Energy Technology Data Exchange (ETDEWEB)

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).

International Nuclear Information System (INIS)

The Surry low power and shutdown probabilistic risk analysis (PRA) is an ongoing project at Brookhaven National Laboratory (BNL) to identify and quantify potential accident scenarios that may occur in a pressurized water reactor (PWR) during low power and shutdown. It was initiated as a result of various incidents and accidents that have occurred within the United States and overseas. The project involves review and evaluation of PWR experience at shutdown, identification of accident scenarios, determination of methods to mitigate the accidents, and performance a level 1 PRA. An evaluation of accident progression, source terms and consequences has also been initiated. The results will be used to address issues related to shutdown conditions. The objective of this paper is to provide a progress report on the project, and ...

International Nuclear Information System (INIS)

We report on severe accident scenarios consequences evaluation in connection to the applied emergency countermeasures and use of the PC COSYMA code. We present some of the results for the reactor core melt accident assumed to happen at the 632 MWE PWR Krsko Nuclear Power Plant in Slovenia. The efficiency of several potential countermeasures in limiting the late health effects was studied. Regarding the source term, the majority of release parameters are as specified for category 2 in the German Risk Study. Site specific data were used. As the outside (meteorologic) conditions during the potential accident onset can be very different, the study limited to the deterministic runs, assuming the wind direction upstream the Sava river into the WNW direction, wind speed of 5 ms -1 and the C Pasquill stability category. The population distribution file was formed from the NEK-FSAR data for ...

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

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)

This report draws an account of the consequences of Chernobyl accident 14 years after the disaster. It is made up of 8 chapters whose titles are: (1) Some figures about Chernobyl accident, (2) Chernobyl nuclear power plant, (3)Sanitary consequences of Chernobyl accident, (4) The management of contaminated lands, (5) The impact in France of Chernobyl fallout, (6) International cooperation, (7) More information about Chernobyl and (8) Glossary.

Energy Technology Data Exchange (ETDEWEB)

The chapter presents the nuclear energy development in 20th century and the most important radiation accidents happened from the point of view of technological risk and high impact consequences: Three Mile Island and Chernobyl.

International Nuclear Information System (INIS)

Industrial radiography accounts for approximately half of all the reported accidents for the nuclear related industry, in both developed and developing countries. This Safety Report is the result of a review made of a large selection of accidents in industrial radiography reported by regulatory authorities, professional associations and scientific journals. A small, representative selection of 43 accident descriptions has been used to illustrate the primary causes of radiography accidents, and a set of measures provided to prevent the recurrence of such accidents or to mitigate the consequences of those that do occur. These accident descriptions were categorized by primary causes as follows: inadequate regulatory control; failure to follow operational procedures; inadequate training; inadequate maintenance; human error; equipment malfunction ...

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)

In the revised GCFR Safety Program Plan a quantitative risk limit line has been adopted to establish requirements for the safety related functions and systems. The risk limit line is derived from an interpretation of NRC established licensing requirements, including those for LMFBR's. Multiple barriers to the progression of accident sequences are defined in the form of six Lines of Protection (LOPs). LOPs-1 to 3 are dedicated to accident prevention and represent the normal operating systems, the dedicated safety systems and the inherent design features, respectively. LOPs-4 to 6 are dedicated to the mitigation of core melt accident consequences and include in-vessel accident containment, secondary containment integrity and radiological attenuation, respectively. Cumulative frequency limits and consequence limits are established for each LOP. Design features ...

International Nuclear Information System (INIS)

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. (AK).

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

Energy Technology Data Exchange (ETDEWEB)

In the framework of the plutonium production optimization in the PWR, many solutions are studied to decrease or recycle the plutonium of the nuclear fuels. Among these solutions, the inert matrix fuels (IMF) are proposed in this thesis. In seven chapters the author presents, the context and the state of the art, the different matrix, the calculi codes such as APOLLO2 or TRIPOLI4 needed to the neutronic analysis, the different fuel assemblies (CERMET UO{sub 2}, MOX, PuO{sub 2} and PuO{sub 2}-UO{sub 2}), the efficiency of the control rods in the case of the PWR, the cross sections problem, preliminary reflexions on critical accidents. (A.L.B.)

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).

Energy Technology Data Exchange (ETDEWEB)

The analytical results of blowdown characteristics and its thrust force were compared with the experiment, which were performed as pipe whip tests under the PWR LOCA conditions on the hypothetical accident of guillotine break of pipes. The blowdown thrust force was obtained by the integral momentum equation about single-phase flow, homogeneous and separated two-phase flow, assuming critical pressure at the exit if critical flow condition was satisfied. The following results are obtained: (1) The node-junction method is useful for the analysis of water hammer phenomena and of the blowdown thrust force. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of analysis and experiment is 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the ...

International Nuclear Information System (INIS)

The multidimensional countercurrent two-phase flow regimes that occur in a pressurized-water reactor (PWR) vessel downcomer during the refill phase of a large-break loss-of-coolant accident are studied using a transparent 1/10 scale model of a PWR vessel. The various flow regimes and their distribution in the downcomer have been identified and mapped for a range of air-water flooding experiments. The two-phase flow patterns that are identified in the downcomer included various types of film flows, droplet flows, countercurrent churn flows and cocurrent flows depending on the flooding condition. Through observation of the two-phase flow dynamics it was deduced that the physical mechanisms associated with the flooding processes could be separated into a liquid entrainment process and a film flow reversal process. In addition to the above exercise, the effect of non-uniform injection of water into the downcomer via different ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

This course in System Reliability and Analysis Techniques focuses on the probabilistic quantification of accident sequences and the link between accident sequences and consequences. Other sessions in this series focus on the quantification of system reliability and the development of event trees and fault trees. This course takes the viewpoint that event tree sequences or combinations of system failures and success are available and that Boolean equations for system fault trees have been developed and are available. 93 figs., 11 tabs.

British Library Electronic Table of Contents (United Kingdom)

Background, Aim, and Scope The identification and assessment of environmental tradeoffs is a strongpoint of life cycle assessment (LCA). A tradeoff made in many product systems is the exchange of potential for occupational accidents with the additional use of energy and materials. Net benefits of safety measures with respect to human health are best illustrated if the consequences avoided and health impacts induced by additional emissions are assessed using commensurable metrics. Our aim is to develop a human health impact indicator for offshore crane lifts. Crane lifts are a major cause of accidents on offshore oil and gas (O & G) rigs, and health impacts from crane lift accidents should be included in comparative LCA of O & G technologies if the alternatives differ in the use of crane li...

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 ...

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, ...

International Nuclear Information System (INIS)

In Aug 2002, the National Nuclear Safety Administration of China issued the policy statement for building new nuclear power plants, which requires the probability based safety goal of severe core damage must be lower than 10"-"5/a. The station blackout accident would be possible to cause a severe accident if there were no effective engineering measures to prevent or mitigate the consequences of the accident. By using MELCOR1.8.5 and KORIGEN codes, the present paper has simulated the station blackout accident for Daya Bay Nuclear Power Plant and calculated the source term and radioactivity of main fission products in the containment in the late phase of the accident. CsI is found the main part of aerosol in the containment. The Xe133 and Xe133m start releasing from the containment after its failure, and the upper limit of the amount of released radioactivity is ...

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 ...

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)

A 4-loop Pressurised Water Reactor (PWR) primary coolant system has been analysed for the postulated Loss of Coolant Accident (LOCA) event in order to derive peak dynamic loads for qualifying the design of equipment supports and pipe whip restraints. Pipe whip restraints as well as pipe and equipment supports are nonlinear by nature because of the presence of gaps and the different directional stiffnesses arising from snubber, steelwork and geometric and material interaction at the concrete to steel embedment. The different structural idealisations for the supports and restraints have an influence on the dynamic response of the structure. In the first of the two part paper a range of idealisation models for the Steam Generator and Reactor Coolant Pump vertical columns ranging from elastic stiffnesses to bilinear stiffnesses with or without preload were examined. Due to both structural and loading complexity, the behaviour of these supports were ...

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.).

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.

International Nuclear Information System (INIS)

An overview of LOCA analysis in Japanese industry is presented. The BASH-M code, developed for large scale LOCA reflooding analysis, is given as an example of verification and improvement of US computer programs are given. The code's application to the operational safety analysis concerns the following main areas: 1D drift flux model base computer program CANAC; CANAC-based advanced training simulator; emergency operating procedures. The author considers also the code application to the following new PWR safety design concepts: use of steam generators for decay heat removal at LOCA conditions; use of horizontal type steam generator for maintaining two-phase natural circulation under the reactor coolant system submerged. 9 figs.

International Nuclear Information System (INIS)

The Level 1 probabilistic safety assessment (PSA) for Wolsong(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 (ASEF) human reliability analysis (HRA) procedure and technique for human error rate prediction (THERP) 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.

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)

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 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 ...

Energy Technology Data Exchange (ETDEWEB)

In nuclear reactor safety research, the question of the possible consequences of delayed core reflood during severe accidents or anticipated transient without scram transients in boiling water reactors (BWRs) has been raised. One can envisage a very low probability accident scenario leading to core uncovery and core heat-up, followed by control blade melting and subsequential delayed reflooding of the core with unborated water before its degradation. Reflooding of the hot core causes significant increases in the peak heating, melting, and hydrogen production rates, thus increasing the probability of core degradation. However, as has been established, debris beds formed from shattered fuel rods and quenched fuel melt will be undermoderated. The reflood process of a voided, intact core was examined using the TRAC/BFI CODE.

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) ...

International Nuclear Information System (INIS)

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 action time by using the MARS platform. ...

Energy Technology Data Exchange (ETDEWEB)

Evaluations were done to determine conditions that could permit nuclear criticality with fissile uranium in low-level waste (LLW) facilities and to estimate potential radiation exposures to personnel if there were such an accident. Simultaneous hydrogeochemical and nuclear criticality studies were done (1) to identity realistic scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) to model groundwater transport of uranium and subsequent concentration via sorption or precipitation, (3) to evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits, and (4) to estimate potential radiation exposures to personnel resulting from criticality consequences. This paper presents the details of the radiation exposure calculations relying on the conditions as determined from the preceding studies detailed in a cited reference.

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 ...

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 analysis of the consequences of a hypothetical meltdown accident in an LMFBR, it is important to estimate the final location of the molten fuel pool in the concrete and ground underlying the reactor vessel. The GROWS program and the AYER program have been developed to calculate the final location of the molten fuel pool as the culmination of the transient analysis of this unusual Stefan problem but these programs require extensive computational resources. The solution is provided to the concrete and ground temperatures surrounding the stationary fuel pool and the related heat flux from the pool to the ground surface outside the containment building. This solution can be used to estimate the final location of the fuel pool and to check the end results of the sophisticated programs.

International Nuclear Information System (INIS)

This document is a synthesis of data and analysis concerning natural circulation cooling in US Pressurized Water Reactors during off-normal operation and accident transients. Its objective is the integration of important research findings concerning PWR natural circulation phenomena into a single reference document. Sources of information include the Nuclear Regulatory Commission, reactor vendors, utility sponsored research groups, utilities, national laboratories, research reports, meeting papers, archival literature, and foreign sources. Three modes of natural circulation are discussed: single-phase, two-phase, and reflux/boiling condensation. General characteristics, analytical expressions, noncondensible gas effects, secondary effects, and nonuniform flow are described with regard to each of the natural circulation modes. Plant operational data, tests in scaled experimental facilities, and analysis with thermal hydraulic system codes have ...

Energy Technology Data Exchange (ETDEWEB)

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water - 310degC and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature ...

Energy Technology Data Exchange (ETDEWEB)

The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping ...

International Nuclear Information System (INIS)

Full text of publication follows: The TUV NORD e.V. is an independent Technical Support Organisation (TSO) performing safety assessments in almost every field of technology. In nuclear safety the TUV can look back on more than 40 years of experience. In the last years in Germany PWR safety analyses were focussed on boron dilution events with the potential of reactivity transients. The possibility of coolant with a low boron concentration collected in localized areas of the reactor coolant system (RCS) can be caused by injection of coolant with less boron content from interfacing systems (external dilution) as well as separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution). Inherent dilution can e.g. occur after reflux-condenser heat transfer after a small break loss of coolant accident (SBLOCA) with a limited operability of the emergency core cooling (ECC) systems. The TUV Nord e.V. was charged ...

Energy Technology Data Exchange (ETDEWEB)

As a candidate of the new concept safety system for the next generation PWR in Japan, the hybrid safety systems, which are combination of the active and the passive safety systems, and passive core cooling system by natural circulation in the reactor coolant loop with horizontal-type steam generators during Loss of Coolant Accidents (LOCAs) are investigated. The passive safety systems are advanced accumulators (ACC), primary-side and secondary-side automatic-depressurization systems (ADS, SADS), and a gravity-driven safety injection system (GDI). The horizontal steam generator design avoids a siphon break caused from the accumulation of non-condensable gases in the tubes by using a vent line in the channel head of the steam generators. This study investigates the passive core cooling characteristics of horizontal-type steam generators under LOCAs. The integrated thermal-hydraulic test has been performed at the Simulation Loop for the Innovative ...

International Nuclear Information System (INIS)

The RADionuclide Transport, Removal, And Dose (RAD-TRAD) code is designed for U.S. Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the off-site 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. 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 removal mechanisms that use, as input, information about the ...

International Nuclear Information System (INIS)

The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report ...

International Nuclear Information System (INIS)

Korea Atomic Energy Research Institute (KAERI) has many R and D facilities in operation, including HANARO research reactor, radioactive waste treatment facility (RWTF), post-irradiation examination facility (PIEF) and irradiated material test facility (IMEF). Recently, nation-wide interest is focused on the safety and security of major industrial facilities. Safe operation of nuclear facilities is imperative because of the consequence of public disaster by radiological release/ contamination, in case of an accident. Recently, Ministry of Science and Technology (MOST) of the Korean government announced amendments of Atomic Energy laws to enforce requirements of the physical protection and radiological emergency. In this paper, the context of amended Atomic Energy laws were reviewed to confirm quality assurance measures and identify additional QA activities, if any, that is required by the amendment

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 ...

International Nuclear Information System (INIS)

In the WWER-440 reactor the primary piping consists of six horizontal loops going radially from the pressure vessel, each loop having a horizontal steam generator. In this reactor type the relatively long primary piping with many curved sections requires special attention in order to successfully eliminate the consequences of the design basis accident. Emergency supports are located in appropriate places to restrict the movements of the pipe. Under normal conditions there is a gap of some centimeters between the pipe and a support so that in the pipe can be deformed freely under changing loads. This paper deals with those energy-absorbing structures used at the Loviisa Nuclear Power Plant for protection against impact loading. Places and circumstances where energy-absorbing structures are employed are specified. Development and design of impact absorber elements are discussed and impact tests are described. (Auth.).

Energy Technology Data Exchange (ETDEWEB)

The A.C.R.O. is an association law 1901 declared at the Calvados prefecture at the date of 14. october 1986 and registered as environment protection. It was created, by more than 900 persons, in the months following the Chernobylsk accident in reaction to a lack of information and means of independent radiation monitoring. The particularity of the association is to own a laboratory of radioactivity analysis. Since the end of the nineties, the concerns include the natural sources of irradiation as the radon and apply to the consequences, out of nuclear industry, of the use of ionizing radiation or radioactive matter. On this last point, the affair of the orphan industrial site Bayard at Saint-Nicolas-d'Aliermont, massively contaminated by radium-226 devoted to the fabrication of alarm clocks, and the appearance of exemption threshold in the European law are elements at the origin of this evolution. (N.C.)

International Nuclear Information System (INIS)

This paper describes the construction of PWR horizontal steam generator in Soviet Union, the water chemistry treatment for secondary side, the design of steam separator, the test of heat transfer characteristics and operation. (author).

International Nuclear Information System (INIS)

... bwr type reactors closures fluidic control devices operation performance pwr

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

Energy Technology Data Exchange (ETDEWEB)

During the last years one focus of German PWR safety analysis was boron dilution events with the potential of reactivity transients. Coolant with a low boron concentration could be collected in localized areas of the reactor coolant system e.g. by separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux- condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. The TUeV NORD SysTec was charged by German supervisory authorities with the assessment of the safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses shall demonstrate that boron dilution events cannot lead to recriticality of the core. Hence the boron concentration at ...

Energy Technology Data Exchange (ETDEWEB)

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. During the course of follower core assessments, TUV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core ...

Energy Technology Data Exchange (ETDEWEB)

Isothermal reaction experiments were performed in the temperature range of 1000 - 1300 C in order to determine the chemical interactions between Zircaloy-4 fuel rod cladding and Inconel-718 spacer grids of Pressurized Water Reactors (PWR) under severe accident conditions. It was not possible to apply even higher temperatures since fast and complete liquefaction of the components occurred as a result of eutectic interactions during heatup. The liquid reaction products formed enhance and accelerate the degradation of the material couples and the fuel elements, respectively. Only small amounts of Inconel are necessary to liquefy large amounts of Zircaloy. Thin oxide layers on the Zircaloy surface delay the beginning of the chemical interactions with Inconel but cannot prevent them. In this work the reaction kinetics have been determined for the system: as-received and pre-oxidized Zircaloy-4/Inconel 718. The interactions can be described by ...

International Nuclear Information System (INIS)

The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P_0 > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences.

International Nuclear Information System (INIS)

During the water detritiation process most of the tritium inventory is transferred from water into the gaseous phase, then it is further enriched and finally extracted and safely stored. The control of tritium inventory is an acute issue from several points of view: - Financially - tritium is an expensive material; - Safeguard - tritium is considered as nuclear material of strategic importance; - Safety - tritium is a radioactive material: requirements for documented safety analysis report (to ensure strict limits on the total tritium allowed) and for evaluation of accident consequences associated with that inventory. Large amounts of tritium can be stored, in a very safely manner, as metal tritides. A bench-scale experiment of a tritium storage bed with integrated system for in-situ tritium inventory accountancy was designed and developed at ICSI Rm. Valcea. The calibration curve and the detection limit for this experimental model of tritium ...

International Nuclear Information System (INIS)

Accidents from three main practices: medical applications, industrial radiography and industrial irradiators are used to illustrate some common causes of accidents and the main lessons to be learned. A brief description of some of these accidents is given. Lessons learned from the described accidents are approached by subjects covering: safety culture, quality assurance, human factors, good engineering practice, defence in depth, security of sources, safety assessment and monitoring and verification compliance. (author)

International Nuclear Information System (INIS)

In the early period after the Chernobyl accident, analysis of patterns of "1"3"1I exposure of the human thyroid showed that contaminated milk was the basic source of "1"3"1I intake among the inhabitants of Russia. The equipment and techniques used for measurement of the "1"3"1I content in the thyroids of these individuals are described in this work. A model of the "1"3"1I intake, taking into account protective actions, and a method of thyroid dose calculation are discussed. The mean thyroid dose and frequency distributions of the thyroid doses to inhabitants of towns and villages of the Bryansk, Tula and Orel regions of Russia are presented. The mean dose to the thyroids of children living in the villages was 2 to 5 times higher than the dose to adult thyroids; for children living in the towns, the mean dose was 1.5 to 12 times higher. The mean thyroid mass in adult inhabitants of the Bryansk region was 27 g, which exceeded the value for a standard man (20 g) and ...

Science.gov (United States)

Recent radiochemical investigations of the PWR reactor coolant have corfirmed earlier observations that the level of activities of 33 m Cs/sup 138/, 2.8 hr Kr , and 8.1 day 1/sup 131/ are more than ten times higher than those predicted for the estimated U contamination of the Zircaloy cladding. The present fission product activity levels have not, as yet, presented any problems in the PWR. (W.L.H.)

Energy Technology Data Exchange (ETDEWEB)

A series of meteorological and tracer experiments, was conducted during May and June 1984 over the 20-km wide /O/resund strait between Denmark and Sweden for the purpose of studying atmospheric dispersion processes over cold water and warm land surfaces and providing the data needed to evaluate meso-scale models in a coastal environment. In concert with these objectives the data from these experiments have been used as part of a continuing effort to evaluate the capability of the three-dimensional MATHEW/ADPIC (M/A) atmospheric dispersion models to simulate pollutant transport and diffusion characteristics of the atmospheric during a wide variety of meteorological conditions. Since previous studies have focused primarily on M/A model evaluations over rolling and complex terrain at inland sites, the /O/resund experiments provide a unique opportunity to evaluate the models in a coastal environment. The M/A models are used by the Atmospheric Release Advisory Capability (ARAC), developed ...

Energy Technology Data Exchange (ETDEWEB)

This project is to provide the regulatory direction of containment bypass during multiple steam generator tube failure issue for the Korean Next Generation Reactors, which is a part of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : the Multiple Steam Generator Tube Repture(MSGTR) event has never been occurred in the history of commercial nuclear reactor operation but single Steam Generator Tube Rupture(SGTR) event is reported to occur every two years. A probabilistic safety analysis study on MSGTR event, however, show its probability of occurrence is to be the same order as the design basis accidents such as LACA. In this regard, the ability of NPPs to cope with MSGTR event is required. Some requirements on initial and boundary conditions are suggested to be used in the analyses of NPPs during MSGTR events. The items that should be considered in establishing regulatory requirements are ...

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...

Energy Technology Data Exchange (ETDEWEB)

A brief guide to training and equipment needed to effectively manage victims of radiation accidents. (DT)

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)

Severe accident management can be defined as the use of existing and alternative resources, systems, and actions to prevent or mitigate a core-melt accident in nuclear power plants. TRAIN (Training pRogram for AMP In NPP), developed for training control room staff and the technical group, is introduced in this report. The TRAIN composes of phenomenological knowledge base (KB), accident sequence KB and accident management procedures with AM strategy control diagrams and information needs. This TRAIN might contribute to training them by obtaining phenomenological knowledge of severe accidents, understanding plant vulnerabilities, and solving problems under high stress. 24 refs., 76 figs., 102 tabs. (Author)

International Nuclear Information System (INIS)

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by ...

International Nuclear Information System (INIS)

... Sergio Department of Materials, University of Oxford, Oxford (United Kingdom)

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)

In the early sixties it was decided to investigate salt formations on its suitability to host heat generating radioactive waste in Germany. In the reference repository concept consequently the emplacement of vitrified waste canisters in deep vertical boreholes inside a salt mine was considered whereas spent fuel should be disposed of in self shielding casks (type POLLUX) in horizontal drifts. The POLLUX casks, 65 t heavy carbon steel casks, will be laid down on the floor of a horizontal drift in one of the disposal zones to be constructed in the salt dome at the 870 m level. The space between casks and drift walls will be backfilled with crushed salt. The transport, the handling und the emplacement of POLLUX casks were subject of successfully performed demonstration and in situ tests in the nineties and resulted in an adjustment of the atomic law. The borehole disposal concept comprises the emplacement of unshielded canisters with vitrified HLW in boreholes with a ...

Energy Technology Data Exchange (ETDEWEB)

In the hypothetical event of a severe accident in a Light Water Reactor, scenarios in which the reactor pressure vessel (RPV) fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The viscosity (in fact, corium rheological behaviour) plays a major role in many phenomena such as core melt down, discharge from reactor pressure vessel, interaction with structural materials (concrete,...) and spreading in a core-catcher. For these reasons, it is important to be able to predict the rheological behaviour of corium melts of different compositions (essentially based on UO{sub 2}, ZrO{sub 2}, Fe{sub x}O{sub y} and Fe for in-vessel scenarios, plus SiO{sub 2} and CaO for ex-vessel scenarios) at temperatures above solidus temperature. In the case of corium-concrete mixtures, the increase of viscosity depends not only on the increase of particles in the melts but also on the increase of the residual liquid phase viscosity (due to ...

Energy Technology Data Exchange (ETDEWEB)

Following the Chernobyl nuclear disaster, a considerable amount of effort and resources were allocated worldwide to designing and developing coherent and comprehensive decision support systems for nuclear or radiological emergency management. They range from simple radiological consequence assessment tools to more advanced systems, incorporating the assessment of countermeasures and their effectiveness. Furthermore, many of these systems have been tailored to answer to national emergency preparedness requirements and in some cases such as the R.O.D.O.S. and A.R.G.O.S. systems they have been successfully deployed in a number of countries. Thus, computer based decision support systems for nuclear emergencies are nowadays a reality in Europe, the US and Japan; however, there was a lack of an adequate information and data exchange mechanism that enabled them to function properly and serve the purpose that triggered their development. Within the EURATOM 5. Framework ...

Energy Technology Data Exchange (ETDEWEB)

Molybdenum is an essential element for living organisms. Moreover its radionuclides may represent an incorporation risk for members of the public and/or radiation workers after a nuclear accident or a release of radioactive materials. However, only few reliable data on Mo biokinetics in humans were available. The results of recent tracer kinetic investigations with stable isotopes have shown several differences from the ICRP data with regard to the processes of intestinal absorption and of excretion. As a consequence, the dose coefficients calculated with a revised biokinetic model deviate from the ICRP estimates. By ingestion of {sup 99}Mo radionuclides with solid food, for example, the dose to the colon may be higher of a factor up to 1 order of magnitude, due to the fraction of non-absorbed material which traverses the gastro-intestinal tract. (orig.) [Deutsch] Molybdaen ist einerseits ein fuer Lebewesen essentielles Spurenelement, ...

International Nuclear Information System (INIS)

There are a total of nine operational nuclear plants in Spain totalling 7.350 MWe. These units produced 54.265 x 106 KWh in 1990, 36% of the total generation in Spain. Seven of these plants are of the PWR type. The first plant in operation was Jose Cabrera (ZORITA) in 1968, one loop Westinghouse plant with a model 24 Steam Generator. Due to the design margin and careful operation of the Steam Generator of this plant its performance have been very good, with only 5% tubes plugged after 23 years of operation. This is one of the few units in the world that remains in phosphate chemistry. During the period 1981-1985 a total of four units, two in Almaraz and two in Asco entered in operation. These three loop s Westinghouse units use model D-3 preheater Steam Generators. The poor design and manufacture of the Steam Generators of these units have caused a large number of problems: mechanical (Preheater and AVB's vibration), denting, and primary and secondary stress ...

Energy Technology Data Exchange (ETDEWEB)

This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

UK PubMed Central (United Kingdom)

A screening questionnaire designed to take an alcohol history was used on 996 patients attending the London Hospital Accident and Emergency Department. Questions concerned with 'binge' drinking detected...Full Text Available

International Nuclear Information System (INIS)

KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

Science.gov (United States)

A post-accident-heat-removal structural effects analysis for the steel liner in the FFTF concrete containment structure is presented. (JWR)

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)

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)

Full text of publication follows: The Nuclear Power Plants (NPPs) are always designed for the highest level of safety against postulated accidents which may be initiated due to internal or external causes. One of the external/internal causes, which may lead to accident in the reactor and its associated systems, is fire in certain vital areas of the plant. Conventionally, the fire containment approach and/or the fire confinement approach is used in designing the fire protection systems of NPPs. Indian NPPs (PHWRs) follow the combined approach to ensure plant safety and all newly designed plants are required to comply with the provisions of Atomic Energy Regulatory Board (AERB) fire safety Guide. In respect of older plants, the reassessment of adequacy of fire safety provisions in the light of current advances has becomes essential so as to decide upon the steps for retrofitting. Keeping this in mind the deterministic fire hazard analysis was ...

Energy Technology Data Exchange (ETDEWEB)

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 ...

International Nuclear Information System (INIS)

Full text of publication follows: The hydrogen mitigation system of 20 igniters and 6 PARs is installed to control the hydrogen in the containment during severe accidents and design basis accidents, respectively, in Shin-Wolsung 1 and 2 nuclear power plants. The igniters are primarily installed at the hydrogen source locations, and the PARs are installed in the open spaces. The PARs will maintain the hydrogen concentration within the containment atmosphere below the limit of 4 v/o in accordance with Regulatory Guide 1.7 during design basis accidents. The igniters will maintain the hydrogen concentration within the containment atmosphere below the limit of 10 v/o in accordance with 10CFR50.34(f) during severe accidents. In addition, the PARs can be used as a supplementary means to control the hydrogen concentration during severe accidents because of their inherent passive ...

International Nuclear Information System (INIS)

Some kinematical consequences of the causal tachyons possible in bi-metric theories are considered. (author).

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 ...

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: A porous media solution PORFLO has been developed for the 3-dimensional two-phase flow by describing the process facility in Cartesian or cylindrical coordinates. The local porosity fraction is applied for distinguishing the fluid filled volumes from the solid structures. The solid structure contribute the two-phase flow through the wall friction, flow area and heat transfer. Optionally the solid structure may contain primary liquid of steam generators, steam in the higher temperature and pressure to be condensed or electrical heating power. By using these optional boundary conditions three different process facilities have been analysed. The thermohydraulic solution based on 5-equation approach, where the conservation equations are solved for the liquid and gas (vapour) mass, mixture momentum (giving the velocity only for the mixture), liquid and gas energy, is described shortly. In addition to that the principles modelling optional boundary ...

International Nuclear Information System (INIS)

The installation procedure, the fabrication, fitting up, positioning, adjustment and welding of piping, examinations, hydrostatics testing and insulation of piping for reactor primary circuit piping of Daya Bay Nuclear power Plant are briefly described.

Energy Technology Data Exchange (ETDEWEB)

Comments are made on the Public Inquiry into CEGB's proposal to construct a pressurized water reactor (PWR) at Sizewell, UK. Aspects discussed include: time elapsed and its possible effect on the result; economics of nuclear power plants compared with coal-fired power plants; changes in real sterling/dollar exchange rates; effect of mineworkers' strike; the UK electric power generating system; AGR reactors compared with PWR reactors; extension of Magnox reactor life; radioactive waste management; political decisions.

International Nuclear Information System (INIS)

Since the PWR plants in Japan have been operated more than 30 years, main instrumentation control equipment of analog systems has been renewed to digital control systems. Renewal works had to be done in short period within periodical inspection term and for several facilities. The Mitsubishi LTD group had been provided with these market needs by its digital control system (MELTAC-NplusR 3) applicable to main instrumentation control equipment for primary and secondary systems and had already finished the renewal for practical plants. (T. Tanaka)

Energy Technology Data Exchange (ETDEWEB)

Resistance of ALLOY 800 as a PWR/SG tubings to IGC/IGSCC and Ni, Co release into high temperature water was evaluated as compared with ALLOYs 600 and 690. The study to improve the resistance to IGSCC and general corrosion was also made. From the results, ALLOY 800 was evaluated to be the most reliable among the high Ni alloys tested. In addition, the alloy would obtain further reliability in IGSCC and general corrosion resistance by the combined addition of Ti and Nb, and the prior oxidation treatments, respectively. (author).

International Nuclear Information System (INIS)

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermal-hydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (author).

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)

Feb 6, 2003 ... Director, Plans and Programs, Headquarters Air Force Materiel Command, .... Commander of the Joint Task Force Southwest Asia at Prince ...

Science.gov (United States)

our exploration of space, in a manner with improved safety. ... a new Space Transportation System. ... Columbia launches as STS-107 on January 16, 2003. ...

UK PubMed Central (United Kingdom)

BackgroundDespite the grave consequences of sexual violence, and it's persistence both within and outside marriages, this subject has received relatively little attention from researchers,...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

Following the Chernobyl accident of April 1986, there has been a continual increase in the numbers of reported cases of childhood thyroid carcinoma. An EC-supported consortium to study the pathology and molecular biology of the thyroid cancers is being coordinated from the University of Cambridge. This paper reports the findings of this study so far, together with its recommendations for further studies. (author).

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)

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)

The term 'Post-CHF' was generally used in the two-phase flow regime in tube flow occurring downstream of the CHF. It has various other names such as dispersed flow, liquid-deficient flow, mist flow and film boiling because the two-phase regime is characterized by a continuous vapor phase with discrete liquid drops and a non-wetted heated surface. The regime has been adopted in a lot of applications including nuclear power plants, fossil power plants, steam generators, refrigeration systems and spray cooling, In particular, this regime has a considerable importance in the areas of light water reactor(LWR) accident analysis (off-normal operating conditions) and design in heat exchangers operating in the once-through mode where subcooled liquid enters the exchanger and superheated vapor exits. Recently, innovative PWRs adopt very high power density increases and so require increased safety margins. For instance, advanced PWRs would be going to use a new-type of spacer ...

Energy Technology Data Exchange (ETDEWEB)

PWR primary circuit radiolysis model describes oxygen/hydrogen behavior in the Westinghouse Sizewell B 4-loop PWR (SNUPPS design). The effect of oxygen ingress have also been evaluated using the same model. There is clear agreement from experimental and modelling data that the dissolved hydrogen concentration required to suppress radiolysis decreases as the temperature increases. There is good evidence from the study at the Belleville PWR that {approx}5 cc(STP)kg{sup -1} H{sub 2} is sufficient to suppress radiolysis during power operation. Modelling indicates that the minimum hydrogen concentration is about 0.5 cc (STP)kg{sup -1} at PWR operating temperatures and that the presence of boric acid has little effect on this value, although it does increase the steady-state concentration of H{sub 2}O{sub 2}. Downstream of the core the concentrations of both oxygen and hydrogen peroxide fall by about two ...

Wastenet

and its consequences on customers and transport services -Dr Alexander Hedderich, Deutsche Bahn AG, Head of Competition (

International Nuclear Information System (INIS)

An uncertainty decrease is an very important issue for enhancing risk-informed (RI) activities worldwide. Especially, a relatively large uncertainty in a level-2 (L2) PSA risk compared with level-1 internal PSA risk has been a bottleneck problem in the RI application to the extent of a severe accident management. According to the ASME PRA standard in which sources of an uncertainty to capture a category-II RI (= Option 2) capability are listed, an uncertainty analysis which identifies the key sources of an uncertainty and includes sensitivity studies for dominant contributors to LERF (Large Early Release Frequency) needs to be provided. To solve these problems, USNRC have developed the 'SPAR-LERF' model related to the L2 RI application and 'L2 uncertainty assessment and improvement' work is being taken as a main PSA2 topic of the SARNET (Severe Accident Research Network of Excellence) program in Europe by OECD/NEA. Domestically, a mid/long-term ...

Energy Technology Data Exchange (ETDEWEB)

The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces were obtained by Navier-Stokes momentum equation for a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained: (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated ...

International Nuclear Information System (INIS)

The purposes of the pipe rupture studies at the Japan Atomic Energy Research Institute are to perform the model tests on the pipe whip of a pipe-restraints system, to get jet impingement force and blowdown thrust force, and to establish the computational method for the analysis of these phenomena. This paper presents the experimental and analytical results of the pipe whip tests carried out under the PWR LOCA conditions using the test pipe of 4-inch diameter and the U-shaped restraints. In the tests, the gap between the test pipe and the restraints was set nearly constant and the overhang length was 250 mm, 400 mm or 650 mm. The dynamic strains and residual deformations of the test pipe and restraints, and the restraint force were measured to clarify the effects of the overhang length on the pipe whip behaviors of the pipe-restraints system. It was confirmed from the pressure data that the present pipe whip tests were performed under the PWR ...

Energy Technology Data Exchange (ETDEWEB)

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking ...

International Nuclear Information System (INIS)

In areas where radiation dose monitoring has not been performed, it is essential to use material available in the environment be able to rapidly assess doses to individuals for immediate emergency medical care or for general estimation of the radiological consequences. It was shown that certain types of telephone cards containing microchips have the potential to be used as individual radiation dosimeters in emergency situations to detect doses over 250 mGy by luminescence measurements. In order to understand the dosimetric properties of chip cards, the components obtained from INFINIEON Company at various stages of production were used for luminescence measurements. It is found that the protecting layer used above the chips so called 'globe top' is the main source of radiation induced signal in chip cards. The globe top produced by INFINIEON at that stage is found to contain SiO2 and Epoxy. In order to improve the dosimetric properties of the chip cards, the raw ...

International Nuclear Information System (INIS)

Flow Accelerated Corrosion (FAC) of Carbon Steel (CS) piping has been one of main issues in Light-Water Nuclear Reactor (LWRs). Wall thinning of CS piping due to FAC increases potential risk of pipe rupture and cost for inspection and replacement of damaged pipes. In particular, corrosion products generated by FAC of CS piping brought steam generator (SG) tube corrosion and degradation of thermal performance, when it intruded and accumulated in secondary side of PWR. To preserve SG integrity by suppressing the corrosion of CS, High-AVT chemistry (Feedwater pH9.8#+-#0.2) has been adopted to Tsuruga-2 (1160 MWe PWR, commercial operation in 1987) in July 2005 instead of conventional Low-AVT chemistry (Feedwater pH 9.3). By the High-AVT adoption, the accumulation rate of iron in SG was reduced to one-quarter of that under conventional Low-AVT. As a result, a tendency to degradation of the SG thermal efficiency was improved. On the other hand, it ...

International Nuclear Information System (INIS)

A reasonable understanding of PWR steam generator corrosion mechanisms such as denting and wastage has been developed, and adequate chemistry control programs defined to obviate the magnitude and effects of these modes of attack. However, relatively unique corrosion attack modes have been encountered at several plants notwithstanding the presence of a reasonable to very good chemistry control program when considered in light of the Steam Generator Owners Group chemistry guidelines. The uniqueness of attack also suggests that parameters not routinely measured or monitored may be playing a significant role. In the authors opinions, the only reasonable method of routinely identifying corrosion accelerating species present in crevices, sludge piles, and deposits in PWR steam generators is by performing detailed chemical return studies during power transients, shutdowns, and long term layups. Although it would be preferable to obtain samples from ...

International Nuclear Information System (INIS)

The Seismic Proving Test of Heavy Component with Energy Absorbing Supports has been conducted to prove the reliability of advanced seismic technology, supporting heavy component such as PWR steam generator with large capacity energy absorbing supports under the sponsorship of Ministry of Economical Trade and Industry. If energy absorbing supports are adopted for NPP heavy components, support structure of facility will be much simplified due to their seismic energy absorbing effect. The paper describes the results of lead damper element test and seismic test at Tadotsu Laboratory, using 1/2.5 scale PWR Steam Generator model supported by Lead Extraction Damper (LED) and development of characteristics evaluation formula of energy absorbing support. (author)

International Nuclear Information System (INIS)

The research committee of the Atomic Energy Society of Japan on water chemistry standard aims at establishing the private standard of water chemistry of nuclear power plants. The committee gathers up 'BWR water chemistry management manual', 'PWR primary system water chemistry management manual' and 'PWR water chemical analysis standard method', and furthermore aims at the standardization of those in future. Looking back on the committee's activities for the past four years, latest results of research of water chemistry mainly contributing to safe and reliable nuclear power plants were described with the future perspective of water chemistry and a demanded break-through. (T.T.)

Energy Technology Data Exchange (ETDEWEB)

Resistance of Alloy 800 as a PWR S/G tube material to IGC/IGSCC and Ni, Co release into water was evaluated as compared with Alloys 600 and 690. The study to improve the resistance to IGSCC and general corrosion was also made, including the effects of shot-peening on the distribution of the residual stress of the U-bent tubes and the susceptibility to SCC, the stabilizing elements on IGC/IGSCC, and prior oxidation on general corrosion. From the results, shot-peened Alloy 800 was estimated to be the most reliable S/G tube material among the high nickel alloys tested. In addition, the Alloy 800 tube would obtain further reliability in IGSCC and general corrosion resistance by the combined addition of Ti, and Nb, and the prior oxidation of the inner tube surface, respectively.

Energy Technology Data Exchange (ETDEWEB)

This paper presents experimental and analytical results of pipe whip tests performed under PWR LOCA conditions using a test pipe of 4-inch diameter and U-shaped restraints. In the tests, the effects of the overhang length on the pipe whip behavior of the piperestraints system were studied by measuring the strains and deformations of the test pipe and restraints, and the restraints forces. The equation for predicting the maximum strain at the outer surface of the pipe was derived using a static equilibrium condition. The calculated maximum strains at the outer surface of the pipe agree fairly well with experimental data. The dynamic response analysis of the pipe-restraints system was conducted by the finite element program ADINA. The applicability of the ADINA program to the pipe whip analysis is made clear through this analysis.

Energy Technology Data Exchange (ETDEWEB)

With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

Science.gov (United States)

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and ...

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.

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)

The investigations carried out on Alloy 800 in aqueous high-temperature environments in France as well as in other countries are reviewed. These studies are mainly concerned with nuclear industry where Alloy 800 can be used as structural material for steam generators of PWR, breeders or HTR. As results referred to in the literature on cracking in caustic environmens do not always agree, a discussion is presented on the matter. The behaviour of Alloy 800 in superheated steam is examined. (Auth.).

International Nuclear Information System (INIS)

Some of the origins of corrosion encountered in the secondary side of pressurized water reactor steam generators are:-sludge accumulation (a mixture of metal oxides, primarily magnetite and copper) on tube sheet and attack of tube support plates by aggressive impurities leading to denting. Although Electricite de France has not suffered from these problems, it has developed a chemical cleaning process to dissolve corrosion products at both locations. (author).

International Nuclear Information System (INIS)

The concept of preliminary transmutation of minor actinides before placement to the long-term storage is considered. The purpose of such preliminary transmutation before ultimate storage is to incinerate a part of actinides and to transform another part into new actinides providing low level of radiotoxicity accumulated in the storage. Modes of transmutation in reactors of PWR, PHWR (CANDU), and Superfenix types are compared. Among power reactors, heavy-water PHWR type reactor is most acceptable for preliminary transmutation. (author)

International Nuclear Information System (INIS)

The corrosion rate of low alloy steel SA-508 and carbon steel A-410b in simulated operation and shutdown conditions of pressurized water reactor has been determined Moreover potentiodynamic polarization curves and galvanic effect through coupling of AISI-304 have been carried out under shutdown simulated condition. (Author) 8 refs.

UK PubMed Central (United Kingdom)

Thus far there are relatively few data on the risk of leukemia among those who were exposed to external radiation during cleanup operations following the Chornobyl nuclear accident, and results...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

The motto of the U.S. Coast Guard, Semper Paratus (Always Ready), should resonate strongly with those of us in the health and safety business, because we must also be ready to deal with a variety of possible radiation accidents that could occur at any time.

UK PubMed Central (United Kingdom)

BackgroundAlthough some previous studies have suggested that posttraumatic growth (PTG) is comprised of several factors with different properties, few have examined both the association...Full Text Available

Science.gov (United States)

at 4:30 P.M. November 30, 2006 - NTSB Sends Investigators to Metro Accident in Alexandria, Virginia November 27, 2006 - (SB-06-67) John Clark Assumes New Scientific Post at...

Energy Technology Data Exchange (ETDEWEB)

This study examines some radiation accidents occurred in the past. This information has been systematically assessed to get global lessons. The experience feedback shows that the most of accidents happened in certain conditions. These conditions can be distributed in four categories: 1- perception and vigilance in occupation: accidental exposure happened by lack of vigilance in details and lack of vigilance and perception; 2- procedures: accidental exposure happened following a lack of procedures or control that were not enough complete, not enough documented or not completely implemented; 3- training and understanding: accidental exposures happened because the personnel was not enough qualified and educated, did not get the general training nor the the necessary specialized training; 4- liabilities: accidental exposures happened following lacks and ambiguity in the definition of functions of the personnel and in the hierarchy liabilities. In ...

Science.gov (United States)

The Global Road Safety Partnership (GRSP) is a global partnership of business, civil society and government working for sustained reduction of road accidents in developing and transition countries. GRSP, which started operations in 1999, has a global secr...

International Nuclear Information System (INIS)

The super simulator 'SAMPSON' has been developed to show that there exist certain safety margins for light water reactors under hypothetical severe accidents and to investigate realistic measures of accident management by simulating accidents with a parallel computer. Heat-up of fuel rods and release of fission products from fuels are important factors to evaluate source terms. Models for fuel rod heat-up, hydrogen production due to cladding oxidation and cladding deformation and failure in the core region have been developed in the fuel rod heat-up analysis module. Fuel temperatures were calculated by solving the heat conduction equation. The calculated results for fuel temperature and hydrogen production were compared with CORA-13 experiment results. The comparisons showed prediction capability for the heat-up of fuel rods. The fission product release analysis module incorporates with models for fission product transport ...

UK PubMed Central (United Kingdom)

BackgroundTraffic injuries can cause physical, psychological, and economical impairment, and affected individuals may also experience shortcomings in their post-accident care and...Full Text Available

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.).

UK PubMed Central (United Kingdom)

Gene expression is a fundamentally stochastic process, with randomness in transcription and translation leading to significant cell-to-cell variations in mRNA and protein levels. This variation...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

This dissertation is devoted to the study of the phenomenological consequences of the three-generation heterotic string models based on the Calabi-Yau compactifications and the N = 2 superconformal constructions.

UK PubMed Central (United Kingdom)

Purpose of reviewThe economic burden due to the sequela of sarcopenia (muscle wasting in the elderly) are staggering and rank similarly to the costs...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

We consider the Pauli theorem on the spin-statistics connection for faster-than-light particles. As the consequence of the unlocalizability of tachyons in space we conclude that their spin-statistics correlations are inverted.

Science.gov (United States)

Metal-semiconductor solar cells reported to date exhibit inherently low output voltages. This effect isa consequence of high diode "saturation" ...

Science.gov (United States)

... M. Taylor, T. Kurt Kyser. (2009) Feather isotope analysis discriminates age-classes of Western, Least, and Semipalmated sandpipers when plumage ... ...

UK PubMed Central (United Kingdom)

BackgroundMany difficult problems in evolutionary genomics are related to mutations that have weak effects on fitness, as the consequences of mutations with large effects are often...Full Text Available

International Nuclear Information System (INIS)

On 21 October 1994, three brothers entered a waste repository at Tammiku, Estonia, without authorization and removed a metal container enclosing a caesium-137 source. During the removal the source was dislodged and fell to the ground. One of the men picked up the source, placed it in his pocket and took it to his home in the nearby village of Kiisa. Very soon after entry into the repository he began to feel ill, and few hours later he began to vomit. The man was subsequently admitted to hospital with severe injuries to his leg and hip and died on 2 November 1994. The injury and subsequent death were not attributed to radiation exposure, and the source remained in the man's house with his wife and stepson and the boy's great-grandmother. The boy was hospitalized on 17 November with severe burns on his hands, and these were identified by a doctor as radiation induced. The authorities were alerted, and the Estonian Rescue Board recovered the source from the house. The source was returned ...

International Nuclear Information System (INIS)

Full text: Full text: The incomplete understanding of the complex mechanisms connected with the interaction between thermal-hydraulic and neutron kinetics still challenges the design and the operation of nuclear reactors and imposes the adoption of conservatism in the evaluation of safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience suggests the revisiting of those areas and the identification of design/operation requirements that can be relaxed. So far, almost all of the safety analyses of research reactors have been performed using conservative computational tools such as channel codes but, nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity. The global aim of the current work is an attempt to apply the best-estimate system thermal-hydraulic code Relap5. For this purpose, the generic IAEA research reactor Benchmark problem is re-considered for proving ...

Energy Technology Data Exchange (ETDEWEB)

An uncertainty decrease is an very important issue for enhancing risk-informed (RI) activities worldwide. Especially, a relatively large uncertainty in a level-2 (L2) PSA risk compared with level-1 internal PSA risk has been a bottleneck problem in the RI application to the extent of a severe accident management. According to the ASME PRA standard in which sources of an uncertainty to capture a category-II RI (= Option 2) capability are listed, an uncertainty analysis which identifies the key sources of an uncertainty and includes sensitivity studies for dominant contributors to LERF (Large Early Release Frequency) needs to be provided. To solve these problems, USNRC have developed the 'SPAR-LERF' model related to the L2 RI application and 'L2 uncertainty assessment and improvement' work is being taken as a main PSA2 topic of the SARNET (Severe Accident Research Network of Excellence) program in Europe by ...

Energy Technology Data Exchange (ETDEWEB)

We obtained the ratios of {sup 134}Cs/{sup 137}Cs and {sup 154}Eu/{sup 137}Cs in spent PWR fuels with gamma scanning equipment of irradiated material examination facility. The fuel segments, of which the burn-up is about 40 GWD/MTU and its cooling time is 8.4 years, are prepared in Post Irradiated Examination Facility and transported to IMEF. By considering of multi-peaks of {sup 134}Cs and {sup 154}Eu, we obtained the relative efficiency of the gamma scanning system as a function of energy. And finally we obtained the number ratios of radioactive nuclides in 72 fuel pellets, radioactive {sup 134}Cs/{sup 137}Cs and {sup 154}Eu/{sup 137}Cs. (author). 4 refs., 27 tabs., 28 figs.

International Nuclear Information System (INIS)

High resolution Monte-Carlo simulations show that the neutron spectrum, fuel burnup and fuel temperature feedback effect of a PWR core loaded with Thoria-based fuel (Th/Pu-O_2) do not significantly differ from the MOX fuelled one due to the similar neutronic characteristics of both fertile materials (Th-232, U-238). The core physics of this fuel variant is characterized by an enhanced moderator/void temperature coefficient (by factor 2.4) and high incineration rate for Pu (approx. 60 %). A PWR core loaded with the Molybdenum-based inert matrix fuel (IMF) - in contrast to MOX-, shows a harder spectrum, resulting in small temperature coefficients of reactivity and particularly in a higher fuel depletion rate as well as an enhanced TRU reduction performance. The incineration of Pu amounts to 46 % resulting, in turn, in generation of minor actinides of about 10 % of the total Pu consumption. The higher excess reactivity resulting from the initial ...

International Nuclear Information System (INIS)

It is one of the most important subjects to estimate the flow-related stability of the heat exchanger tubes. A large scale model steam generator has been developed to verify the stability of the tubes in the Japanese PWR steam generators for the two-phase flow-induced vibration and to accumulate related technical data of thermal-hydraulic and flow-induced vibration of U-bend tube bundle. The model steam generator has 230 U-bend tubes of 46 different radius and 5 columns for each of practical diameter and material, and the anti vibration bars are inserted into each spacing between tube arrays. The freon R123 has been used as the secondary side fluid in stead of water-steam two-phase. In the test, void fraction and interfacial velocities in U-bend and straight tube-bundle are measured with bi-optical probes, and vibration responses of some selected tubes are measured with strain gauges and accelerators. It is verified that the U-bend tubes are stable when they are ...

International Nuclear Information System (INIS)

Denting studies have been undertaken in order to assess the influence of the most important parameters which could initiate corrosion of the carbon-steel occurring in the tube-tube support plate crevices of some PWR steam generators. Tests have been carried out in model boilers, feedwater being polluted with sea or river water. Specific effect of chloride or sulfate and influence of oxygen content, magnetite addition and pH value were investigated. In magnetite prepacked crevices, denting is obtained within 1000 hours for sea-water pollution of 0.3 ppm chloride at the blowdown. In neutral chloride or in river water denting is observed only with oxygen addition. Denting prevention is effective in the case of an on-line addition of phosphate, boric acid or calcium hydroxide. For denting stopping, boric acid or calcium hydroxide is efficient even with a high sea-water pollution. Soaks cannot stop denting if they are not followed by an on-line treatment (boric acid, ...

International Nuclear Information System (INIS)

This project evaluated the feasibility of using a low-concentration (approx. 0.5 wt %) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The primary objective was to develop a dilute process that could be safely applied at scheduled intervals, such as during normal refueling outages, to maintain a clean operating condition in the steam generator. The dilute chemical cleaning process developed in this project was demonstrated successfully on two model generators which were operated on faulted chemistry by DOE/CRC at Commonwealth's State Line Facility. Unit 5 was cleaned after 48 days of operation with 1% seawater fouling, and Unit 6 was cleaned after 112 days of operations with Lake Michigan water. This report describes work leading to the model generator cleaning demonstrations and provides details of the cleaning operation for each model steam ...

Energy Technology Data Exchange (ETDEWEB)

The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test loop have been ...

Energy Technology Data Exchange (ETDEWEB)

Thermal-hydraulics analysis can be used to determine location and magnitude of leaks inside a pressurized water reactor (PWR) containment, as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside the containment. In addition, such a program allows for the elimination of pipe whip restraints and jet impingement shields, eliminating costs for maintenance of these supports and shields in older plants and lowering construction costs for new plants. Previously, only simple single-node containment models were used for determining leakage magnitude. This paper presents a more sophisticated multinode approach for determining the magnitude and location. The resulting sensitivities to leak can be programmed into the plant's computer system. In this way, the plant's computer ...

Energy Technology Data Exchange (ETDEWEB)

The development of KNGR (Korean Next Generation Reactor) is now in progress. KAERI is developing KNGR which is a advanced active PWR (pressurized water reactor) and 1350 MW electric capacities and is by based on UCN(Ulchin) 3 and 4 nuclear power plant which is a Korean standard PWR. In this report, the PZR (pressurizer) and Related System Design Features for KNGR which include PZR volume, PPCS (pressurizer safety valve)were analyzed. First, the Design Parameters between KNGR compared to UCH 3 and 4 were compared, and second, advanced design features of KNGR compared to UCN 3 and 4 were analyzed. After the present analysis, it has been concluded that the safety margins for the PZR level and pressure of KNGR were more increased by the larger PZR volume than those of UCN 3 and 4, for PZR minimum water level at reactor/turbine trip and PZR maximum pressure at LOCV(loss of condenser vacuum) of KNGR were higher and lower, respectively than those of ...

Energy Technology Data Exchange (ETDEWEB)

Forced vibration experiment has been conducted for the reactor building of Sendai Unit 1 nuclear power plant. The beam vibrational behaviors of the outer shielding building and the internal concrete structure have been observed by using a 50 tf vibration for low frequency region, and a 10 tf vibration for high frequency region, respectively. The outline of the experimental methods, the data handling system and the major results of experiment are described. The experimental results were simulated by an analytical model. The proper vibrational frequency and the vibration modes obtained by the analysis were compared with those obtained by the experiment. By these comparisons, the adequacy of the analytical method employed for the design was confirmed.

International Nuclear Information System (INIS)

Axial offset anomaly (AOA) is caused by the deposition of crud on the fuel cladding of a PWR. When significant levels of crud build up on the cladding, boron can accumulate in the pores of the crud as a concentrated solution or solid phase, and cause the flux depression. Numerous studies have been conducted on the primary water chemistry to reduce the amount of crud in the primary circuit to avoid radioactivity buildup and unexpected power transition in the plant. However, experiments on the crud are restricted in the laboratory because the crud is a highly radioactive material. The objective of this study is to develop a test method for simulating the deposition of crud in a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

The report deals with: (1) two-phase flow regime transitions, (2) liquid entrainment in break flow, (3) vapor pull-through, and (4) CCFL in horizontal ducts. The first three processes influence the mass flow through the break, whereas the fourth one imposes a limit on liquid flow from the steam generator through the hot leg break into the core. Correlations available in the literature which deal with these processes are presented and applied to a hot leg of a PWR, LOFT and Semiscale for quantitative estimates, as well as for determining the scale distortion in the latter two facilities.

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code, named as SPACE, for a pressurized water reactor is under development to obtain a licensing to be used for the PWR design and to hold entire proprietary rights. The task of KAERI is to develop the physical models and correlations which are required to solve the field equations. It can be divided into four parts; i) flow regime determination, ii) wall heat transfer, iii) wall and interfacial friction, iv) interfacial heat and mass transfer. This paper will describe the process to develop the models for the two-phase flow regime maps in the horizontal and vertical pipes.

Energy Technology Data Exchange (ETDEWEB)

In the tests, the effects of the overhang length on the pipe whip behavior of the pipe-restraints system were studied by measuring the strains and deformations of the test pipe and restraints, and the restraints forces. The equation for predicting the maximum strain at the outer surface of the pipe was derived using a static equilibrium condition. The calculated maximum strains at the outer surface of the pipe agree fairly well with experimental data. The dynamic response analysis of the pipe-restraints system was conducted by the finite element program ADINA. The applicability of the ADINA program to the pipe whip analysis is made clear through this analysis.

International Nuclear Information System (INIS)

The importance that must be attached to the phenomenon of stress corrosion cracking of austenitic alloys is emphasized. The relation between chemical composition of various alloys and their sensitivity to cracking is shown with particular reference to the behaviour of Alloy 800. The different effects of alkaline anc chloride environments are discussed. Studies are reported of the general corrosion of Alloy 800 and other alloys in an environment representative of the primary coolant of PWR reactors; and of the behaviour of various alloys (including Alloy 800) in the conditions envisaged for their use for steam generators with superheat up to about 550 deg.C. (U.K.).

Science.gov (United States)

Steam generators in PWR plants have been subject to denting corrosion as a result of nonprotective magnetite forming on the carbon steel support plate causing the voluminous corrosion product that eventually crimps (dents) the heat transfer tube at the support plate interface. This project was designed to determine the causes of denting and the usefulness of water chemistry changes meant to arrest denting. This volume of the final report describes laboratory research on the correlation of water chemistry, superheat, and oxygen ingress with denting in steam generators.

Energy Technology Data Exchange (ETDEWEB)

The stability-enchancing two-step (SETS) method has been used successfully in the Transient Reactor Analysis Code (TRAC) for several years. The method consists of a basic semi-implicit step combined with a stabilizer step that, taken together, eliminate the material Courant stability limit associated with standard semi-implicit numerical methods. This approach toward stability requires significantly fewer computational operations than a fully implicit method, but currently maintains the first-order accuracy in space and time of its semi-implicit predecessors.

Energy Technology Data Exchange (ETDEWEB)

This article gives an account of the recent development of light water reactors new concepts in the world. Different projects are being studied. The CE80+ from Combustion Engineering (CE) is a 1350 MWe-PWR-type reactor whose primary circuit is confined in a spherical metallic containment. This reactor was certified by NRC (national regulatory commission) in mid-1996. The APWR (advanced pressurized water reactor) is developed by MHI (Mitsubishi heavy industries) in a collaboration with Westinghouse, this PWR-type reactor fitted with 4 loops derived from the SP90 model that was developed by Westinghouse during the eighties. 2 units of ABWR (advanced boiling water reactor) were commissioned in Japan in 1996 and 1997, ABWR was certified by NRC in mid-1996. The BWR90+ is developed by ABB-atom (Sweden) and it represents a cautious advanced version of the BWR75. Passive reactors are reactors that rely only on potential energy (compressed gas, ...

Energy Technology Data Exchange (ETDEWEB)

Due to the fact that, the workers' behavior is characterized by its complexity and diversity, this issue has been seen as a great 'black box' in discussions regarding the Management Systems of SHE. Associated with this issue other arises: How conscious people? How to engage them with the process? How to improve the risk control? How to motivate the prevention? Most of these responses are discussed in the Social and Human Sciences for many years. However, it is necessary to closer the technical-operational knowledge and the human aspects, applying in the organizations' daily work, to make the working environment more safe. The purpose of this study, therefore, is examining the possibility of reducing accidents through the identification and treatment of deviations (unsafe acts and unsafe conditions), cause the whole accident, be it serious or not, begins with a small deviation. It was used as a reference tool, ...

International Nuclear Information System (INIS)

The paper discusses the safety problems connected with the conversion to dense storage of RBMK-1000 spent fuel in reactor cooling pools and independent storage facilities. Recourse to dense storage has been made for a number of reasons, among which are the absence of spent fuel shipments from the nuclear power plant site, prolongation of storage time and a partial change in storage conditions. Increasing the storage density per unit volume of the storage facility and turning to new technical procedures (as against the basic design) call for further investigation of safety problems. The safety assessment of the dense storage mode includes: (1) Selecting a list of initiating events for design basis and unforeseeable accidents; (2) Assessing dense storage safety under normal as well as design basis accident conditions; (3) Safety analysis and development of measures to compensate for unforeseen accidents. Based on the studies ...

Energy Technology Data Exchange (ETDEWEB)

Use of hydrogen in large scale applications is more usual than public is mentioning normally. Nevertheless reserve against hydrogen can be observed up to highest level decision-makers. Possibly a main reason can be found and eliminated by fixing: Some spectacular accidents happened in the past and found great interest. The publication of impressive accidents and the follow up of the events course was very carefully. The research in finding causalities in former decisions and follow up was not in the interest of some people or institutions. Important facts are even not noticed by insiders, but would have been very important for future decision makings and public acceptance of new applications. It will be demonstrated in three historical examples. Much more examples would be available and each one could help to find new applications for a saver and effective use of hydrogen in power supply. Awaking from new reserves could be avoided. Additional 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)

A nonlinear mathematical model is presented of a WWER-440 nuclear power plant horizontal steam generator. On the proposed model is based a computer program for investigating transients in steam generators during loss of coolant accidents. Processes taking place at the primary side of the steam generator are described by a set of partial differential equations while those at the secondary side of the steam generator are described by plain differential equations with the variables being complex time functions. The model takes account of the coolant as both a single- and two-phase medium, of changes in the direction of the primary coolant flow and of changes in the direction of heat transfer. Heat transfer through the wall is based on a simple model of heat transfer through a thin-walled tube and includes a correction for the heat resistance of the wall. (author).

Energy Technology Data Exchange (ETDEWEB)

Most of previous study for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contract, the local failure of the RPV nozzle region has not been studied in detail. The existence and features of nozzle failure in LAVA-ICI specimen of KAERI and LHF-4 specimen of Sandia National Lab., are observed. It is confirmed that the nozzle failure of LHF-4 specimen is due to the hoop stress in the RPV. The tensile tests in various temperatures and the creep rupture tests in various temperatures and stresses, are accomplished. The finite element analysis for LAVA-ICI experiment was confirmed, and the stress and deformation analysis results are used in LAVA-ICI experiment. 17 refs., 34 figs., 3 tabs. (Author)

Science.gov (United States)

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)

This document contains a summarization of the results and insights from the Level 1 accident sequence analyses of internally initiated events, internally initiated fire and flood events, seismically initiated events, and the Level 2/3 risk analysis of internally initiated events (excluding fire and flood) for Surry, Unit 1. The analysis was confined to mid-loop operation, which can occur during three plant operational states (identified as POSs R6 and R10 during a refueling outage, and POS D6 during drained maintenance). The report summarizes the Level 1 information contained in Volumes 2--5 and the Level 2/3 information contained in Volume 6 of NUREG/CR-6144.

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.

UK PubMed Central (United Kingdom)

Accessory cusp present on the occlusal surface may seldom pose problems. While its presence may not be a cause for alarm in most instances, it can sometimes lead to serious consequences if it is damaged....Full Text Available

UK PubMed Central (United Kingdom)

Chronic heart failure (HF) is a cardiovascular disease of cardinal importance because of several factors: a) an increasing occurrence due to the aging of the population, primary and secondary prevention...Full Text Available

UK PubMed Central (United Kingdom)

BackgroundSeverely disabling chronic pain in the adult population is strongly associated with a range of negative health consequences for individuals and high health care costs,...Full Text Available

UK PubMed Central (United Kingdom)

Latent inhibition (LI) manifests as poorer conditioning to a CS that has previously been presented without consequence. There is some evidence that LI can be potentiated by reduced mesoaccumbal dopamine...Full Text Available

UK PubMed Central (United Kingdom)

A large percentage of patients with HIV/AIDS will develop dermatological complications. Consequently, all practising clinicians and pathologists in regions with a high prevalence of HIV/AIDS must be...Full Text Available

UK PubMed Central (United Kingdom)

It has been sixty years since the Millers first described the covalent binding of carcinogens to tissue proteins. Protein covalent binding was gradually overshadowed by the emergence of DNA...Full Text Available

UK PubMed Central (United Kingdom)

BackgroundIn cancer research, the association between a gene and clinical outcome suggests the underlying etiology of the disease and consequently can motivate further studies. The...Full Text Available

UK PubMed Central (United Kingdom)

Insulin resistance plays a central role in type 2 diabetes and obesity, which develop as a consequence of genetic and environmental factors. Dietary changes including high fat diet (HFD) feeding promotes...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

The obligations under international law to reduce transfrontier air pollution is discussed in five chapters from various aspects. Consequences for the European Communities are gone into in five further chapters. (orig./HP)