception of poor health, though much weaker associations between local area-level dosage and other specific self-reported health conditions. There is also some evidence to suggest that those who lived in areas more exposed to Chernobyl-induced radiation have sig...

International Nuclear Information System (INIS)

More than twenty years after commencing commercial operation in 1983, Wolsong Unit 1(W1- NPP), the first CANDU Pressurized Heavy Water Reactor (PHWR) in Korea, has been undergoing refurbishment. Safety analyses were required to evaluate the safety of W1-NPP because significant amount of equipment has been refurbished. To evaluate the effectiveness of W1-NPP after these upgrades, new safety analyses were performed using the same technical standards of Wolsong Units 2, 3, 4 (W234-NPP) for Design Basis Accidents (DBA). The refurbished W1- NPP is expected to be licensed for full power operation based on the verified safety analysis results that are obtained by using the upgraded computer codes and newly adopted technical standards of W234-NPP

Energy Technology Data Exchange (ETDEWEB)

Since 1976, the Nuclear Engineering Laboratory of the Technical Research Centre of Finland and Lappeenranta University of Technology have cooperated in the field of nuclear reactor thermal-hydraulics. During these years, a series of experimental facilities (REWET-I, -II, -III, VEERA) simulating pressurized water reactors (PWRs) have been built. The newest facility, PACTEL (Parallel Channel Test Loop), is an experimental out-of-pile facility designed to simulate the major components and system behaviour of a commercial PWR during postulated small and medium size break loss-of-coolant accidents (LOCAs), natural circulation and operational transients. A PACTEL natural circulation experiment has been carried out as an OECD/NEA international standard problem ISP 33. (2 refs., 3 figs., 2 tabs.).

International Nuclear Information System (INIS)

The sequence of events from the assembly of a space nuclear power system to its integration in the Space Shuttle Transportation System (STS) is considered. First, the sequence followed for SNAP-10A, the only free world space reactor electric power system ever launched and operated in space, is reviewed. Before shipment, the SNAP-10A reactor was raised to operating temperature using electrically supplied heat and operated at low power for control calibration. Next we discuss shipment to the launch site, a phase that is critical because of the potential for various accidents. Once the power system arrives at the launch site, the processing sequence is performed. This sequence includes checkout, mating with the payload or upper stage launch vehicle, and integration into the STS.

International Nuclear Information System (INIS)

This paper addresses the unique related aspects of the LMFBR concept which are of significance to containment design and structural analysis. Topics covered include: Primary boundary integrity assurance; Effects of sodium spills on integrity of structures; Provisions being considered for containment of melted cores; Fuel handling accidents. Specific reference is made to the FFTF and the Clinch River Breeder Reactor Project designs and methods of treatment of the above problems. In particular, the part played by tests, such as those carried out on a simulated FFTF model, and the planned structural reliability and related programs are considered. Where practicable, these topics are addressed in a manner which places FFTF and CRBR in context with other LMFBR's and point to a possible direction for future American LMFBR designs. (Auth.).

International Nuclear Information System (INIS)

The aim of the risk assessment consists of a comparative security evaluation for the SNR-300 and the PWR Biblis B. The failure analysis focusses on the reactor core; in addition, possible fission product release from the spent fuel pits is examined. By reliability analyses, the frequency of events leading to incidents is determined together with the probability of core destruction. In the accident analysis, the kind and frequency of failure of the activity barriers, i.e., primary system (reactorvessel) and inner and outer containment are investigated for the various incident sequences. The radionuclide release into the environment is classified into five different release categories. Besides internal failures, external causes (especially earthquakes and plane crashes) are considered under the aspect of their risk contribution. (RF).

Energy Technology Data Exchange (ETDEWEB)

One of the alternative ideas to solve the spent fuel issues, the partitioning and transmutation (P and T) technology has been developed for decades. Moreover, the concept of LILW production from P and T are proposed by Bowman. A PEACER (Proliferationresistant, Environmental-friendly Accident-tolerant, Continuable and Economical Reactor), based on pyrochemical process and Pb-Bi coolant transmutation reactor, has been conceptually designed to be able to convert all PWR spent fuel into low and intermediate level waste for near-surface disposal. In this study, the acceptance criteria for near-surface disposal facility is derived by the methodology for establishment of acceptance criteria. Then acceptable TRU decontamination factor (DF) and LLFP removal efficiency in order to meet acceptance criteria is evaluated.

Energy Technology Data Exchange (ETDEWEB)

A cold-leg small-break loss-of-coolant accident (SBLOCA) experiment was performed on the PACTEL facility to study the behavior of natural circulation in a VVER-440 reactor geometry. The facility is a volumetrically scaled (1:305) integral test loop simulating the VVER-440 reactors used in Finland. The test results were used to assess the computer codes RELAP5/MOD3.1 and APROS 3.0 for VVER reactors. The behavior of the horizontal steam generator and the effect of the hot-leg loop seal were of particular interest. The specific parameters to be compared included the primary pressure and the downcomer mass flow rate.

International Nuclear Information System (INIS)

Protective goals set the line of orientation of tasks and activities in the field of accident prevention. They have to be based on safety-science methods in order to develop from the conceptual idea to the practically feasible solution, while using the scientific methods to take into account the facts and the capabilities of a situation and, proceeding from them, finding an efficient and rational, optimal pragmatic approach by way of various strategies or tactics. In this process, the activities of defining, informing, thinking and developing need the proper terminology. Safety is absence of danger, protection is limitation of danger and prevention of damage. So it is protection what is needed with danger being given, and risks have to be minimized. Riskology is a novel method of safety science, combining risk analysis and risk control into a systematic concept which is practice-oriented. Applying this to the field of nuclear engineering, the hitherto achieved ...

International Nuclear Information System (INIS)

In the Carpathian Basin, significant percentage of watershed area and floodplains of rivers are utilised agriculturally. Several potential sources of poisonous metal pollution have been identified in these areas. Because of spills from some of them a few severe accidents have happened especially in the watershed area of Tisza River during the last decades. The motivation of our present work was to produce "6"2","6"5Zn and "2"0"3Pb radioisotopes because they can be used especially as tracers for studying the kinetics of uptake, transport and accumulation of zinc and lead by plants under different circumstances. (orig.)

International Nuclear Information System (INIS)

This report presents a view of and commentary on the current status of equipment qualification (EQ) in nuclear industries of the major western nations. The introductory chapters discuss the concepts of EQ, the elements of EQ process and highlight some of the key issues in EQ. A brief review of industry practices and some of the prevalent industrial standards is presented, followed by an overview of current regulatory positions in the USA, France, Germany and Sweden. A summary and commentary on the latest research findings on issues relating to accident simulation, to aging simulation and some special topics related to EQ, has been contributed by Franklin Research Centre of Philadelphia. The last part of the report deals with equipment qualification in Canada and gives recommendations on EQ for new plants as well as currently operational CANDU nuclear power plants.

Energy Technology Data Exchange (ETDEWEB)

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors ...

International Nuclear Information System (INIS)

By the department Radio-ecology of the Laboratory for Radiation Research, in the period 1981 up to 1989 inclusive, the transfer has been studied, from soil to plant, of a number of important activation and fission products, originating in the nuclear-power production in nuclear power plants. The purpose of this study was twofold: on the one side the quantification of this transfer for various agrarian systems and on the other side to find out in how far, after an accidental contamination, certain agriculture activities can influence essentially the transfer and subsequently the radiation burden for the population. Emphasis lay, the last years, in particular upon this second aspect. The results of this study form essential basic data for diffusion models for radioactive materials which, in turn, are important in estimating the effects of measures. (author). 6 refs.; 4 figs.

Energy Technology Data Exchange (ETDEWEB)

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

Energy Technology Data Exchange (ETDEWEB)

The program introduced here for determining and describing the multi-axial creep of pipes is based, on the one hand, on the results of the nuclear process heat prototype plant material program and, on the other hand, on the possible load conditions which arise for evaluating accidents or extreme working situations. The basis of a theoretical description of multi-axial creep is the invariant theory in which both the von Mises configuration change hypothesis and the Norton creep law are included. Combined tension and torsion are also considered in detail, the superimposition of cyclic stresses in the tensile threshold area is discussed and cases of partial relaxation are explained. Experimental results for the discussed loads are introduced, which have led to satisfactory agreement between theory and experiment. (orig./MM).

Energy Technology Data Exchange (ETDEWEB)

This experimental research investigates mixed convection and heat transfer augmentation by gaseous forced jets in a large enclosure, at conditions simulating those of passive containment cooling systems for Gen III+ passively safe reactors. The experiment is designed to measure the key parameters governing heat transfer augmentation by forced jets, and to investigate the effects of geometric factors, including the jet diameter, jet injection orientation, interior structures, and enclosure aspect ratio. The tests cover a variety of injection modes leading to flow configurations of interest for mixing and stratification phenomena in containments under accident conditions. Correlations for heat transfer augmentation by forced jets are developed and compared with experimental data. The characteristic recirculation speed inside the enclosure is introduced and analyzed. Steady stratified temperature distributions are compared with model simulations of the BMIX++ code.

Energy Technology Data Exchange (ETDEWEB)

The Level 1 probabilistic safety assessment (PSA) for Wolsung(WS) 2/3/4 nuclear power plant (NPPs) in design stage is performed using the methodologies being equivalent to PWR PSA. Accident sequence evaluation program (ASEP) human reliability analysis (HRA) procedure and technique for human error rate prediction (THERR) are used in HRA of WS 2/3/4 NPPs PSA. The= purpose of this paper is to introduce the procedure and methodology of HRA in WS 2/3/4 NPPs PSA. Also, this paper describes the interim results of importance analysis for human actions modeled in WS 2/3/4 PSA and the findings and recommendations of administrative control of secondary control area from the view of human factors. (Author) 10 refs., 2 tabs.

British Library Electronic Table of Contents (United Kingdom)

Fault symptoms of running gearboxes must be detected as early as possible to avoid serious accidents. Diverse advanced methods are developed for this challenging task. However, for multiwavelet transforms, the fixed basis functions independent of the input dynamic response signals will possibly reduce the accuracy of fault diagnosis. Meanwhile, for multiwavelet denoising technique, the universal threshold denoising tends to overkill important but weak features in gear fault diagnosis. To overcome the shortcoming, a novel method incorporating customized (i.e., signal-based) multiwavelet lifting schemes with sliding window denoising is proposed in this paper. On the basis of Hermite spline interpolation, various vector prediction and update operators with the desirable properties of biorthog...

British Library Electronic Table of Contents (United Kingdom)

There is little flame spread data for homogeneous energetic materials and no data for nitramines. We report the results of flame spread experiments of PBX 9501 (HMX (cyclotetramethylenetetranitramine) based explosive). The horizontal flame spread rate, Sf, is of the same order of magnitude as normal deflagration and varies nearly as the square root of pressure, as our scaling analysis presented here predicts. In the vertical orientation, the flame propagation downward was observed to be slightly faster than horizontal flame spread, presumably because of the melt layer flowing downward on the sample. In an accident scenario, a charge may be fractured or the surface roughened. Consequently, we also examined the effect of roughness. Minor roughness created by explosives machining was found to...

Energy Technology Data Exchange (ETDEWEB)

Sheltering, evacuation and distribution of stable iodine tablets are considered to be major early emergency actions aiming at diminishing the consequences after a release of radioactive materials from nuclear power plants into the air. Whether in real situations emergency managers will act accordingly is hard to predict. Uncertainties associated with these decisions are termed 'volitional' uncertainties. These uncertainties, however, cannot be assessed by expert judgements as they express the decision at stake in an emergency situation. Uncertainties on the times to implement countermeasures and on the times for the general population to respond to these measures can be assessed by experts, as they represent 'lack-of-knowledge' uncertainties. This paper describes the difference in approach of both types of uncertainties and shows the results of expert judgements on the latter type of uncertainties in early emergency actions. Ten experts from seven ...

Energy Technology Data Exchange (ETDEWEB)

RELAP5/MOD3 code was developed for western type power water reactors with vertical steam generators. Thus, this code should be validated also for VVER design with horizontal steam generators. The validation work, which has been started in Lappeenranta University of Technology (LUT), has already shown some weaknesses of the code. For example the flow inside a steam generator horizontal tube in some accident cases is not correctly modelled by the code. It may be the result of erroneous prediction of the flow regime. The aim of the study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal tubes. (18 refs.).

Energy Technology Data Exchange (ETDEWEB)

This document provides environmental information on postulated closure options for the Savannah River Laboratory Seepage Basins at the Savannah River Plant and was developed as background technical documentation for the Department of Energy`s proposed Environmental Impact Statement (EIS) on waste management activities for groundwater protection at the plant. The results of groundwater and atmospheric pathway analyses, accident analysis, and other environmental assessments discussed in this document are based upon a conservative analysis of all foreseeable scenarios as defined by the National Environmental Policy Act (CFR, 1986). The scenarios do not necessarily represent actual environmental conditions. This document is not meant to be used as a closure plan or other regulatory document to comply with required federal or state environmental regulations.