Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

International Nuclear Information System (INIS)

Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

1986-01-01

As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

Aircraft accident investigation: the decision-making in initial action scenario.

Science.gov (United States)

Barreto, Marcia M; Ribeiro, Selma L O

2012-01-01

In the complex aeronautical environment, the efforts in terms of operational safety involve the adoption of proactive and reactive measures. The process of investigation begins right after the occurrence of the aeronautical accident, through the initial action. Thus, it is in the crisis scenario, that the person responsible for the initial action makes decisions and gathers the necessary information for the subsequent phases of the investigation process. Within this scenario, which is a natural environment, researches have shown the fragility of rational models of decision making. The theoretical perspective of naturalistic decision making constitutes a breakthrough in the understanding of decision problems demanded by real world. The proposal of this study was to verify if the initial action, after the occurrence of an accident, and the decision-making strategies, used by the investigators responsible for this activity, are characteristic of the naturalistic decision making theoretical approach. To attend the proposed objective a descriptive research was undertaken with a sample of professionals that work in this activity. The data collected through individual interviews were analyzed and the results demonstrated that the initial action environment, which includes restricted time, dynamic conditions, the presence of multiple actors, stress and insufficient information is characteristic of the naturalistic decision making. They also demonstrated that, when the investigators make their decisions, they use their experience and the mental simulation, intuition, improvisation, metaphors and analogues cases, as strategies, all of them related to the naturalistic approach of decision making, in order to satisfy the needs of the situation and reach the objectives of the initial action in the accident scenario.

Severe Accident Test Station Design Document

Energy Technology Data Exchange (ETDEWEB)

Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

Severe Accident Test Station Design Document

International Nuclear Information System (INIS)

Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

2015-01-01

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

Accident analysis of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.; Chi, D. Y

1998-03-01

Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

Nuclear Fuel Behaviour during Reactivity Initiated Accidents. Workshop Proceedings

International Nuclear Information System (INIS)

2010-01-01

A reactivity initiated accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. The power increase may damage the reactor core. The main objective of the workshop was to review the current status of the experimental and analytical studies of the fuel behavior during the RIA transients in PWR and BWR reactors and the acceptance criteria for RIA in use and under consideration. The workshop was organized in an opening session and 5 technical sessions: 1) Recent experimental results and experimental techniques used; 2) Modelling and Data Interpretation; 3) Code Assessment; 4) RIA Core Analysis and 5) Revision and application of safety criteria

Hot Cell Installation and Demonstration of the Severe Accident Test Station

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

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

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

Energy Technology Data Exchange (ETDEWEB)

Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

TIARA: treatment initiatives after radiological accidents

International Nuclear Information System (INIS)

Menetrier, F.; Berard, Ph.; Joussineau, S.; Stradling, N.; Hodgson, A.; List, V.; Morcillo, M.A.; Paile, W.; Holt, D.C.B.; Eriksson, T.

2007-01-01

This paper describes the objectives, and reviews the progress, of the European project 'Treatment Initiatives After Radiological Accidents' (TIARA). TIARA forms part of the 'Preparatory Action for Security Research' (PASR) launched by the European Commission in 2004. The Preparatory Action is intended to reach preliminary conclusions on the needs for the security of EU citizens. It prepared a comprehensive Security Research Programme as part of the Commission's Seventh Framework Programme proposal, which was adopted in 2006 and launched in 2007. The principal purpose of TIARA is to constitute a European network that will participate in facilitating the management of a crisis in the event of the malevolent dispersal of radionuclides into the public environment. (authors)

Severe accident testing of a personnel airlock

International Nuclear Information System (INIS)

Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

1988-01-01

Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the U.S. Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800 degrees F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. The authors provide a detailed report on the test program

77 FR 10666 - Pipeline Safety: Post Accident Drug and Alcohol Testing

Science.gov (United States)

2012-02-23

... 199 [Docket No. PHMSA-2011-0335] Pipeline Safety: Post Accident Drug and Alcohol Testing AGENCY... operators of Liquefied Natural Gas (LNG) facilities to conduct post- accident drug and alcohol tests of..., operators must drug and alcohol test each covered employee whose performance either contributed to the...

Severe Accident Test Station Activity Report

Energy Technology Data Exchange (ETDEWEB)

Pint, Bruce A [ORNL; Terrani, Kurt A [ORNL

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

Causes of several accidents in gamma radiography testing units

International Nuclear Information System (INIS)

Vykrocil, L.

1979-01-01

Three cases are described of radiation accidents in gamma flaw-detection work-places in the West Bohemian Region. The causes of the accidents stemmed from the unsatisfactory technical condition of the materials testing equipment used and nonobservance of regulations for work with radioactive sourr.es. It is necessary for precluding similar accident to improve preventive care of gamma flaw-detection equipment and to educate personnel who would be considered for coping with the situation when control over the radiation source is lost. (Ha)

Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Hobbins, R.R.

1981-01-01

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

Identification of NPP accidents using support vector classification

Energy Technology Data Exchange (ETDEWEB)

Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

2016-10-15

In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

2016-05-15

The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

MCC-15: waste/canister accident testing and analysis method

International Nuclear Information System (INIS)

Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

1985-02-01

The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

Reactivity initiated accidents and loss of shutdown - 20 years later

International Nuclear Information System (INIS)

Luxat, J.C.

2007-01-01

A review of the safety of Ontario's nuclear power reactors was conducted in 1987 after the Chernobyl accident. As part of this review an analysis was performed of a Loss of Coolant Accident in a Pickering A unit with coincident failure to shutdown. This analysis showed that the power excursion was halted by channel and calandria vessel failures leading to moderator fluid displacement. The containment structure did not fail and, at worst might suffer minor cracking at the top of the dome of the reactor building. Overall the dose consequences of such an accident were no worse than the limiting design basis dual failure event. In the intervening twenty years following this analysis, Significant experimental information has been obtained that relates to power pulse behaviour. This information, together with conservatisms in he original analysis, are reviewed and assessed in this paper. In addition, the issue of reactivity initiated events in other reactor types is reviewed to identify the reactor design characteristics that are of importance in these events. Contrary to popular belief the existence of positive coolant void reactivity is not as significant a factor as it is sometimes stated to be. On balance, with appropriate design measures, no one reactor type can be claimed to be 'more safe' than another. The underlying basis for this statement is articulated in this paper. (author)

49 CFR 655.44 - Post-accident testing.

Science.gov (United States)

2010-10-01

... best information available at the time of the decision, that the covered employee's performance can be... best available information at the time of the determination that the employee's performance could not... test any other covered employee whose performance could have contributed to the accident, as determined...

The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

2002-01-01

In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Radiation damage to the thyroid and metabolic changes in cattle in the initial and remote period after the Chernobyl accident

International Nuclear Information System (INIS)

Iljazov, R.G.; Yunousova, R.M.

1997-01-01

The initial period after the Chernobyl accident was the most dangerous for animals kept in the zone of radioactive contamination. Dose burdens from I-isotopes on the thyroid gland of cattle in the initial period after the accident contributed significantly into the alteration of the hormonal status, physiological state and productive, qualities of cattle on farms of the Gomel area of Belarus

Performance Analysis Review of Thorium TRISO Coated Particles during Manufacture, Irradiation and Accident Condition Heating Tests

International Nuclear Information System (INIS)

2015-03-01

Thorium, in combination with high enriched uranium, was used in all early high temperature reactors (HTRs). Initially, the fuel was contained in a kernel of coated particles. However, particle quality was low in the 1960s and early 1970s. Modern, high quality, tristructural isotropic (TRISO) fuel particles with thorium oxide and uranium dioxide (UO 2 ) had been manufactured since 1978 and were successfully demonstrated in irradiation and accident tests. In 1980, HTR fuels changed to low enriched uranium UO 2 TRISO fuels. The wide ranging development and demonstration programme was successful, and it established a worldwide standard that is still valid today. During the process, results of the thorium work with high quality TRISO fuel particles had not been fully evaluated or documented. This publication collects and presents the information and demonstrates the performance of thorium TRISO fuels.This publication is an outcome of the technical contract awarded under the IAEA Coordinated Research Project on Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy, initiated in 2012. It is based on the compilation and analysis of available results on thorium TRISO coated particle performance in manufacturing and during irradiation and accident condition heating tests

Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, Maxime

2014-01-01

This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

Applicability of simplified methods to evaluate consequences of criticality accident using past accident data

International Nuclear Information System (INIS)

Nakajima, Ken

2003-01-01

Applicability of four simplified methods to evaluate the consequences of criticality accident was investigated. Fissions in the initial burst and total fissions were evaluated using the simplified methods and those results were compared with the past accident data. The simplified methods give the number of fissions in the initial burst as a function of solution volume; however the accident data did not show such tendency. This would be caused by the lack of accident data for the initial burst with high accuracy. For total fissions, simplified almost reproduced the upper envelope of the accidents. However several accidents, which were beyond the applicable conditions, resulted in the larger total fissions than the evaluations. In particular, the Tokai-mura accident in 1999 gave in the largest total specific fissions, because the activation of cooling system brought the relatively high power for a long time. (author)

The influence of chemistry on severe accident phenomena in integral tests

International Nuclear Information System (INIS)

Hobbins, R.R.; Osetek, D.J.; Hagrman, D.L.

1988-01-01

The influence of chemical processes on severe accident phenomena in integral tests is reviewed and recommendations for areas of additional work are made. The results reviewed include those from tests conducted in the in-pile facilities at ACRR, PBF, and TREAT and the TMI-2 accident. Progress has been made in understanding the influence of chemistry on important severe accident phenomena such as core melt progression, hydrogen generation, aerosol generation and transport, and fission product release and transport (including revaporization). An example is the chemistry of volatile fission products, especially iodine and tellurium. Areas where understanding is inadequate are also apparent, such as chemical interactions between fission product vapors and aerosols. Influential chemical processes reviewed include oxidation by steam and interactions among control, structural, fuel, fission product, and aerosol materials

Fukushima, one year later. Initial analyses of the accident and its consequences

International Nuclear Information System (INIS)

2012-01-01

The earthquake of magnitude 9 of March 11, 2011 with an epicenter 80 km east of the Japanese island of Honshu, and the subsequent tsunami, severely affected the region of Tohoku, with major consequences for its population and infrastructure. Devastating the site of the Fukushima Dai-ichi nuclear power plant, these natural events were the cause of the core meltdowns of three nuclear reactors and the loss of cooling of several spent fuel pools. Explosions also occurred in reactor buildings 1 through 4 due to hydrogen produced during fuel degradation. Very significant radioactive releases into the environment took place. The accident was classified level 7 on the International Nuclear Event Scale (INES). This report provides an assessment and perspective on the information gathered by IRSN during the first twelve months following the disaster in an effort to understand the condition of the installations, evaluate the releases and analyze and evaluate the consequences of the accident on workers and the impact on the population and the environment. On the basis of available information, the report provides an initial analysis of the chain of events. It should be noted that a year after the accident, the full sequence of events is still not understood. Operating experience feedback from the 1979 Three Mile Island accident in the United States, in which reactor core damage was not confirmed until 1986, suggests that it may be several years before a detailed scenario can be constructed of the accident that led to radioactive releases. It will require access to the damaged installations. The situation at the site remains dangerous (reactor pressure vessels and containments are not leak-tight, diffuse releases, etc.). If it has significantly improved as a result of the significant resources deployed by the Tokyo Electro Power Company (TEPCO) to regain control of the installations, this effort must continue over the long term to begin evacuation of fuel from pools (in two

Radiation accidents

International Nuclear Information System (INIS)

Nenot, J.C.

1996-01-01

Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report

International Nuclear Information System (INIS)

Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

2016-06-01

The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

Initiating events of accidents in the practice of oil well logging in Cuba

International Nuclear Information System (INIS)

Alles Leal, A.; Perez Reyes, Y.; Dumenigo Gonzalez, C.

2013-01-01

The oil well logging is an extremely important activity within the oil industry, but in turn, brings risks that occasionally result in damage to health, the environment and economic losses. In this context, risk analysis has become an important tool to control them through their prediction and the study of the factors that determine them, enabling substantiated decisions to, first, foresee accidents and, secondly, to minimize their consequences. This paper proposes the elaboration of a list of initiating events of accidents in the practice of oil well logging which is one of the most important aspects for further evaluation of radiation safety of this practice. For its determination the technique employed to identify risks was 'Failure Modes and Effects Analysis (FMEA)' by applying it to the different stages and processes of practice. (Author)

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

Energy Technology Data Exchange (ETDEWEB)

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.; Woolstenhulme, Nicolas E.; Ban, Heng; Wachs, Daniel M.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure of the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.

Severe accident phenomena

International Nuclear Information System (INIS)

Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

1995-02-01

Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-06-01

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO₂) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

Comparison of two simulation methods for testing of algorithms to detect cyclist and pedestrian accidents in naturalistic data

OpenAIRE

Madsen, Tanja; Christensen, Mads; Sloth Andersen, Camilla; Varhelyi, Andras; Laureshyn, Aliaksei; Moeslund, Thomas; Lahrmann, Harry

2017-01-01

Naturalistic studies can potentially be used to detect accidents of vulnerable road users and thus overcome the large degree of under-reporting in the official accident records. In this study, simulated cycling and walking accidents were performed by a stuntman and with a crash test dummy to test how they differ from each other and the potential implications of using simulated accidents as an alternative to real accidents. The study consisted of simulations of common accident types for cyclis...

Total Monte-Carlo method applied to the assessment of uncertainties in a reactivity-initiated accident

Energy Technology Data Exchange (ETDEWEB)

Cruz, D.F. da; Rochman, D.; Koning, A.J. [Nuclear Research and Consultancy Group NRG, Petten (Netherlands)

2014-07-01

The Total Monte-Carlo (TMC) method has been applied extensively since 2008 to propagate the uncertainties in nuclear data for reactor parameters and fuel inventory, and for several types of advanced nuclear systems. The analyses have been performed considering different levels of complexity, ranging from a single fuel rod to a full 3-D reactor core at steady-state. The current work applies the TMC method for a full 3-D pressurized water reactor core model under steady-state and transient conditions, considering thermal-hydraulic feedback. As a transient scenario the study focused on a reactivity-initiated accident, namely a control rod ejection accident initiated by a mechanical failure of the control rod drive mechanism. The uncertainties on the main reactor parameters due to variations in nuclear data for the isotopes {sup 235},{sup 238}U, {sup 239}Pu and thermal scattering data for {sup 1}H in water were quantified. (author)

Comparison of two simulation methods for testing of algorithms to detect cyclist and pedestrian accidents in naturalistic data

OpenAIRE

Madsen, Tanja Kidholm Osmann; Christensen, Mads Bock; Andersen, Camilla Sloth; Várhelyi, András; Laureshyn, Aliaksei; Moeslund, Thomas B.; Lahrmann, Harry Spaabæk

2017-01-01

Naturalistic studies can potentially be used to detect accidents of vulnerable road users and thus overcome the large degree of under-reporting in the official accident records. In this study, simulated cycling and walking accidents were performed by a stunt man and with a crash test dummy to test how they differ from each other and the potential implications of using simulated accidents as an alternative to real accidents. The study consisted of simulations of common accident types for cycli...

Results of stress tests of European nuclear power plants after the Fukushima-Daiichi accident

International Nuclear Information System (INIS)

Kovacs, Zoltan; Novakova, Helena

2012-01-01

In response to the Fukushima-Daiichi accident, the European Council laid down the requirement that a transparent and comprehensive risk assessment exercise ('stress tests') be carried out at each European nuclear power plant. The stress tests concentrated on the nuclear power plants' safety margins in the light of the lessons learned from the accident. The reviews focused on natural external events including earthquake, tsunami and extreme weather, loss of safety functions, and severe accident management. The stress test procedure comprised 3 steps: (i) The nuclear facility operators performed the stress tests and prepared proposals for safety improvements. (ii) The national regulators performed independent reviews of the stress tests and prepared national reports. (iii) The reports submitted by the national regulators were subjected to review at a European level. The article describes the scope of the stress tests and their results, verified at the European level. (orig.)

A consistent approach to assess safety criteria for reactivity initiated accidents

International Nuclear Information System (INIS)

Sartoris, C.; Taisne, A.; Petit, M.; Barre, F.; Marchand, O.

2010-01-01

In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on: -A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena; -The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results; -The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up. This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO 2 rod at Hot Zero Power.

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

PWR accident management realated tests: some Bethsy results

International Nuclear Information System (INIS)

Clement, P.; Chataing, T.; Deruaz, R.

1993-01-01

The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

Tests of qualification of national components of nuclear power plants under design basis accident

International Nuclear Information System (INIS)

Mesquita, A.Z.

1990-01-01

With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

Superheated-steam test of ethylene propylene rubber cables using a simultaneous aging and accident environment

International Nuclear Information System (INIS)

Bennett, P.R.; St Clair, S.D.; Gilmore, T.W.

1986-06-01

The superheated-steam test exposed different ethylene propylene rubber (EPR) cables and insulation specimens to simultaneous aging and a 21-day simultaneous accident environment. In addition, some insulation specimens were exposed to five different aging conditions prior to the 21-day simultaneous accident simulation. The purpose of this superheated-steam test (a follow-on to the saturated-steam tests (NUREG/CR-3538)) was to: (1) examine electrical degradation of different configurations of EPR cables; (2) investigate differences between using superheated-steam or saturated-steam at the start of an accident simulation; (3) determine whether the aging technique used in the saturated-steam test induced artificial degradation; and (4) identify the constituents in EPR that affect moisture absorption

Test study on safety features of station blackout accident for nuclear main pump

International Nuclear Information System (INIS)

Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

2009-01-01

The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

Analysis of pressurized water reactor accidents in reactivity disturbances. II

International Nuclear Information System (INIS)

Tinka, I.

1978-01-01

The logic structure of program FATRAP is described. The time course of reactivity temporal and spatial distributions of neutron flux density and power, characteristic temperatures of the individual reactor zones and the heat flux density from cladding to the coolant can be obtained as the main results. The basic program funcitons were tested for a point and a one-dimensional model. In the basic test the absorption rod was removed uncontrollably at a preset speed for 0.5 s with the reactivity feedback operative. A second test simulated the action of the accident protection system with a delay of 0.1 s started when the 7500 MW power had been obtained. The last test consisted in simulating a start-up accident with an initial power of 2.25 MW. For the said chosen accident models reactivity feedback is responsible for the formation of the appropriate power peak while the accident protection attendance alone can considerably reduce temperatures during the process. (J.F.)

Introduction to Large-sized Test Facility for validating Containment Integrity under Severe Accidents

International Nuclear Information System (INIS)

Na, Young Su; Hong, Seongwan; Hong, Seongho; Min, Beongtae

2014-01-01

An overall assessment of containment integrity can be conducted properly by examining the hydrogen behavior in the containment building. Under severe accidents, an amount of hydrogen gases can be generated by metal oxidation and corium-concrete interaction. Hydrogen behavior in the containment building strongly depends on complicated thermal hydraulic conditions with mixed gases and steam. The performance of a PAR can be directly affected by the thermal hydraulic conditions, steam contents, gas mixture behavior and aerosol characteristics, as well as the operation of other engineering safety systems such as a spray. The models in computer codes for a severe accident assessment can be validated based on the experiment results in a large-sized test facility. The Korea Atomic Energy Research Institute (KAERI) is now preparing a large-sized test facility to examine in detail the safety issues related with hydrogen including the performance of safety devices such as a PAR in various severe accident situations. This paper introduces the KAERI test facility for validating the containment integrity under severe accidents. To validate the containment integrity, a large-sized test facility is necessary for simulating complicated phenomena induced by an amount of steam and gases, especially hydrogen released into the containment building under severe accidents. A pressure vessel 9.5 m in height and 3.4 m in diameter was designed at the KAERI test facility for the validating containment integrity, which was based on the THAI test facility with the experimental safety and the reliable measurement systems certified for a long time. This large-sized pressure vessel operated in steam and iodine as a corrosive agent was made by stainless steel 316L because of corrosion resistance for a long operating time, and a vessel was installed in at KAERI in March 2014. In the future, the control systems for temperature and pressure in a vessel will be constructed, and the measurement system

Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

The Initial Atmospheric Transport (IAT) Code: Description and Validation

Energy Technology Data Exchange (ETDEWEB)

Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-10-01

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

Monitoring Severe Accidents Using AI Techniques

International Nuclear Information System (INIS)

No, Young Gyu; Kim, Ju Hyun; Na, Man Gyun; Ahn, Kwang Il

2011-01-01

It is very difficult for nuclear power plant operators to monitor and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. The objective of this study is to develop and verify the monitoring for severe accidents using artificial intelligence (AI) techniques such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH) and fuzzy neural network (FNN). The SVC and PNN are used for event classification among the severe accidents. Also, GMDH and FNN are used to monitor for severe accidents. The inputs to AI techniques are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. In this study, 3 types of initiating events such as the hot-leg LOCA, the cold-leg LOCA and SGTR are considered and it is verified how well the proposed scenario identification algorithm using the GMDH and FNN models identifies the timings when the reactor core will be uncovered, when CET will exceed 1200 .deg. F and when the reactor vessel will fail. In cases that an initiating event develops into a severe accident, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful for operators to perform severe accident management

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

Severe accident tests and development of domestic severe accident system codes

International Nuclear Information System (INIS)

2013-01-01

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

Westinghouse accident tolerant fuel program. Current results and future plans

Energy Technology Data Exchange (ETDEWEB)

Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

2016-07-15

This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

Energy Technology Data Exchange (ETDEWEB)

Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

2016-09-15

At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

International Nuclear Information System (INIS)

Vigil, R.A.; Jacobus, M.J.

1994-04-01

Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

Review of accident analyses performed at Mochovce NPP

International Nuclear Information System (INIS)

Siko, D.

2000-01-01

In this paper the review of accident analysis performed in NPP Mochovce V-1 is presented. The scope of these safety measures was defined and development in the T SSM for NPP Mochovce Nuclear Safety Improvements Report' issued in July 1995. The main objectives of these safety measures were the followings: (a) to establish the criteria for selection and classification of accidental events, as well as defining the list of initiating events to be analysed. Accident classification to the individual groups must be performed in accordance with RG 1.70 and IAEA recommendations 'Guidelines for Accidental Analysis of WWER NPP' (IAEA-EBR-WWER-01) to select boundary cases to be calculated from the scope of initiating events; (b ) to elaborate the accident analysis methodology that also includes acceptance criteria for their result evaluation, initial and boundary conditions, assumption related with the application of the single failure criteria, requirements on the analysis quality, used computer codes, as well as NPP models and input data for the accident analysis; (c) to perform the accident analysis for the Pre-operational Safety Report (POSAR); (d) to provide a synthetic report addressing the validity range of codes models and correlations, the assessment against relevant tests results, the evidence of the user qualification, the modernisation and nodding scheme for the plant and the justification of used computer codes. Analyses results showed that all acceptance criteria were met with satisfactory margin and design of the NPP Mochovce is accurate. (author)

Dominant accident sequences in Oconee-1 pressurized water reactor

International Nuclear Information System (INIS)

Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

1985-04-01

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

A Study on the Operation Strategy for Combined Accident including TLOFW accident

International Nuclear Information System (INIS)

Kim, Bo Gyung; Kang, Gook Young; Yoon, Ho Joon

2014-01-01

It is difficult for operators to recognize the necessity of a feed-and-bleed (F-B) operation when the loss of coolant accident and failure of secondary side occur. An F-B operation directly cools down the reactor coolant system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. The plant is not always necessary the F-B operation when the secondary side is failed. It is not necessary to initiate an F-B operation in the case of a medium or large break because these cases correspond to low RCS pressure sequences when the secondary side is failed. If the break size is too small to sufficiently decrease the RCS pressure, the F-B operation is necessary. Therefore, in the case of a combined accident including a secondary cooling system failure, the provision of clear information will play a critical role in the operators' decision to initiate an F-B operation. This study focuses on the how we establish the operation strategy for combined accident including the failure of secondary side in consideration of plant and operating conditions. Previous studies have usually focused on accidents involving a TLOFW accident. The plant conditions to make the operators confused seriously are usually the combined accident because the ORP only focuses on a single accident and FRP is less familiar with operators. The relationship between CET and PCT under various plant conditions is important to decide the limitation of initiating the F-B operation to prevent core damage

Stress in accident and post-accident management at Chernobyl

International Nuclear Information System (INIS)

Girard, P.; Dubreuil, G.H.

1996-01-01

The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an anlysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of post-accident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. (Author)

Researches of WWER fuel rods behaviour under RIA accident conditions

International Nuclear Information System (INIS)

Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

2003-01-01

Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

NIRS report of the criticality accident in a uranium conversion test plant in Tokai-mura

International Nuclear Information System (INIS)

2001-01-01

This report is a detailed account of the roles that National Institute of Radiological Sciences (NIRS) played at the criticality accident in the title, which occurred at around 10:35, on Sep. 30, 1999 and resulted in death of two workers after all, and is published to discharge NIRS responsibilities in regards to the accident. The accident caused many residents concern on their health and rumors had both social and economic consequences. The report involves chapters of detailed outline of the accident; demand for acceptance of the victims and communications until the identification of the criticality'' accident; the acceptance and initial treatment; the exposure dose estimation (based on acute symptoms, on physics, on chromosomal analyses and on neutron-activated dental metals, and detailed analyses for dose distribution); decision made for therapeutic strategies; cooperation with the Network Council for Radiation Emergency and with other medical facilities; the urgent import of medicine; treatment and processes (patients, nursing system and radiation injuries); radiation protection in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hopefully useful in preventing the occurrence of future accidents. (N.I.)

High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

Energy Technology Data Exchange (ETDEWEB)

Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2013-12-23

The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H₂ environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H₂O with the alumina reaction tube at 1700°C resulted in Al(OH)₃

Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables

International Nuclear Information System (INIS)

Jacobus, M.J.

1992-11-01

This report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. Three sets of cables were aged for up to 9 months under simultaneous thermal (≅100 degrees C) and radiation (≅0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate irradiation (≅6 kGy/hr) and high temperature steam followed the aging. Also exposed to the accident conditions was a fourth set of cables, which were unaged. The test results indicate that, properly installed, most of the various miscellaneous cable products tested should be able to survive an accident after 60 years for total aging doses of at least 150 kGy or higher (depending on the material) and for moderate ambient temperatures on the order of 45--55 degrees C (potentially higher or lower, depending on material specific activtion energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation

A defense in depth approach for nuclear power plant accident management

Energy Technology Data Exchange (ETDEWEB)

Chih-Yao Hsieh; Hwai-Pwu Chou [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu, TW (China)

2015-07-01

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident

A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Fujishiro, Toshio

1984-03-01

Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

Comparison of interior crashworthiness observed in passenger train accidents and 8G dynamic seat sled tests

Science.gov (United States)

2012-04-17

The Office of Research and Development of the Federal Railroad Administration conducts engineering research to address protection of passengers and crew during train accidents. This research includes accident investigations and dynamic seat testing t...

Electron probe X-ray microanalysis of boar and inobuta testes after the Fukushima accident

International Nuclear Information System (INIS)

Yamashiro, Hideaki; Abe, Yasuyuki; Hayashi, Gohei; Urushihara, Yusuke; Kuwahara, Yoshikazu; Suzuki, Masatoshi; Kobayashi, Jin; Kino, Yasuyuki; Fukuda, Tomokazu; Tong, Bin; Takino, Sachio; Sugano, Yukou; Sugimura, Satoshi; Yamada, Takahisa; Isogai, Emiko; Fukumoto, Manabu

2015-01-01

We aimed to investigate the effect of chronic radiation exposure associated with the Fukushima Daiichi Nuclear Power Plant (FNPP) accident on the testes of boar and inobuta (a hybrid of Sus scrofa and Sus scrofa domestica). This study examined the contamination levels of radioactive caesium (Cs), especially 134 Cs and 137 Cs, in the testis of both boar and inobuta during 2012, after the Fukushima accident. Morphological analysis and electron-probe X-ray microanalysis (EPMA) were also undertaken on the testes. The 134 Cs and 137 Cs levels were 6430 ± 23 and 6820 ± 32 Bq/kg in the boar testes, and 755 ± 13 and 747 ± 17 Bq/kg in the inobuta testes, respectively. The internal and external exposure of total 134 Cs and 137 Cs in the boar testes were 47.1 mGy and 176.2 mGy, respectively, whereas in the inobuta testes, these levels were 6.09 mGy and 59.8 mGy, respectively. Defective spermatogenesis was not detected by the histochemical analysis of radiation-exposed testes for either animal. In neither animal were Cs molecules detected, using EPMA. In conclusion, we showed that adverse radiation-induced effects were not detected in the examined boar and inobuta testes following the chronic radiation exposure associated with the FNPP accident

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Directory of Open Access Journals (Sweden)

Isabelle Guénot-Delahaie

2018-03-01

Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

2012-10-15

APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

[Theory and testing of an accident risk assessment system based on prior experience].

Science.gov (United States)

Montresor, Michele; Ricci, Paolo; Giroletti, Elio

2015-01-01

to improve the "National Project: Integrated investigations for an indepth analysis of cases of Fatal Accidents", a project which, on one hand, is too open to interpretation of events, while, on the other, does not offer the possibility to analyse external factors which are often at the basis of accidents in the workplace. identification and weighting criteria regarding causes of accident have been established and correlated by means of a specific algorithm, with the aim of making them numerically measurable. This has made it possible to use them as indicators to identify lines of priority in prevention planning. The theoretical model has been tested in an analysis of 35 work accidents which occurred in a firm in Mantova. the model has been evaluated in comparison to the analysis which was previously used to examine cases of work-related accidents and it has proved to be more efficient in the move towards establishing preventative action at the beginning of a chain of events. While maintaining the "Learning from mistakes" model, the method here proposed represents an extension and an implementation of previous practices. It is an effective operative method for companies, offering both a qualitative and quantitative analysis of work-related accidents with a view to their prevention.

An initial assessment of the Chernobyl-4 reactor accident release source

International Nuclear Information System (INIS)

Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

1986-07-01

The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

The report of the criticality accident in a uranium conversion test plant in Tokai-mura

International Nuclear Information System (INIS)

Murata, Hajime; Akashi, Makoto

2002-03-01

The criticality accident in the title occurred at around 10:35, on Sep. 30, 1999, cost the lives of two workers and caused many residents concern on their health. Moreover, rumors had both social and economic consequences. This report is a detailed account of the roles that many individuals and groups in the National Institute of Radiological Sciences (NIRS) performed in a range of the areas, and is published to discharge NIRS responsibilities in regards to the accident. The report involves chapters of detailed outline of the accident; acceptance of the victims and communications until the identification of the ''criticality'' accident; initial treatment; dose estimation (medical, hematological, physical and biological ones and that by dental metals activated by the neutron); decision making for therapeutic strategies; cooperation with the Network Council for Radiation Emergency Medicine and other medical facilities; emergency importation of medical supplies; treatment and progress (nursing system and radiation injuries); protection from radiation in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hoped to be useful in preventing the occurrence of future accidents. (K.H.)

Computerized accident management support system: development for severe accident management

International Nuclear Information System (INIS)

Garcia, V.; Saiz, J.; Gomez, C.

1998-01-01

The activities involved in the international Halden Reactor Project (HRP), sponsored by the OECD, include the development of a Computerized Accident Management Support System (CAMS). The system was initially designed for its operation under normal conditions, operational transients and non severe accidents. Its purpose is to detect the plant status, analyzing the future evolution of the sequence (initially using the APROS simulation code) and the possible recovery and mitigation actions in case of an accident occurs. In order to widen the scope of CAMS to severe accident management issues, the integration of the MAAP code in the system has been proposed, as the contribution of the Spanish Electrical Sector to the project (with the coordination of DTN). To include this new capacity in CAMS is necessary to modify the system structure, including two new modules (Diagnosis and Adjustment). These modules are being developed currently for Pressurized Water Reactors and Boiling Water REactors, by the engineering of UNION FENOSA and IBERDROLA companies (respectively). This motion presents the characteristics of the new structure of the CAMS, as well as the general characteristics of the modules, developed by these companies in the framework of the Halden Reactor Project. (Author)

Effect of Strathclyde police initiative "Operation Blade" on accident and emergency attendances due to assault.

OpenAIRE

Bleetman, A; Perry, C H; Crawford, R; Swann, I J

1997-01-01

OBJECTIVE: To review assault victim attendance at the accident and emergency department of Glasgow Royal Infirmary before and after a police initiative to curb knife carrying and tackle violent assaults ("Operation Blade"). METHODS: Assault victim attendance was reviewed for the month before the implementation of Operation Blade and for one month a year later. The number of victims requiring treatment in the resuscitation room for stab wounds before, during, and after Operation Blade was also...

Global process industry initiatives to reduce major accident hazards

Energy Technology Data Exchange (ETDEWEB)

Pitblado, Robin [DNV Energy Houston, TX (United States). SHE Risk Management; Pontes, Jose [DNV Energy Rio de Janeiro, RJ (Brazil). Americas Region; Oliveira, Luiz [DNV Energy Rio de Janeiro, RJ (Brazil)

2008-07-01

Since 2000, disasters at Texas City, Toulouse, Antwerp, Buncefield, P-36 and several near total loss events offshore in Norway have highlighted that major accident process safety is still a serious issue. Hopes that Process Safety Management or Safety Case regulations would solve these issues have not proven true. The Baker Panel recommended to BP several actions mainly around leadership, incentives, metrics, safety culture and more effective implementation of PSM systems. In Europe, an approach built around mechanical integrity and safety barriers, especially relating to technical safety systems, is being widely adopted. DNV has carried out a global survey of process industry initiatives, by interview and by literature review, for both upstream and downstream activities, to identify what the industry itself is planning to implement to enhance process safety in the next 5 - 10 years. This shows that an approach combining Baker Panel and EU barrier approaches and some nuclear industry real-time risk management approaches might be the best means to achieve a factor of 3-4 improvement in process safety. (author)

Accident sequence quantification with KIRAP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

Accident sequence quantification with KIRAP

International Nuclear Information System (INIS)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

THE ROAD ACCIDENT FUND AND SERIOUS INJURIES: THE NARRATIVE TEST

Directory of Open Access Journals (Sweden)

Magda Slabbert

2012-08-01

Full Text Available The Road Accident Fund Amendment Act 19 of 2005 came into effect on 1 August 2008. This Act limits the Road Accident Fund’s liability for compensation in respect of claims for non-pecuniary loss to instances where a “serious injury” has been sustained. A medical practitioner has to determine whether or not the claimant has suffered a serious injury by undertaking an assessment prescribed in the Regulations to the Act. The practitioner has to complete a RAF 4 report. In doing so the practitioner must assess the injury in terms of the American Medical Association’s Guides to the Evaluation of Permanent Impairment (6th ed. If the injury is considered to have resulted in less than 30 per cent of the whole person impairment the medical practitioner should apply the narrative test. The article focuses on the narrative test but also discusses reasons why the regulations do not fulfil the requirements of the Act; reasons why the Guides is not adequate to the task; the impact of the circumstances of an injured person on disability; problems with the existing wording of the narrative test; shortcomings on the RAf 4 form; the administrative process as well as the appeal tribunals.

Overview of core disruptive accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.

1977-01-01

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

Accident-resistant container: safety for warhead transport. Executive summary

International Nuclear Information System (INIS)

Berry, R.E.

1975-11-01

Development testing of model and full-scale hardware to the abnormal environments created during a cargo aircraft crash has demonstrated that the accident-resistant container (ARC) can protect an enclosed warhead from these abnormal environments. This protection reduces the probability of initiation of the warhead HE. Transfer of the plutonium limit to the ARC may permit transporting increased numbers of warheads on a single transport vehicle. Testing of one warhead configuration has been completed. Production can be initiated for transporting that system in the ARC. Other systems need test evaluation and certification before being transported in the ARC

Behavior of small-sized BWR fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

1992-01-01

The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

Improvement in post test accident analysis results prediction for the test no. 2 in PSB test facility by applying UMAE methodology

International Nuclear Information System (INIS)

Dubey, S.K.; Petruzzi, A.; Giannotti, W.; D'Auria, F.

2006-01-01

This paper mainly deals with the improvement in the post test accident analysis results prediction for the test no. 2, 'Total loss of feed water with failure of HPIS pumps and operator actions on primary and secondary circuit depressurization', carried-out on PSB integral test facility in May 2005. This is one the most complicated test conducted in PSB test facility. The prime objective of this test is to provide support for the verification of the accident management strategies for NPPs and also to verify the correctness of some safety systems operating only during accident. The objective of this analysis is to assess the capability to reproduce the phenomena occurring during the selected tests and to quantify the accuracy of the code calculation qualitatively and quantitatively for the best estimate code Relap5/mod3.3 by systematically applying all the procedures lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE), developed at University of Pisa. In order to achieve these objectives test facility nodalisation qualification for both 'steady state level' and 'on transient level' are demonstrated. For the 'steady state level' qualification compliance to acceptance criteria established in UMAE has been checked for geometrical details and thermal hydraulic parameters. The following steps have been performed for evaluation of qualitative qualification of 'on transient level': visual comparisons between experimental and calculated relevant parameters time trends; list of comparison between experimental and code calculation resulting time sequence of significant events; identification/verification of CSNI phenomena validation matrix; use of the Phenomenological Windows (PhW), identification of Key Phenomena and Relevant Thermal-hydraulic Aspects (RTA). A successful application of the qualitative process constitutes a prerequisite to the application of the quantitative analysis. For quantitative accuracy of code prediction Fast Fourier Transform Based

Nuclear Reactor RA Safety Report, Vol. 16, Maximum hypothetical accident

International Nuclear Information System (INIS)

1986-11-01

Fault tree analysis of the maximum hypothetical accident covers the basic elements: accident initiation, phase development phases - scheme of possible accident flow. Cause of the accident initiation is the break of primary cooling pipe, heavy water system. Loss of primary coolant causes loss of pressure in the primary circuit at the coolant input in the reactor vessel. This initiates safety protection system which should automatically shutdown the reactor. Separate chapters are devoted to: after-heat removal, coolant and moderator loss; accident effects on the reactor core, effects in the reactor building, and release of radioactive wastes [sr

Categorization of PWR accident sequences and guidelines for fault trees: seismic initiators

International Nuclear Information System (INIS)

Kimura, C.Y.

1984-09-01

This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to determine the applicability of the dominant accident sequences of Zion 1 to other PWR plants. The functional event trees were developed to describe the system functions that must work or not work in order for a certain accident sequence to happen, one for pipe breaks and one for transients

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

Energy Technology Data Exchange (ETDEWEB)

Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

Severe accident analysis methodology in support of accident management

International Nuclear Information System (INIS)

Boesmans, B.; Auglaire, M.; Snoeck, J.

1997-01-01

The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

Testing of an accident consequence assessment model using field data

International Nuclear Information System (INIS)

Homma, Toshimitsu; Matsubara, Takeshi; Tomita, Kenichi

2007-01-01

This paper presents the results obtained from the application of an accident consequence assessment model, OSCAAR to the Iput dose reconstruction scenario of BIOMASS and also to the Chernobyl 131 I fallout scenario of EMRAS, both organized by International Atomic Energy Agency. The Iput Scenario deals with 137 Cs contamination of the catchment basin and agricultural area in the Bryansk Region of Russia, which was heavily contaminated after the Chernobyl accident. This exercise was used to test the chronic exposure pathway models in OSCAAR with actual measurements and to identify the most important sources of uncertainty with respect to each part of the assessment. The OSCAAR chronic exposure pathway models had some limitations but the refined model, COLINA almost successfully reconstructed the whole 10-year time course of 137 Cs activity concentrations in most requested types of agricultural products and natural foodstuffs. The Plavsk scenario provides a good opportunity to test not only the food chain transfer model of 131 I but also the method of assessing 131 I thyroid burden. OSCAAR showed in general good capabilities for assessing the important 131 I exposure pathways. (author)

High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

Reactivity Initiated Accident Test Series: Test RIA 1-2. Quick look report

International Nuclear Information System (INIS)

Martinson, Z.R.; Semken, R.S.; Smith, R.H.; Osetek, D.J.

1978-12-01

The primary objectives of Test RIA 1-2 were to (a) characterize the response of preirradiated fuel rods during an RIA event conducted at boiling water reactor (BWR) hot-startup conditions for an axial peak pellet surface energy of 200 cal/g UO 2 , and (b) evaluate the effect of internal rod pressure on preirradiated fuel rod response during an RIA event. The test consisted of four, individually shrouded, pressurized water reactor-type fuel rods previously irradiated to burnups of about 4800 MWd/t. In addition to the power calibration and preconditioning, the fuel rods were subjected to a single power burst that deposited a total pellet surface energy of approximately 200 cal/gm UO 2 at the axial peak power location (estimated using the core power chambers to relate steady state and transient powers). The test data indicate that the two irradiated fuel rods prepressurized to 2.41 MPa did not fail. FRAP-T4 calculations had predicted that prompt cladding rupture would occur for pellet surface energy depositions of 206 cal/g or greater. Although the two fuel rods prepressurized to 2.41 MPa did not fail, the data indicate that at least one of the two fuel rods prepressurized to 0.1 MPa did fail. Based on the core power chamber data, this rod failure indicates a threshold for the preirradiated fuel rods near or below 200 cal/g UO 2 total pellet surface energy at the axial flux peak

Instrumentation Performance during the TMI-2 Accident

International Nuclear Information System (INIS)

Rempe, Joy L.; Knudson, Darrell L.

2013-06-01

The accident at the Three Mile Island Unit 2 (TMI- 2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts. (authors)

Expert software for accident identification

International Nuclear Information System (INIS)

Dobnikar, M.; Nemec, T.; Muehleisen, A.

2003-01-01

Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

Monitoring severe accidents using AI techniques

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Institute of Nuclear Nonproliferation and Control, Daejon (Korea, Republic of)

2012-05-15

After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

Monitoring severe accidents using AI techniques

International Nuclear Information System (INIS)

No, Young Gyu; Ahn, Kwang Il; Kim, Ju Hyun; Na, Man Gyun; Lim, Dong Hyuk

2012-01-01

After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

Probability of spent fuel transportation accidents

International Nuclear Information System (INIS)

McClure, J.D.

1981-07-01

The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

Development of an Accident Diagnostic Scheme Using Artificial Intelligence Techniques (I)

Energy Technology Data Exchange (ETDEWEB)

Na, M. G.; Lee, S. H.; Kim, D. S.; No, Y. G.; Lee, S. W. [Chosun University, Gwangju (Korea, Republic of); Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-06-15

As a means to effectively manage the severe nuclear accidents, it is important to identify and diagnose the accident initiating events during an initial short time interval after the accidents by observing the major controlling parameters. Main objective of this study is to develop the diagnostic approach for the accurate prediction of accident initiating events using artificial intelligence techniques. For this, first, a variety of artificial intelligence techniques such as Finn, Gmbh, and Sm were examined through this study. Among them, Sc and Gmbh model were assessed as a useful approach to predict the break location and the break size of Local. In order to verify the proposed algorithm, the 111 accident simulation data (based on Map) were applied to train the Sc and Gmbh models, and the test data was used to independently verify whether or not the SVC and GMDH models work well. The analysis of the maximum errors and RMS errors, and the performance of the GMDH according to the existence of measurement errors and SIS actuation showed that the proposed SVC and GMDH models can accurately classify the break locations and accurately predict the break size. As the time-integrated signals were used for inputs into the GMDH model within a period of 60 second after a reactor scram, the actuation of the safety systems such as safety injection system (SIS), auxiliary feed water system, and containment spray system, were not considered in this study. It is because the initial 60 second time-integrated signals were used and the safety systems usually start to actuate after a more than 60 second time delay after the reactor scram

Development of an Accident Diagnostic Scheme Using Artificial Intelligence Techniques (I)

International Nuclear Information System (INIS)

Na, M. G.; Lee, S. H.; Kim, D. S.; No, Y. G.; Lee, S. W.; Ahn, K. I.

2010-06-01

As a means to effectively manage the severe nuclear accidents, it is important to identify and diagnose the accident initiating events during an initial short time interval after the accidents by observing the major controlling parameters. Main objective of this study is to develop the diagnostic approach for the accurate prediction of accident initiating events using artificial intelligence techniques. For this, first, a variety of artificial intelligence techniques such as Finn, Gmbh, and Sm were examined through this study. Among them, Sc and Gmbh model were assessed as a useful approach to predict the break location and the break size of Local. In order to verify the proposed algorithm, the 111 accident simulation data (based on Map) were applied to train the Sc and Gmbh models, and the test data was used to independently verify whether or not the SVC and GMDH models work well. The analysis of the maximum errors and RMS errors, and the performance of the GMDH according to the existence of measurement errors and SIS actuation showed that the proposed SVC and GMDH models can accurately classify the break locations and accurately predict the break size. As the time-integrated signals were used for inputs into the GMDH model within a period of 60 second after a reactor scram, the actuation of the safety systems such as safety injection system (SIS), auxiliary feed water system, and containment spray system, were not considered in this study. It is because the initial 60 second time-integrated signals were used and the safety systems usually start to actuate after a more than 60 second time delay after the reactor scram

Epidemiological and immunological studies of radiation accidents and nucleare tests participants

International Nuclear Information System (INIS)

Shubik, V. M.; Bronstein, I. E.; Koroleva, T.M.; Strelnicova, T.M.; Sukalskay, S. J.

2004-01-01

Results of long term studies of epidemiological and immunological problems after radiation accidents in Ural. At Chernobyl and nuclear weapons tests in Semi-palatinsk and Novaya Zemlya nuclear tests sites are presented. Changes in Health and immunity status of emergency team workers (liquida-tors) and participants on nuclear weapon tests were recorded in long term studies af-ter 10 and more years after radiation exposure. Some changes (decrease in ly-sozyme activity, disimmunoglobulinemia) could be attributed to the old age of exam-ined persons and concomitant cardiovasculatory, respiratory and other diseases An-other ones were related to the autoimmune syndromes. Humoral and cellular auto-immune changes were more pronounced in liquidators and participants then in controls. concentrations of antitissue antibodies in exposed cohort was three times higher than in control. Level of antibodies to thyroid antigens (microsoms and thy-roglobulines) were five times higher in liquidators of Chernobyl accident. The pos-sible role of humoral and cell autoimmune changes in the development of cardiovascular, liver, kidney and thyroid is considered. Considerable increase in some cytocine concentrations in blood of participants was found. For example increased concentration of TNF was recorded in half of par-ticipants from Novaya Zemlya in comparison to similar changes in only twenty pro-cents of controls. In half of participants from Semipalatinsk site the virus antigens in epithelium of higher respiratory tract (mostly adenoviruses) were found, with 22% in control group. In health and immunity studies of population from the contaminated areas after accidents and nuclear tests (Ural, Bryansk, Russian arktics) the demographics changes, mortality structure changes, oncological mortality and immunological deficiencies were found. The recorded effects might by considered as a results of combined effect of ra-diological and non-radiological factors. The potentiated effect of chronic

A comparison of the hazard perception ability of accident-involved and accident-free motorcycle riders.

Science.gov (United States)

Cheng, Andy S K; Ng, Terry C K; Lee, Hoe C

2011-07-01

Hazard perception is the ability to read the road and is closely related to involvement in traffic accidents. It consists of both cognitive and behavioral components. Within the cognitive component, visual attention is an important function of driving whereas driving behavior, which represents the behavioral component, can affect the hazard perception of the driver. Motorcycle riders are the most vulnerable types of road user. The primary purpose of this study was to deepen our understanding of the correlation of different subtypes of visual attention and driving violation behaviors and their effect on hazard perception between accident-free and accident-involved motorcycle riders. Sixty-three accident-free and 46 accident-involved motorcycle riders undertook four neuropsychological tests of attention (Digit Vigilance Test, Color Trails Test-1, Color Trails Test-2, and Symbol Digit Modalities Test), filled out the Chinese Motorcycle Rider Driving Violation (CMRDV) Questionnaire, and viewed a road-user-based hazard situation with an eye-tracking system to record the response latencies to potentially dangerous traffic situations. The results showed that both the divided and selective attention of accident-involved motorcycle riders were significantly inferior to those of accident-free motorcycle riders, and that accident-involved riders exhibited significantly higher driving violation behaviors and took longer to identify hazardous situations compared to their accident-free counterparts. However, the results of the regression analysis showed that aggressive driving violation CMRDV score significantly predicted hazard perception and accident involvement of motorcycle riders. Given that all participants were mature and experienced motorcycle riders, the most plausible explanation for the differences between them is their driving style (influenced by an undesirable driving attitude), rather than skill deficits per se. The present study points to the importance of

Documentation for initial testing and inspections of Beneficial Uses Shipping System (BUSS) Cask

International Nuclear Information System (INIS)

Lundeen, J.E.

1994-01-01

The purpose of this report is to compile data generated during the initial tests and inspections of the Beneficial Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in section 8.1 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The BUSS Cask body and lid are each one-piece forgings fabricated from ASTM A473, Type 304 stainless steel. The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Chapter 8 of the BUSS Cask SARP requires several acceptance tests and inspections, each intended to evaluate the performance of different components of the BUSS Cask system, to be performed before its first use. The results of the tests and inspections required are included in this document

Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

International Nuclear Information System (INIS)

Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I.; Elkin, I.V.

2001-01-01

The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG and begins to remove

Fukushima-Daiichi after the severe accident (estimation)

International Nuclear Information System (INIS)

Otcenasek, Petr

2011-01-01

All facts about the Fukushima-Daiichi NPP and about the accident known at the time of publication are summarized and expected remedial actions and consequences of the accidents are deduced. The paper is structured as follows: (1) Accident initiation is known; (2) Logically inferred results; (3) Framework identification; (4) Survey; and (5) Economic and strategic impacts of the accident. Worldwide solidarity is mentioned in conclusion. (P.A.)

Occupational accidents aboard merchant ships

DEFF Research Database (Denmark)

Hansen, H.L.; Nielsen, D.; Frydenberg, Morten

2002-01-01

Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may...... be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were...... aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded...

Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

International Nuclear Information System (INIS)

Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

2015-01-01

The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

The influence of simultaneous or sequential test conditions in the properties of industrial polymers, submitted to PWR accident simulations

International Nuclear Information System (INIS)

Carlin, F.; Alba, C.; Chenion, J.; Gaussens, G.; Henry, J.Y.

1986-10-01

The effect of PWR plant normal and accident operating conditions on polymers forms the basis of nuclear qualification of safety-related containment equipment. This study was carried out on the request of safety organizations. Its purpose was to check whether accident simulations carried out sequentially during equipment qualification tests would lead to the same deterioration as that caused by an accident involving simultaneous irradiation and thermodynamic effects. The IPSN, DAS and the United States NRC have collaborated in preparing this study. The work carried out by ORIS Company as well as the results obtained from measurement of the mechanical properties of 8 industrial polymers are described in this report. The results are given in the conclusion. They tend to show that, overall, the most suitable test cycle for simulating accident operating conditions would be one which included irradiation and consecutive thermodynamic shock. The results of this study and the results obtained in a previous study, which included the same test cycles, except for more severe thermo-ageing, have been compared. This comparison, which was made on three elastomers, shows that ageing after the accident has a different effect on each material [fr

LMFBR accident delineation study: approach and preliminary results

International Nuclear Information System (INIS)

Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

1979-01-01

Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

Relevance of IAEA tests to severe accidents in nuclear fuel cycle transport

International Nuclear Information System (INIS)

Wilkinson, W.L.

2004-01-01

The design and performance standards for packages used for the transport of nuclear fuel cycle materials, are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials, TS-R-1, in order to ensure safety under both normal and accident conditions of transport. The underlying philosophy is that safety is vested principally in the package and the design and performance criteria are related to the potential hazard. Type B packages are high duty packages which are used for the transport of the more radioactive materials, notably spent fuel and vitrified high-level waste (VHLW). Tests are specified in the IAEA Regulations to ensure the integrity of these packages in potential transport accidents involving impacts, fires or immersion in water. The mechanical tests for Type B packages include drop tests onto an unyielding surface without giving rise to a significant release of radioactivity. The objects which a package could impact in real life transport accidents, such as concrete roads, bridge abutments and piers, will yield to some extent and absorb some of the energy of the moving package. Impact tests onto an unyielding surface are therefore relevant to impacts onto real-life objects at much higher speeds. The thermal test specifies that Type B packages should be able to withstand a fully engulfing fire of 8000 C for 30 minutes. Analytical studies backed up by experimental tests have shown that these packages can withstand such conditions without significant release of radioactivity. The Regulations also specify immersion tests for Type B packages; 15 metres for 8 hours without significant release of radioactivity and, in addition for spent fuel and VHLW packages, 200 metres for 1 hour without rupture of the containment. Studies have shown that spent fuel and VHLW casks would meet these conditions. Therefore, there is a large body of evidence to show that the current IAEA Type B test requirements are severe and cover all the situations which can

Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

2016-04-15

Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

Accident management insights after the Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

2014-01-01

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

Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

1981-05-01

This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

Test Data for USEPR Severe Accident Code Validation

Energy Technology Data Exchange (ETDEWEB)

J. L. Rempe

2007-05-01

This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

Benchmarking MARS (accident management software) with the Browns Ferry fire

International Nuclear Information System (INIS)

Dawson, S.M.; Liu, L.Y.; Raines, J.C.

1992-01-01

The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS code in simulatng a plant transient, MARS is being benchmarked with the available reactor pressure vessel (RPV) pressure and level data from the Browns Ferry fire. The MRS software uses the Modular Accident Analysis Program (MAAP) code as its basis to calculate plant response under accident conditions. MARS uses a limited set of plant data to initialize and track the accidnt progression. To perform this benchmark, a simulated set of plant data was constructed based on actual report data containing the information necessary to initialize MARS and keep track of plant system status throughout the accident progression. The initial Browns Ferry fire data were produced by performing a MAAP run to simulate the accident. The remaining accident simulation used actual plant data

Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

Essay on the pertinence of Luscher's abbreviate test in psychological evaluation of the radioactive accident victims of Goiania

International Nuclear Information System (INIS)

Costa Neto, Sebastiao Benicio da

1995-01-01

The essay on the pertinence of Luscher's abbreviate test in psychological evaluation of the radioactive accident victims of Goiania - Brazilian city - occurred in 1987 is consequence of confront of data obtained in two distinct situations having for criterion: time, efficiency and pertinence. Besides of this, they are introduced palografic and the house-tree-person - HTP - tests. These tests aimed at the common psychological characteristics verification to radioactive accident victims' personality of Goiania and to the data existential moment for those people. Among the three tests, the one of Luscher was what obtained the best interviewees acceptance index

Initial medical management of criticality accident victim; Conduite a tenir aux victimes d'un accident de criticite

Energy Technology Data Exchange (ETDEWEB)

Miele, A; Bebaron-Jacobs, L

2005-07-01

The extremely severe criticality accidents known to this day, and the subsequent deaths recorded (Sarov 1997 and Tokai Mura 1999), demonstrate the need for sustained surveillance and constant adapted training for the teams in charge of irradiated and/or contaminated victims. The aim of this work group, composed of occupational health services and associated medical biology laboratories, is to present, in leaflet format, the essential data on the documentation and the conduct to be held when facing the victims of a criticality accident. The studies of this work group confirm the difficulties involved in managing this type of accident, both from the dosimetric evaluation point of view and from the therapeutic management point of view. That is why several research themes and perspectives are developed. During the different phases of victim triage, the recommendations given on these leaflets describe the operational conducts to be held. This work will have to be updated according to the evolution in knowledge and means: short and long term effects of exposure to neutrons, multi-competence hospital cooperation, expertise networks related to dosimetric reconstitution. (authors)

The consequences of the Chernobyl accident - the radioecological database Redac of the French-German initiative

Energy Technology Data Exchange (ETDEWEB)

Deville-Cavelin, G.; Biesold, H.; Chabanyuk, V. [Radioprotection and Nuclear Safety Institute (IRSN), Dir. of Environment and Intervention (DEI) - CEA Cadarache, 13 - Saint-Paul-lez-Durance (France)

2004-07-01

The French-German Initiative for Chernobyl (FGI), implemented by IRSN and GRS from 1997 until the end of 2003, included the 'Project on the Radioecological Consequences of the Accident'. The most relevant fields of radioecology and post-accidental aspects have been studied, such as radionuclides transfers to plants, to animals, by surface runoff, in the aquatic environment and in the urban environment, wastes management and countermeasures. The main goal was to collect and harmonise, from Belarus, Russia and Ukraine, the highest possible amount of data and results on these different topics. These data have been verified, validated and organized in a common geo-referenced database REDAC (Radioecological Database After Chernobyl). For linking the different data, maps of initial and present contamination by {sup 137}Cs and {sup 90}Sr have been drawn up and relevant environmental non-radioactive data have been included. The operational database built will also allow the management of the wastes disposal sites. Countermeasures used after the accident for urban areas, natural and agricultural environment, have been described and classified. A methodology for evaluating their effectiveness has been developed. This database constitutes a tool for the development and validation of operational, assessment and explicative models. This allows the quantification and assessment of radionuclide transfer in the different compartments of ecosystems. So the main parameters influencing the transfers can be identified. REDAC should be completed by further investigations, for example on transuranic elements and extended to larger geographical zones. The database should also be combined with others provided by different organisations (IAEA, IRSN, UIR, ). (author)

The consequences of the Chernobyl accident - the radioecological database Redac of the French-German initiative

International Nuclear Information System (INIS)

Deville-Cavelin, G.; Biesold, H.; Chabanyuk, V.

2004-01-01

The French-German Initiative for Chernobyl (FGI), implemented by IRSN and GRS from 1997 until the end of 2003, included the 'Project on the Radioecological Consequences of the Accident'. The most relevant fields of radioecology and post-accidental aspects have been studied, such as radionuclides transfers to plants, to animals, by surface runoff, in the aquatic environment and in the urban environment, wastes management and countermeasures. The main goal was to collect and harmonise, from Belarus, Russia and Ukraine, the highest possible amount of data and results on these different topics. These data have been verified, validated and organized in a common geo-referenced database REDAC (Radioecological Database After Chernobyl). For linking the different data, maps of initial and present contamination by 137 Cs and 90 Sr have been drawn up and relevant environmental non-radioactive data have been included. The operational database built will also allow the management of the wastes disposal sites. Countermeasures used after the accident for urban areas, natural and agricultural environment, have been described and classified. A methodology for evaluating their effectiveness has been developed. This database constitutes a tool for the development and validation of operational, assessment and explicative models. This allows the quantification and assessment of radionuclide transfer in the different compartments of ecosystems. So the main parameters influencing the transfers can be identified. REDAC should be completed by further investigations, for example on transuranic elements and extended to larger geographical zones. The database should also be combined with others provided by different organisations (IAEA, IRSN, UIR, ). (author)

Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

Energy Technology Data Exchange (ETDEWEB)

Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I. [Electrogorsk Research and Engineering Center, EREC, Moscow (Russian Federation); Elkin, I.V. [RRC ' ' Kurchatov Institute, Moscow (Russian Federation)

2001-07-01

The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG

Psychometric testing of children prenatally irradiated during the Chernobyl accident

International Nuclear Information System (INIS)

Bajrakova, A.; Vasilev, G.; Khristova, M. N.; Chobanova, N.; Tsenova, T.; Jordanova, M.; Lalova, J.; Vasileva, F.; Mikhajlova, Z.; Trifonova, S.

1993-01-01

The investigation involved 50 children aged median 6 years and 6 months. The group was selected in view of the critical period for occurrence of radiation-related deviations in mental development (8-15 gestation weeks) and the period of maximum irradiation during the Chernobyl accident. Assessment of the individual exposure and analysis of possible impacts from non-radiation risk factors were based on guided parental history reports. The dose of accidental irradiation was determined using the radiological data for the country. A Bulgarian standardization of the Wechsler Intelligence Scale for Children (WISC-R) was used. The procedure includes 5 verbal and 5 nonverbal subtests. Results were compared with those from a countrywide control group of children (including a large city, a small town, a village). The analysis indicated higher mean IQ scores in the investigated children. The children were additionally studied by original tests for attention and gnosis-praxis functions using tactile and visual modalities. The tests included intra- and transmodal versions, bilateral simultaneous presentation of stimuli with verbal and nonverbal characteristics in applying analytical and global strategies. Comparisons were made with results for children in the same age range, who had been studied prior to the Chernobyl accident. The evidence surprisingly varied, taking into account the small size of the investigation group. A longitudinal follow-up of this population thus appears to be appropriate. (author)

Safety in the Chemical Laboratory. Epidemiology of Accidents in Academic Chemistry Laboratories, Part 2. Accident Intervention Study, Legal Aspects, and Observations.

Science.gov (United States)

Hellmann, Margaret A.; And Others

1986-01-01

Reports on a chemistry laboratory accident intervention study conducted throughout the state of Colorado. Addresses the results of an initial survey of institutions of higher learning. Discusses some legal aspects concerning academic chemistry accidents. Provides some observations about academic chemistry laboratory accidents on the whole. (TW)

Overview of severe accident research at the USNRC

International Nuclear Information System (INIS)

Basu, S.; Ader, C.E.

1999-01-01

This paper summarizes the U.S. Nuclear Regulatory Commission's (USNRC) severe accident research activities, in particular, progress made in the past year toward the resolution and/or improved understanding of a number of severe accident issues. The direct containment heating (DCH) is nearing resolution for Combustion Engineering and Babcock and Wilcox type pressurized water reactors (PWRs) are well as for ice condensers. Additionally, two lower pressure DCH tests were conducted recently at the Sandia National Laboratories (SNL) under the NRC/IPSN/FzK sponsorship to provide data regarding intentional depressurization as an accident management strategy to mitigate DCH loads. In the area of lower head integrity, the experimental program to investigate boiling heat transfer on downward facing curved surfaces with insulation was completed. Finally, the SNL program investigating the creep rupture behavior of the lower head under the combined thermo-mechanical loading was completed recently. Additional lower head experiments at SNL are being planned as an OECD project. During the past year, the USNRC participated in two programs aimed at extending the data base on hydrogen combustion into more prototypic situations. Testing was performed at the Brookhaven National Laboratory (BNL) to investigate detonation transmission at elevated temperatures. In a cooperative program under the sponsorship of NRC/IPSN/FzK, Russian Research Center (RRC) investigated hydrogen combustion issues at large scale at the RUT facility. The experimental program at the SNL to examine the performance of Passive Autocatalytic Recombiners (PARs) was completed also this year. In the fuel-coolant interaction (FCI) area, the experimental work at the Argonne National Laboratory (ANL) to investigate chemical augmentation of FCI energetics was completed as was the experimental work at the University of Wisconsin (UW) involving one-dimensional propagation experiments (similar to KROTOS). The USNRC is

The 1986 Chernobyl accident; Der Unfall von Tschernobyl 1986

Energy Technology Data Exchange (ETDEWEB)

Kerner, Alexander; Stueck, Reinhard; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching bei Muenchen, Koeln (Germany). Bereich Reaktorsicherheitsanalysen; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

2011-02-15

April 26, 2011 marks the 25th anniversary of the Chernobyl reactor accident, the worst incident in the history of the peaceful utilization of nuclear power. While investigations of the course of events and the causes of the accident largely present a uniform picture, descriptions still vary widely when it comes to the impact on the population and the environment. This treatment of the Chernobyl accident constitutes a summary of facts about the initiation of the accident and the sequence of events that followed. In addition, measures are described which were taken to exclude any repetition of a disaster of this kind. The health consequences and the socio-economic impact of the accident are not discussed in any detail. The first section contains an introduction and an overview of the Soviet RBMK (Chernobyl) reactor line. In section 2, fundamental characteristics of this special type of reactor, which was exclusively built in the former Soviet Union, are discussed. This information is necessary to understand the sequence of accident events and provides an answer to the frequent question whether that accident could be transferred to reactors in this country. The third section outlines the history of the accident caused ultimately by a commissioning test never performed before. The section is completed by a brief description of radiological releases and the state of the plant after the accident when entombed in the ''sarcophagus.'' The different causes are then summarized and the modifications afterwards made to RBMK reactors are outlined. (orig.)

Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

International Nuclear Information System (INIS)

Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

1996-01-01

The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

Review of specific radiological accident considerations

International Nuclear Information System (INIS)

Elder, J.

1984-01-01

Specific points of guidance provided in the forthcoming document A Guide to Radiological Accident Considerations for Siting and Design of Nonreactor Nuclear Facilities are discussed. Of these, the following are considered of particular interest to analysts of hypothetical accidents: onsite dose limits; population dose, public health effects, and environmental contamination as accident consequences which should be addressed; risk analysis; natural phenomena as accident initiators; recommended dose models; multiple organ equivalent dose; and recommended methods and parameters for source terms and release amount calculations. Comments are being invited on this document, which is undergoing rewrite after the first stage of peer review

ASSESSMENT OF RELIABILITY AND RISK DEGREE FOR ACCIDENT INITIATION AT SLIME STORAGES OF 4th MINING ADMINISTRATION, JSC “BELARUSKALI”

Directory of Open Access Journals (Sweden)

P. M. Bohaslauchyk

2016-01-01

Full Text Available Definition of reliability for dams of slime storage embankment is given on the basis of reliability theory and characteristics of reliability and their analysis are presented in the paper. The paper specifies qualitative indices for earth dams which are subdivided in two groups: applicability factors and structural reliability factors. A short analysis of all possible causes for accident initiation at earth dams has been made and the analysis has permitted to pinpoint eleven main objects for diagnosis for slime storage dams. In order to assess risk degree of accident initiation at JSC “Belaruskali” slime storages all possible causes of emergency cases and their probability of occurrence have been analyzed in the paper. The paper acknowledges the fact that dam malfunction is possible, as a rule, due to violation of operational rules and regulations. Main parameters of slime storage state which are to be controlled regularly in the process of its operation have been noted in the paper. Observation results over slime storages, calculations of dam slope stability for normal operation (a principal calculation case and operating irregularities in water seals (a special calculation case. As a stability margin factor is close to 1.0 for a special calculation case, an extreme position of depression curve has been determined for all design sections. It has been recommended to carry out a constant control over its position, and in the case when it reaches its peak value it is necessary to undertake appropriate measures in order to reduce its value. Final expert estimations on probability of accident initiation at the investigated slime storage dams of the 4th Mining Administration, JSC “Belaruskali” have been prepared on the basis of the analysis comprising all the required factors. A conclusion has been made about low risk degree of their destruction.

A neutron dosemeter for nuclear criticality accidents.

Science.gov (United States)

d'Errico, F; Curzio, G; Ciolini, R; Del Gratta, A; Nath, R

2004-01-01

A neutron dosemeter which offers instant read-out has been developed for nuclear criticality accidents. The system is based on gels containing emulsions of superheated dichlorodifluoromethane droplets, which vaporise into bubbles upon neutron irradiation. The expansion of these bubbles displaces an equivalent volume of gel into a graduated pipette, providing an immediate measure of the dose. Instant read-out is achieved using an array of transmissive optical sensors which consist of coupled LED emitters and phototransistor receivers. When the gel displaced in the pipette crosses the sensing region of the photomicrosensors, it generates a signal collected on a computer through a dedicated acquisition board. The performance of the device was tested during the 2002 International Accident Dosimetry Intercomparison in Valduc, France. The dosemeter was able to follow the initial dose gradient of a simulated accident, providing accurate values of neutron kerma; however, the emulsion was rapidly depleted of all its drops. A model of the depletion effects was developed and it indicates that an adequate dynamic range of the dose response can be achieved by using emulsions of smaller droplets.

A neutron dosemeter for nuclear criticality accidents

International Nuclear Information System (INIS)

D'Errico, F.; Curzio, G.; Ciolini, R.; Del Gratta, A.; Nath, R.

2004-01-01

A neutron dosemeter which offers instant read-out has been developed for nuclear criticality accidents. The system is based on gels containing emulsions of superheated dichlorodifluoromethane droplets, which vaporise into bubbles upon neutron irradiation. The expansion of these bubbles displaces an equivalent volume of gel into a graduated pipette, providing an immediate measure of the dose. Instant read-out is achieved using an array of transmissive optical sensors which consist of coupled LED emitters and phototransistor receivers. When the gel displaced in the pipette crosses the sensing region of the photo microsensors, it generates a signal collected on a computer through a dedicated acquisition board. The performance of the device was tested during the 2002 International Accident Dosimetry Intercomparison in Valduc (France)). The dosemeter was able to follow the initial dose gradient of a simulated accident, providing accurate values of neutron kerma; however, the emulsion was rapidly depleted of all its drops. A model of the depletion effects was developed and it indicates that an adequate dynamic range of the dose response can be achieved by using emulsions of smaller droplets. (authors)

Occupational accidents among mototaxi drivers.

Science.gov (United States)

Amorim, Camila Rego; de Araújo, Edna Maria; de Araújo, Tânia Maria; de Oliveira, Nelson Fernandes

2012-03-01

The use of motorcycles as a means of work has contributed to the increase in traffic accidents, in particular, mototaxi accidents. The aim of this study was to estimate and characterize the incidence of occupational accidents among the mototaxis registered in Feira de Santana, BA. This is a cross-sectional study with descriptive and census data. Of the 300 professionals registered at the Municipal Transportation Service, 267 professionals were interviewed through a structured questionnaire. Then, a descriptive analysis was conducted and the incidence of accidents was estimated based on the variables studied. Relative risks were calculated and statistical significance was determined using the chi-square test and Fisher's exact test, considering p accidents were observed in 10.5% of mototaxis. There were mainly minor injuries (48.7%), 27% of them requiring leaves of absence from work. There was an association between the days of work per week, fatigue in lower limbs and musculoskeletal complaints, and accidents. Knowledge of the working conditions and accidents involved in this activity can be of great importance for the adoption of traffic education policies, and to help prevent accidents by improving the working conditions and lives of these professionals.

Accident situations tests HTR fuel with the device Kufa

International Nuclear Information System (INIS)

Kellerbauer, A. I.; Freis, D.

2010-01-01

The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

Resolve. Version 2.5: Flammable Gas Accident Analysis Tool Acceptance Test Plan and Test Results

International Nuclear Information System (INIS)

LAVENDER, J.C.

2000-01-01

RESOLVE. Version 2 .5 is designed to quantify the risk and uncertainty of combustion accidents in double-shell tanks (DSTs) and single-shell tanks (SSTs). The purpose of the acceptance testing is to ensure that all of the options and features of the computer code run; to verify that the calculated results are consistent with each other; and to evaluate the effects of the changes to the parameter values on the frequency and consequence trends associated with flammable gas deflagrations or detonations

Generic implications of the Chernobyl accident

International Nuclear Information System (INIS)

Sege, G.

1989-01-01

The US Nuclear Regulatory Commission (NRC) staff's assessment of the generic implications of the Chernobyl accident led to the conclusion that no immediate changes in the NRC's regulations regarding design or operation of US commercial reactors are needed. However, further consideration of certain issues was recommended. This paper discusses those issues and the studies being addressed to them. Although 24 tasks relating to light water reactor issues are identified in the Chernobyl follow-up research program, only four are new initiatives originating from Chernobyl implications. The remainder are limited modifications of ongoing programs designed to ensure that those programs duly reflect any lessons that may be drawn from the Chernobyl experience. The four new study tasks discussed include a study of reactivity transients, to reconfirm or bring into question the adequacy of potential reactivity accident sequences hitherto selected as a basis for design approvals; analysis of risk at low power and shutdown; a study of procedure violations; and a review of current NRC testing requirements for balance of benefits and risks. Also discussed, briefly, are adjustments to ongoing studies in the areas of operational controls, design, containment, emergency planning, and severe accident phenomena

Radiation protection survey of research and development activities initiated after the Chernobyl accident. Review report

International Nuclear Information System (INIS)

Burkart, W.

1989-01-01

The compilation of research and development activities in the various fields of radiation protection in OECD Member countries which have been undertaken or planned specifically to address open questions arising from the Chernobyl reactor accident experience shows a potential for international cooperative arrangements and/or coordination between national programmes. Both the preliminary review of the answers, which only cover a part of the relevant activities in OECD Member countries, and a computerized literature search indicate that the multidisciplinarity of the research area under consideration will call for special efforts to efficiently implement new models and new quantitative findings from the different fields of activity to provide an improved basis for emergency management and risk assessment. Further improvements could also be achieved by efforts to initiate new activities to close gaps in the programmes under way, to enhance international cooperation, and to coordinate the evaluation of the results. This preliminary review of the answers of 17 Member countries to the questionnaire on research and development activities initiated after the Chernobyl accident is not sufficient as a basis for a balanced decision on those research areas most in need for international cooperation and coordination. It may however serve as a guide for the exploration of the potential for international cooperative arrangements and/or coordination between national programmes by the CRPPH. Even at this preliminary stage, several specific activities are proposed to the NEA/OECD by Member countries. Whole body counting and the intercomparison of national data bases on the behaviour of radionuclides in the environment did attract most calls for international cooperation sponsored by the NEA

Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

International Nuclear Information System (INIS)

Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

2010-01-01

In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

Initial medical management of criticality accident victim; Conduite a tenir aux victimes d'un accident de criticite

Energy Technology Data Exchange (ETDEWEB)

Miele, A.; Bebaron-Jacobs, L

2005-07-01

The extremely severe criticality accidents known to this day, and the subsequent deaths recorded (Sarov 1997 and Tokai Mura 1999), demonstrate the need for sustained surveillance and constant adapted training for the teams in charge of irradiated and/or contaminated victims. The aim of this work group, composed of occupational health services and associated medical biology laboratories, is to present, in leaflet format, the essential data on the documentation and the conduct to be held when facing the victims of a criticality accident. The studies of this work group confirm the difficulties involved in managing this type of accident, both from the dosimetric evaluation point of view and from the therapeutic management point of view. That is why several research themes and perspectives are developed. During the different phases of victim triage, the recommendations given on these leaflets describe the operational conducts to be held. This work will have to be updated according to the evolution in knowledge and means: short and long term effects of exposure to neutrons, multi-competence hospital cooperation, expertise networks related to dosimetric reconstitution. (authors)

Pilot program: NRC severe reactor accident incident response training manual: Public protective actions: Predetermined criteria and initial actions

International Nuclear Information System (INIS)

Martin, J.A. Jr.; McKenna, T.J.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Public Protective Actions - Predetermined Criteria and Initial Actions is the fourth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume reviews public protective action criteria and objectives, their bases and implementation, and the expected public response. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

Cosyma a new programme package for accident consequence assessment

International Nuclear Information System (INIS)

Kelly, G.N.

1991-01-01

This report gives details of a new programme package for accident consequence assessment, prepared under the CEC's Maria programme (Methods for assessing the radiological impact of accidents) initiated in 1982 to review and build on the nuclear accident consequence assessment methods in use within the European Community

Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

International Nuclear Information System (INIS)

O'Kula, K.R.; Chung, D.Y.

1999-01-01

Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

Corporate Cost of Occupational Accidents

DEFF Research Database (Denmark)

Rikhardsson, Pall M.; Impgaard, M.

2004-01-01

method could be used in all of the companies without revisions. The evaluation of accident cost showed that 2/3 of the costs of occupational accidents are visible in the Danish corporate accounting systems reviewed while 1/3 is hidden from management view. The highest cost of occupational accidents......The systematic accident cost analysis (SACA) project was carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. Its focused on developing and testing a method for evaluating...... occupational costs of companies for use by occupational health and safety professionals. The method was tested in nine Danish companies within three different industry sectors and the costs of 27 selected occupational accidents in these companies were calculated. One of the main conclusions is that the SACA...

RIA type tests in ACPR

International Nuclear Information System (INIS)

Preda, Marin; Stefan, Violeta; Ancuta, Mirela; Negut, Gheorghe

2008-01-01

For a better NPP operation fuel behavior in accidental conditions (LOCA) is of great interest. Irradiation tests in ACPR can give interesting data on the CANDU fuel behavior in such kind of accidents. These data can be used for simulation, calibration and validation of fuel computer codes. The tests were accomplished in the TRIGA Annular Core Pulse Reactor (ACPR). Reactivity insertion accidents are not specific for the CANDU reactors but this type of test can contribute to a better understanding of CANDU type fuel behavior during various conditions.The tests were accomplished in the ambient pressure and temperature with fresh fuel probes. These tests gave similar results on the clad-fuel, and clad-coolant interactions which occur in LOCA accidents. The tests will be continued with instrumentation improvements what will give better statistics and information on fuel behavior in accident conditions. The structure of the paper is the following: - 1. Introduction; - 2. Irradiation device description; - 2.1. Capsule main parameters; - 2.1.1. Initial conditions; - 2.1.2. Test conditions; - 3. Tests objectives; - 4. Test of fuel; - 5. Results; - 5.1. Temperature effects; - 5.2. Pressure effects; - 6. Conclusion

Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

2015-05-15

accident is very hazardous. If the operator initiates F and B operation properly under the combined accident including TLOFW accident, the operators can prevent the core damage. Since F and B operation is last resort to prevent core damage and necessary conditions of F and B operation are very complicated, the consequence of these events should be considered in PSA model to improve emergency response capabilities under the rare events. Dynamic PSA modeling is better to estimate the effects of heading order and timing issues. Especially, dynamic PSA can model accident sequences and estimate their probabilities through integrated, time-dependent, probabilistic and deterministic models of NPPs, based on the thermal-hydraulic processes and operator behavior in accident conditions. We will develop the dynamic PSA model for the combined accident including TLOFW accident in the further study.

The Heat Flux Analysis in an Annulus Narrows Gap With Initial Temperature Variations Using HeaTiNG-01 Test Section

International Nuclear Information System (INIS)

Mulya Juarsa; Efrizon Umar; Andhang Widi Harto

2009-01-01

An experiment to understand the complexity of boiling phenomena on a narrow gap, which has occurs in severe accident at TMI-2 NPP is necessary to be done in aimed to increase the understanding of accident management. The goal of research is to obtain a heat flux and critical heat flux (CHF) value during boiling heat transfer process in a narrow gap annulus. The method of research is experimental using HeaTiNG-01 test section. The experiment has been done with heating-up heated rod until a certain initial temperature, for this experiment, three initial temperature variations was decided at 650°C, 750°C dan 850°C. Then, a cooling process in heated rod by saturated water was recorded based on temperature data changes. Temperature data was used to calculate a value of heat flux and wall superheat temperature, until the results could be defined in boiling curve. The result of this research shows that, although the initial temperature of heated rod was different, the value of CHF is almost similar with CHF average 253.7 kW/m 2 with the changes of only 4.7%. The event of boiling in a narrow gap is not included pool boiling category based on the comparison of film boiling area of the experiment to Bromley correlations. (author)

Proposal of the concept of selection of accidents that release large amounts of radioactive substances in the high temperature engineering test reactor

International Nuclear Information System (INIS)

Ono, Masato; Honda, Yuki; Takada, Shoji; Sawa, Kazuhiro

2015-01-01

In Position, construction and equipment of testing and research reactor to be subjected to the use standards for rules Article 53 (prevention of expansion of the accident to release a large amount of radioactive material) generation the frequency is a lower accident than design basis accident, when what is likely to release a large amount of radioactive material or radiation from the facility has occurred, and take the necessary measures in order to prevent the spread of the accident. There is provided a lower accident than frequency design basis accidents, for those that may release a large amount of radioactive material or radiation. (author)

Head impact in a snowboarding accident.

Science.gov (United States)

Bailly, N; Llari, M; Donnadieu, T; Masson, C; Arnoux, P J

2017-09-01

To effectively prevent sport traumatic brain injury (TBI), means of protection need to be designed and tested in relation to the reality of head impact. This study quantifies head impacts during a typical snowboarding accident to evaluate helmet standards. A snowboarder numerical model was proposed, validated against experimental data, and used to quantify the influence of accident conditions (speed, snow stiffness, morphology, and position) on head impacts (locations, velocities, and accelerations) and injury risk during snowboarding backward falls. Three hundred twenty-four scenarios were simulated: 70% presented a high risk of mild TBI (head peak acceleration >80 g) and 15% presented a high risk of severe TBI (head injury criterion >1000). Snow stiffness, speed, and snowboarder morphology were the main factors influencing head impact metrics. Mean normal head impact speed (28 ± 6 km/h) was higher than equivalent impact speed used in American standard helmet test (ASTM F2040), and mean tangential impact speed, not included in standard tests, was 13.8 (±7 km/h). In 97% of simulated impacts, the peak head acceleration was below 300 g, which is the pass/fail criteria used in standard tests. Results suggest that initial speed, impacted surface, and pass/fail criteria used in helmet standard performance tests do not fully reflect magnitude and variability of snowboarding backward-fall impacts. © 2016 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Use of simulators in severe accident management

International Nuclear Information System (INIS)

Evans, R.C.

1994-01-01

The U.S. nuclear utility industry is moving in a deliberate fashion through a coordinated industry severe accident working group to study and augment, where appropriate, the existing utility organizational and emergency planning structure to address accident and severe accident management. Full-scope simulators are used extensively to train licensed operators for their initial license examinations and continually thereafter in licensed operator requalification training and yearly examinations. The goal of the training (both initial and requalification) is to ensure that operators possess adequate knowledge, skills and abilities to prevent an event from progressing to core damage. The use of full-scope simulators in severe accident management training is in large part viewed by the industry as being premature. The working group study has not progressed to the point where the decision to employ full-scope simulators can be logically considered. It is not however premature to consider part-task or work station simulators as invaluable research tools to support the industry's study. These simulators could be employed, subject to limitations in the current state of knowledge regarding severe accident progression and phenomenological responses, in the validation and verification (V and V) of severe accident models or codes as they are developed. The U.S. nuclear utility industry has made substantial strides in the past 12 years in the accident prevention, mitigation and management arena. These strides are a product of the industry's preference for a logical and systematic approach to change. (orig.)

Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture

International Nuclear Information System (INIS)

Rassame, Somboon; Griffiths, Matthew; Yang, Jun; Lee, Doo Yong; Ju, Peng; Choi, Sung Won; Hibiki, Takashi; Ishii, Mamoru

2014-01-01

Highlights: • Basic understanding of the venting phenomena in the SP during a LOCA was obtained. • A series of experiment is carried out using the PUMA-E test facility. • Two phases of experiments, namely, an initial and a quasi-steady phase were observed. • The maximum void penetration depth was experienced during the initial phase. - Abstract: During the initial blowdown period of a Loss of Coolant Accident (LOCA), the non-condensable gas initially contained in the BWR containment is discharged to the pressure suppression chamber through the blowdown pipes. The performance of Emergency Core Cooling System (ECCS) can be degraded due to the released gas ingestion into the suction intakes of the ECCS pumps. The understanding of the relevant phenomena in the pressure suppression chamber is important in analyzing potential gas intrusion into the suction intakes of ECCS pumps. To obtain the basic understanding of the relevant phenomena and the generic data of void distribution in the pressure suppression chamber during the initial blowdown period of a LOCA, tests with various blowdown conditions were conducted using the existing Suppression Pool (SP) tank of the integral test facility, called Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility, a scaled downcomer pipe installed in the PUMA-E SP, and air discharge pipe system. Two different diameter sizes of air injection pipe (0.076 and 0.102 m), a range of air volumetric flux (7.9–24.7 m/s), initial void conditions in an air injection pipe (fully void, partially void, and fully filled with water) and different air velocity ramp rates (1.0, 1.5, and 2.0 s) are used to investigate the impact of the blowdown conditions to the void distribution in the SP. Two distinct phases of experiments, namely, an initial and a quasi-steady phase were observed. The maximum void penetration depth was experienced during the initial phase. The quasi-steady phase provided less void

On high-temperature reactor accident topology

International Nuclear Information System (INIS)

Fassbender, J.; Kroeger, W.; Wolters, J.

1981-01-01

American and German risk studies for an HTGR and independent investigations of hypothetical accident sequences led to a fundamental understanding of the topology of HTGR accident sequences. The dominating importance of core heat-up accidents was confirmed and the initiating events were identified. Complications of core heat-up accidents by air or water ingress are of minor importance for the risk, whereas the long-term development of accidents during days and weeks plays an important role for the environmental impact. The risk caused by an HTGR at a German site cannot yet be determined exactly, because no modern German HTGR design has passed a licensing procedure. Cautious estimates show that risk will appear to be substantially smaller than the LWR risk. The main reasons are the considerably reduced release of fission procucts and the slow development of core heat-up accidents leaving much time for measures which reduce the risk. (orig.) [de

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

International Nuclear Information System (INIS)

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

1996-12-01

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

Tchernobyl accident

International Nuclear Information System (INIS)

1986-06-01

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

Overview and status of the SIMMER testing program

International Nuclear Information System (INIS)

Scott, J.H.

1979-01-01

Los Alamos Scientific Laboratory has undertaken an extensive experiment analysis program to test the results of SIMMER Liquid Metal Fast Breeder Reactor (LMFBR) accident calculations. Initially, we will test the postdisassembly work-energy partition problem. The SIMMER-calculated order-of-magnitude reduction of available kinetic energy following a severe hypothetical core-disruptive accident (HCDA) can be attributed to (1) purely fluid-dynamic effects; and (2) rate-controlled effects, such as phase transitions and heat transfer. We have chosen to test separately each class of mitigator. In this paper we review the experiments initially chosen for testing of each class of mitigator and report on the status of the analyses. We enumerate several problems in SIMMER that experiment analysis has disclosed. Finally, needs for future experiments are discussed

Criticality accident:

International Nuclear Information System (INIS)

Canavese, Susana I.

2000-01-01

A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

2017-08-15

An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Directory of Open Access Journals (Sweden)

Takeshi Takeda

2017-08-01

Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Treatment initiatives after radiological accidents: TIARA first step

International Nuclear Information System (INIS)

Menetrier, F.; Berard, P.; Joussineau, S.; Stradling, N.; Hodgson, V.; List, MA.; Morcillo, W.; Paile, D.; Holt, T.; Eriksson

2006-01-01

Full text of publication follows: T.I.A.R.A. [Treatment Initiatives After Radiological Accidents] project is a consortium of 8 European partners. This project is part of the Preparatory Action on Security Research recently launched by the European Commission. The Preparatory Action is intended to reach preliminary conclusions on the needs for the security of European Union citizens before the launch of the Security Research Programme in 2007. The principal purpose of T.I.A.R.A. is to constitute a European network which will participate in enhancing the management of a crisis in the hypothesis of a malevolent dispersal of radionuclides in a public place. The main concern is to identify and define effective medical treatments for internal radioactive contamination. A preview of the state of treatment of contamination by radionuclides (especially actinides) in Europe highlights the following points: a decrease in the number of physicians with experience of treatment, a need for generalised agreement on treatment decisions and protocols, unanticipated operational issues and research into new treatments. If treatment is to be effective then several factors must be addressed and these include: firstly, the availability of effective specific treatment for the radionuclides involved, their rapid transport to and distribution of the drugs at the place of the malevolent dispersal and the easy administration of the drug even if numerous people are contaminated. The objectives of T.I.A.R.A. are threefold. First to provide straightforward guidance on dose assessment and efficacy of treatment which is readily understood by health physicists and physicians who do not have detailed knowledge and experience in radiological protection matters. Second, to foresee the operational needs for treating persons when there are mass casualties. Third, to monitor scientific and technological development on research into new treatments. Progress in all these aspects of the project will be

Severe accident risks: An assessment for five US nuclear power plants

International Nuclear Information System (INIS)

1991-01-01

This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United State. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide releases and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. This report, Volume 3, contains two appendices. Appendix D summarizes comments received, and staff responses, on the first (February 1987) draft of NUREG-1150. Appendix E provides a similar summary of comments and responses, but for the second (June 1989) version of the report

Application of FFTBM to severe accidents

International Nuclear Information System (INIS)

Prosek, A.; Leskovar, M.

2005-01-01

In Europe an initiative for the reduction of uncertainties in severe accident safety issues was initiated. Generally, the error made in predicting plant behaviour is called uncertainty, while the discrepancies between measured and calculated trends related to experimental facilities are called the accuracy of the prediction. The purpose of the work is to assess the accuracy of the calculations of the severe accident International Standard Problem ISP-46 (Phebus FPT1), performed with two versions of MELCOR 1.8.5 for validation purposes. For the quantitative assessment of calculations the improved fast Fourier transform based method (FFTBM) was used with the capability to calculate time dependent code accuracy. In addition, a new measure for the indication of the time shift between the experimental and the calculated signal was proposed. The quantitative results obtained with FFTBM confirm the qualitative conclusions made during the Jozef Stefan Institute participation in ISP-46. In general good agreement of thermal-hydraulic variables and satisfactory agreement of total releases for most radionuclide classes was obtained. The quantitative FFTBM results showed that for the Phebus FPT1 severe accident experiment the accuracy of thermal-hydraulic variables calculated with the MELCOR severe accident code is close to the accuracy of thermal-hydraulic variables for design basis accident experiments calculated with best-estimate system codes. (author)

Ex-vessel debris coolability test during severe accident (COTELS project)

International Nuclear Information System (INIS)

Ogasawara, H.

1998-01-01

The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

Supporting system in emergency response plan for nuclear material transport accidents

International Nuclear Information System (INIS)

Nakagome, Y.; Aoki, S.

1993-01-01

As aiming to provide the detailed information concerning nuclear material transport accidents and to supply it to the concerned organizations by an online computer, the Emergency Response Supporting System has been constructed in the Nuclear Safety Technology Center, Japan. The system consists of four subsystems and four data bases. By inputting initial information such as name of package and date of accident, one can obtain the appropriate initial response procedures and related information for the accident immediately. The system must be useful for protecting the public safety from nuclear material transport accidents. But, it is not expected that the system shall be used in future. (J.P.N.)

Friction testing for abnormal wet weather accident locations : all Louisiana districts for the period 1995 : technical assistance report.

Science.gov (United States)

2000-06-01

This report contains the results of friction testing conducted by the pavement/systems group of the Louisiana Transportation Research Center (LTRC) based on accidents occurring in 1995. This testing is conducted on all Louisiana locations which have ...

The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

2005-01-01

The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

Severe accident recriticality analyses (SARA)

DEFF Research Database (Denmark)

Frid, W.; Højerup, C.F.; Lindholm, I.

2001-01-01

with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

Accidents - Chernobyl accident

International Nuclear Information System (INIS)

2004-01-01

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

Driving force of PCMI failure under reactivity initiated accident conditions and influence of hydrogen embrittlement on failure limit

International Nuclear Information System (INIS)

Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi

2005-09-01

In order to clarify the driving force of PCMI (Pellet/Cladding Mechanical Interaction) failure on high burnup fuels and to investigate the influence of hydrogen embrittlement on failure limit under RIA (Reactivity Initiated Accident) conditions, RIA-simulation experiments were performed on fresh fuel rods in the NSRR (Nuclear Safety Research Reactor). The driving force of PCMI was restricted only to thermal expansion of pellet by using fresh UO 2 pellets. Fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuel rods. In seven experiments out of fourteen, test rods resulted in PCMI failure, which has been observed in the NSRR tests on high burnup PWR fuels, in terms of the transient behavior and the fracture configuration. This indicates that the driving force of PCMI failure is sufficiently explained with thermal expansion of pellet and a contribution of fission gas on it is small. A large number of incipient cracks were generated in the outer surface of the cladding even on non-failed fuel rods, and they stopped at the boundary between hydride rim, which was a hydride layer localized in the periphery of the cladding, and metallic layer. It suggests that the integrity of the metallic layer except for the hydride rim has particular importance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride layer. (author)

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

IRSN-Ancli seminar on the post-accident context

International Nuclear Information System (INIS)

Didier, Damien; Leroyer, Veronique; Gariel, Jean-Christophe; Meier, Christine; Petitfrere, Michael; Meraux-Netillard, Isabelle; Lerouxel, Roland; Gandouen, Gael; Boutin, Dominique; Charre, Jean-Pierre; Noe, Maite; Quenneville, Celine; Farandeau, Sebastien; Mouchet, Chantal; Pineau, Coralie; Rollinger, Francois; GARIEL, Jean-Christophe; Ando, Ryoko; Nishida, Shoshi; Miazaki, Makoto; Hayano, Ryugo; Lheureux, Yves; Lochard, Jacques; Boilley, David; Godet, Jean-Luc

2014-10-01

The first session addressed the context of post-accident management: main challenges of radiation protection in case of nuclear accident, management of energy situations (specific intervention plans of nuclear plants), elements of doctrine for the post-accident management of an accident. The second session addressed the preparedness of territories to post-accident management: preparation to post-accident management in the Montbeliard district, emergency and post-accidental situation (preparedness at the district scale, example of Loiret), and return on experience from the post-accident exercise in Cattenom. The third session addressed the action undertaken by the ANCCLI and IRSN for the awareness of post-accidental problematic (experiments in Saclay, Marcoule, Gravelines and Golfech, lessons learned from the pilot phase and perspectives). The last session addressed the post-accidental management of the Fukushima accident: approach of the IRSN to learn lessons from the dialogue initiative in Fukushima, round table on challenges on the long term of post-accidental management, Japanese witnesses

Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report; Unfallanalysen in Kernkraftwerken nach anlagenexternen ausloesenden Ereignissen und im Nichtleistungsbetrieb. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

2016-06-15

The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

Water tests for determining post voiding behavior in the LMFBR

International Nuclear Information System (INIS)

Hinkle, W.D.

1976-06-01

The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described were conceived with the objective of obtaining basic information and data relating to this possibility

Strategy generation in accident management support

International Nuclear Information System (INIS)

Sirola, M.

1995-01-01

An increased interest for research in the field of Accident Management can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accident in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The ideal of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information form the plant will help the strategy planning. (author). 12 refs, 2 figs

Internal dose assessment in radiation accidents

International Nuclear Information System (INIS)

Toohey, R.E.

2003-01-01

Although numerous models have been developed for occupational and medical internal dosimetry, they may not be applicable to an accident situation. Published dose coefficients relate effective dose to intake, but if acute deterministic effects are possible, effective dose is not a useful parameter. Consequently, dose rates to the organs of interest need to be computed from first principles. Standard bioassay methods may be used to assess body contents, but, again, the standard models for bioassay interpretation may not be applicable because of the circumstances of the accident and the prompt initiation of decorporation therapy. Examples of modifications to the standard methodologies include adjustment of biological half-times under therapy, such as in the Goiania accident, and the same effect, complicated by continued input from contaminated wounds, in the Hanford 241 Am accident. (author)

Calculation of spent fuel pool severe accident with MELCOR

International Nuclear Information System (INIS)

Deng Jian; Xiang Qing'an; Zhou Kefeng

2014-01-01

A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

Safety analysis of accident localization system

International Nuclear Information System (INIS)

1999-01-01

A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

Probabilistic Accident Progression Analysis with application to a LMFBR design

International Nuclear Information System (INIS)

Jamali, K.M.

1982-01-01

A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-15

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

International Nuclear Information System (INIS)

Park, Soo Young; Ahn, Kwang Il

2012-01-01

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

International Nuclear Information System (INIS)

Sowder, Andrew

2013-01-01

Department of Energy is providing substantial support for initial R and D on accident-tolerant fuel concepts with an aggressive target of a lead test assembly (LTA) in an LWR by 2022. EPRI proposes an additional stretch goal of commercialisation of a new LWR fuel by 2030. The scale of and resource demands associated with these R and D targets require a global collaborative structure to leverage resources, create an environment for innovation and co-operation, and foster necessary partnerships and arrangements among the many key players and roles spanning government, academic, and industrial sectors. EPRI is proposing a voluntary, open, and non-binding structure to quickly build momentum and to maximise early engagement and information exchange among key stakeholders. The flexibility of this organisational model offers an environment that is compatible with and encourages engagement, innovation, and development of the more formal arrangements and partnerships that will be needed to commercialise current R and D concepts. The opportunity for transformation of LWR fuel performance under normal and accident conditions is now. Accordingly, the time for action is now. Commercialisation of accident-tolerant fuel in the near future can only be realised with collaboration among governments, industry and academia on a scale commensurate with the challenges at hand

Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

Energy Technology Data Exchange (ETDEWEB)

Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

International Nuclear Information System (INIS)

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

1987-01-01

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

The Chernobyl accident - five years later

International Nuclear Information System (INIS)

Mueck, K.

1991-06-01

At the fifth anniversary of the Chernobyl accident the initial situation at that time, the control of the consequences to Austria in the present light, as well as the knowledge gained from the accident and its consequences are described. A final estimate and appraisal of the total population dose by the accident alloted according to the individual exposure pathways and the dose reductions due to countermeasures by the authorities are given. The dose reduction in the following years is described. Five years later the external exposure was reduced to about 6 % of the values of the first year, the ingestion dose to about 5 % of the first-year-values. Finally, the current radiation situation is described and the dose contribution by foodstuff with elevated activity concentration is estimated. Also the consequences from the experience and knowledge obtained by the accident are described. (author)

Personal nuclear accident dosimetry at Sandia National Laboratories

International Nuclear Information System (INIS)

Ward, D.C.; Mohagheghi, A.H.; Burrows, R.

1996-09-01

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%

A critical assessment of energy accident studies

International Nuclear Information System (INIS)

Felder, Frank A.

2009-01-01

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases.

A critical assessment of energy accident studies

Energy Technology Data Exchange (ETDEWEB)

Felder, Frank A. [Edward J. Bloustein School of Planning and Public Policy, Rutgers, The State University of New Jersey, 33 Livingston Avenue, New Brunswick, NJ 08901 (United States)

2009-12-15

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases. (author)

CFD Analyses of Air-Ingress Accident for VHTRs

Science.gov (United States)

Ham, Tae Kyu

The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air

The Chernobyl accident and its consequences.

Science.gov (United States)

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

2011-05-01

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

Ductile crack initiation in the Charpy V-notch test

International Nuclear Information System (INIS)

Server, W.L.; Norris, D.M. Jr.; Prado, M.E.

1978-01-01

Initiation and growth of a crack in the Charpy V-notch test was investigated by performing both static and impact controlled deflection tests. Test specimens were deformed to various deflections, heat-tinted to mark crack extension and broken apart at low temperature to allow extension measurements. Measurement of the crack extension provided an estimate of crack initiation as defined by different criteria. Crack initiation starts well before maximum load, and is dependent on the definition of ''initiation''. Using a definition of first micro-initiation away from the ductile blunting, computer model predictions agreed favorably with the experimental results

Chernobyl reactor accident

International Nuclear Information System (INIS)

Malinauskas, A.P.; Buchanan, J.R.; Lorenz, R.A.; Yamashita, T.

1986-01-01

On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

Mitigation of Severe Accident Consequences Using Inherent Safety Principles

International Nuclear Information System (INIS)

Wigeland, R.A.; Cahalan, J.E.

2009-01-01

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

Lessons learned from accidents in radiotherapy. An IAEA Safety Report

International Nuclear Information System (INIS)

Ortiz, P.

1998-01-01

Radiotherapy is a very special application from the view point of protection because humans are deliberately exposed to high doses of radiation, and no physical barrier can be placed between the source and the patient. It deserves, therefore, special considerations from the point of view of potential exposure. An IAEA's Safety Report (in preparation) reviews a large collection of accident information, their initiating events and contributing factors, followed by a set of lessons learned and measures for prevention. The most important causes were: deficiencies in education and training, lack of procedures and protocols for essential tasks (such as commissioning, calibration, commissioning and treatment delivery), deficient communication and information transfer, absence of defence in depth and deficiencies in design, manufacture, testing and maintenance of equipment. Often a combination of more than one of these causes was present in an accident, thus pointing to a problem of management. Arrangements for a comprehensive quality assurance and accident prevention should be required by regulations and compliance be monitored by a Regulatory Authority. (author)

The effect of a yellow bicycle jacket on cyclist accidents

DEFF Research Database (Denmark)

Lahrmann, Harry; Madsen, Tanja Kidholm Osmann; Olesen, Anne Vingaard

2018-01-01

Highlights •A randomised controlled trial with 6793 cyclists shows a reduced accident risk due to a yellow bicycle jacket. •The test group had 47% fewer multiparty accidents with personal injury. •The test group had 55% fewer multiparty accidents against motorised vehicles....

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

International Nuclear Information System (INIS)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-01

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

An assessment of core wide coherency effects in the multichannel modeling of the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, M.; Gubernatis, P.; Suteau, C.; Le Tellier, R.; Lecerf, J.

2014-01-01

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly to- channel arrangement. The multiple-assembly-per-channel approach introduces core wide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces core wide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly to- channel arrangement. (authors)

Identification and evaluation of PWR in-vessel severe accident management strategies

International Nuclear Information System (INIS)

Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

1992-03-01

This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

Criticality accident alarm system

International Nuclear Information System (INIS)

Malenfant, R.E.

1991-01-01

The American National Standard ANSI/ANS-8.3-1986, Criticality Accident Alarm System provides guidance for the establishment and maintenance of an alarm system to initiate personnel evacuation in the event of inadvertent criticality. In addition to identifying the physical features of the components of the system, the characteristics of accidents of concern are carefully delineated. Unfortunately, this ANSI Standard has led to considerable confusion in interpretation, and there is evidence that the ''minimum accident of concern'' may not be appropriate. Furthermore, although intended as a guide, the provisions of the standard are being rigorously applied, sometimes with interpretations that are not consistent. Although the standard is clear in the use of absorbed dose in free air of 20 rad, at least one installation has interpreted the requirement to apply to dose in soft tissue. The standard is also clear in specifying the response to both neutrons and gamma rays. An assembly of uranyl fluoride enriched to 5% 235 U was operated to simulate a potential accident. The dose, delivered in a free run excursion 2 m from the surface of the vessel, was greater than 500 rad, without ever exceeding a rate of 20 rad/min, which is the set point for activating an alarm that meets the standard. The presence of an alarm system would not have prevented any of the five major accidents in chemical operations nor is it absolutely certain that the alarms were solely responsible for reducing personnel exposures following the accident. Nevertheless, criticality alarm systems are now the subject of great effort and expense. 13 refs

Test reactor risk assessment methodology

International Nuclear Information System (INIS)

Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

1976-04-01

A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

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

International Nuclear Information System (INIS)

Majumdar, S.

1996-09-01

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

Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

2015-10-15

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

International Nuclear Information System (INIS)

Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

2015-01-01

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

The development of severe accident analysis technology

Energy Technology Data Exchange (ETDEWEB)

Kim, Heuy Dong; Cho, Sung Won; Kim, Sang Baek; Park, Jong Hwa; Lee, Kyu Jung; Park, Lae Joon; Hu, Hoh; Hong, Sung Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-07-01

The objective of the development of severe accident analysis technology is to understand the severe accident phenomena such as core melt progression and to provide a reliable analytical tool to assess severe accidents in a nuclear power plant. Furthermore, establishment of the accident management strategies for the prevention/mitigation of severe accidents is also the purpose of this research. The study may be categorized into three areas. For the first area, two specific issues were reviewed to identify the further research direction, that is the natural circulation in the reactor coolant system and the fuel-coolant interaction as an in-vessel and an ex-vessel phenomenological study. For the second area, the MELCOR and the CONTAIN codes have been upgraded, and a validation calculation of the MELCOR has been performed for the PHEBUS-B9+ experiment. Finally, the experimental program has been established for the in-vessel and the ex-vessel severe accident phenomena with the in-pile test loop in KMRR and the integral containment test facilities, respectively. (Author).

The nuclear accidents: Causes and consequences

International Nuclear Information System (INIS)

Rochd, M.

1988-01-01

The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-01-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-10-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

International Nuclear Information System (INIS)

Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

2017-01-01

Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.

Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

Energy Technology Data Exchange (ETDEWEB)

Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

2016-07-15

Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

Accident management for severe accidents

International Nuclear Information System (INIS)

Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

1988-01-01

The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

International Nuclear Information System (INIS)

Fehrenbach, P.J.; Wood, J.C.

1987-06-01

Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

Research investigation report on Fukushima Daiichi nuclear accident

International Nuclear Information System (INIS)

2012-03-01

This report was issued in February 2012 by Rebuild Japan Initiative Foundation's Independent Investigation Commission on the Fukushima Daiichi Nuclear Accident, which consisted of six members from the private sector in independent positions and with no direct interest in the business of promoting nuclear power. Commission aimed to determine the truth behind the accident by clarifying the various problems and reveal systematic problems behind these issues so as to create a new starting point by identifying clear lessons learned. Report composed of four chapters; (1) progression of Fukushima accident and resulting damage (accident management after Fukushima accident, and effects and countermeasure of radioactive materials discharged into the environment), (2) response against Fukushima accident (emergency response of cabinet office against nuclear disaster, risk communication and on-site response against nuclear disaster), (3) analysis of historical and structural factors (technical philosophy of nuclear safety, problems of nuclear safety regulation of Fukushima accident, safety regulatory governance and social background of 'Safety Myth'), (4) Global Context (implication in nuclear security, Japan in nuclear safety regime, U.S.-Japan relations for response against Fukushima accident, lessons learned from Fukushima accident - aiming at creation of resilience). Report could identify causes of Fukushima accident and factors related to resulting damages, show the realities behind failure to prevent the spread of damage, and analyze the overall structural and historical background behind the accidents. (T. Tanaka)

Examining accident reports involving autonomous vehicles in California.

Directory of Open Access Journals (Sweden)

Francesca M Favarò

Full Text Available Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017. The data provides important information on autonomous vehicles accidents' dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama.

Examining accident reports involving autonomous vehicles in California.

Science.gov (United States)

Favarò, Francesca M; Nader, Nazanin; Eurich, Sky O; Tripp, Michelle; Varadaraju, Naresh

2017-01-01

Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017). The data provides important information on autonomous vehicles accidents' dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama.

Analysis of severe accidents on fast reactor test loop

International Nuclear Information System (INIS)

Cenerini, R.; Verzelletti, G.; Curioni, S.

1975-01-01

The Pec reactor is a sodium cooled fast reactor which is being designed for the primary purpose of accomodating closed sodium cooled test loops for the developmental and proof testing of fast reactor fuel assemblies. The test loops are located in the central test region of reactor. The basic function for which the loop is designed is burn-up to failure testing of fuel under advanced performance conditions. It is therefore necessary to design the loop for failure conditions. Basically two types of accidents can occur within the loops: rupture of gas plenum in the fuel pins and coolant starvation. Explosive tests on Pec loop, whose first set is described in this report, are devoted to investigate the effects of an accidental energy release on loop containment. The loop model reproduces in the test section the prototype dimensions in radial scale 1:1. Using a wire explosive charge of 300mm, the height of test section is sufficient for determining the containment capability of the loop that has a nearly constant deformation in a length of. 3-4 time the diameter. The inertial effects of the coolant column are reproduced by two tubes at the extremities of test section, closed with top plugs. Some tests has been performed by wrapping around the test section four layers of steel wire in order to evaluate the influence on the containment of tungsten wire that is foreseen in prototype loop. The influence of the coolant around the loop was evaluated by inserting the model in water. Dummy sub-assemblies was used and explosive substitutes the central rods. Piezoelectric pressure transducers were mounted on the three plugs and radial deformation was measured directly at different height. From experiments performed it resulted the importance of harmonic wires and inertial reaction of external water on loop containment; maximum containable energy is about 50 Cal with E.1 explosive

Test set for initial value problem solvers

NARCIS (Netherlands)

W.M. Lioen (Walter); J.J.B. de Swart (Jacques)

1998-01-01

textabstractThe CWI test set for IVP solvers presents a collection of Initial Value Problems to test solvers for implicit differential equations. This test set can both decrease the effort for the code developer to test his software in a reliable way, and cross the bridge between the application

The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

2012-10-15

Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

Laterality, spatial abilities, and accident proneness.

Science.gov (United States)

Voyer, Susan D; Voyer, Daniel

2015-01-01

Although handedness as a measure of cerebral specialization has been linked to accident proneness, more direct measures of laterality are rarely considered. The present study aimed to fill that gap in the existing research. In addition, individual difference factors in accident proneness were further examined with the inclusion of mental rotation and navigation abilities measures. One hundred and forty participants were asked to complete the Mental Rotations Test, the Santa Barbara Sense of Direction scale, the Greyscales task, the Fused Dichotic Word Test, the Waterloo Handedness Questionnaire, and a grip strength task before answering questions related to number of accidents in five areas. Results indicated that handedness scores, absolute visual laterality score, absolute response time on the auditory laterality index, and navigation ability were significant predictors of the total number of accidents. Results are discussed with respect to cerebral hemispheric specialization and risk-taking attitudes and behavior.

Characteristics of initial deposition and behavior of radiocesium in forest ecosystems of different locations and species affected by the Fukushima Daiichi Nuclear Power Plant accident

International Nuclear Information System (INIS)

Komatsu, Masabumi; Kaneko, Shinji; Ohashi, Shinta; Kuroda, Katsushi; Sano, Tetsuya; Ikeda, Shigeto; Saito, Satoshi; Kiyono, Yoshiyuki; Tonosaki, Mario; Miura, Satoru; Akama, Akio; Kajimoto, Takuya; Takahashi, Masamichi

2016-01-01

After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, information about stand-level spatial patterns of radiocesium initially deposited in the surrounding forests was essential for predicting the future dynamics of radiocesium and suggesting a management plan for contaminated forests. In the first summer (approximately 6 months after the accident), we separately estimated the amounts of radiocesium ("1"3"4Cs and "1"3"7Cs; Bq m"−"2) in the major components (trees, organic layers, and soils) in forests of three sites with different contamination levels. For a Japanese cedar (Cryptomeria japonica) forest studied at each of the three sites, the radiocesium concentration greatly differed among the components, with the needle and organic layer having the highest concentrations. For these cedar forests, the proportion of the "1"3"7Cs stock in the aboveground tree biomass varied from 22% to 44% of the total "1"3"7Cs stock; it was 44% in highly contaminated sites (7.0 × 10"5 Bq m"−"2) but reduced to 22% in less contaminated sites (1.1 × 10"4 Bq m"−"2). In the intermediate contaminated site (5.0–5.8 × 10"4 Bq m"−"2), 34% of radiocesium was observed in the aboveground tree biomass of the Japanese cedar stand. However, this proportion was considerably smaller (18–19%) in the nearby mixed forests of the Japanese red pine (Pinus densiflora) and deciduous broad-leaved trees. Non-negligible amounts of "1"3"4Cs and "1"3"7Cs were detected in both the sapwood and heartwood of all the studied tree species. This finding suggested that the uptake or translocation of radiocesium had already started within 6 months after the accident. The belowground compartments were mostly present in the organic layer and the uppermost (0–5 cm deep) mineral soil layer at all the study sites. We discussed the initial transfer process of radiocesium deposited in the forest and inferred that the type of initial deposition (i.e., dry versus wet radiocesium deposition

Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

International Nuclear Information System (INIS)

Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo

2016-01-01

An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted

Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

Energy Technology Data Exchange (ETDEWEB)

Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted.

Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2013-01-01

Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

Development of the simulation system IMPACT for analysis of nuclear power plant severe accidents

International Nuclear Information System (INIS)

Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

1997-01-01

The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system IMPACT for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT's distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data

Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

International Nuclear Information System (INIS)

Lee, Yongjin; Jae, Moosung

2014-01-01

The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

What one should know about radiation. Comparison of radiation burden from the Chernobyl accident and the atomic weapons test

Energy Technology Data Exchange (ETDEWEB)

Burtscher, A

1986-01-01

The natural radiation burden, that due to the Chernobyl accident and the atmospheric nuclear weapons tests in Austria are compared. The overall Chernobyl burden is estimated at 50-70% of the annual natural burden and thus less than the burden from atmospheric nuclear weapons tests. (G.Q.).

In-vessel natural circulation during a hypothetical loss-of-heat-sink accident in the Fast Flux Test Facility

International Nuclear Information System (INIS)

Perkins, K.R.; Bari, R.A.; Pratt, W.T.

1979-05-01

The capability to remove decay heat from the FFTF core via in-vessel natural circulation has been analyzed for the preboiling phase using a lumped parameter model. The results indicate that boiling will occur in the average fuel assembly for a wide spectrum of initial conditions which appear to be representative of the hypothetical loss-of-heat-sink accident. Two-phase pressure drop calculations indicate that, once the saturation temperature is reached, coolability can only be assured for decay heat levels which are less than 0.5% of the operating power. A review of the limited sodium boiling data indicates that boiling-induced natural circulation may support up to 4% of the operating power, but geometric atypicalities and a large degree of inlet subcooling for the existing data limit the applicability to the loss-of-heat-sink accident in FFTF

Operator modeling of a loss-of-pumping accident using MicroSAINT

International Nuclear Information System (INIS)

Olsen, L.M.

1992-01-01

The Savannah River Laboratory (SRL) human factors group has been developing methods for analyzing nuclear reactor operator actions during hypothetical design-basis accident scenarios. The SRL reactors operate at a lower temperature and pressure than power reactors resulting in accident sequences that differ from those of power reactors. Current methodology development is focused on modeling control room operator response times dictated by system event times specified in the Savannah River Site Reactor Safety Analysis Report (SAR). The modeling methods must be flexible enough to incorporate changes to hardware, procedures, or postulated system event times and permit timely evaluation. The initial model developed was for the loss-of-pumping accident (LOPA) because a significant number of operator actions are required to respond to this postulated event. Human factors engineers had been researching and testing a network modeling simulation language called MicroSAINT to simulate operators' personal and interpersonal actions relative to operating system events. The LOPA operator modeling project demonstrated the versatility and flexibility of MicroSAINT for modeling control room crew interactions

Nuclear Facility Accident (NFAC) Unit Test Report For HPAC Version 6.3

Energy Technology Data Exchange (ETDEWEB)

Lee, Ronald W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Computational Sciences and Engineering Division; Morris, Robert W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Computational Sciences and Engineering Division; Sulfredge, Charles David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Computational Sciences and Engineering Division

2015-12-01

This is a unit test report for the Nuclear Facility Accident (NFAC) model for the Hazard Prediction and Assessment Capability (HPAC) version 6.3. NFAC’s responsibility as an HPAC component is three-fold. First, it must present an interactive graphical user interface (GUI) by which users can view and edit the definition of an NFAC incident. Second, for each incident defined, NFAC must interact with RTH to create activity table inputs and associate them with pseudo materials to be transported via SCIPUFF. Third, NFAC must create SCIPUFF releases with the associated pseudo materials for transport and dispersion. The goal of NFAC unit testing is to verify that the inputs it produces are correct for the source term or model definition as specified by the user via the GUI.

Touch-sensitive colour graphics enhance monitoring of loss-of-coolant accident tests

International Nuclear Information System (INIS)

Snedden, M.D.; Mead, G.L.

1982-01-01

A stand-alone computer-based system with an intelligent colour termimal is described for monitoring parameters during loss-of-coolant accident tests. Colour graphic displays and touch-sensitive control have been combined for effective operator interaction. Data collected by the host MODCOMP II minicomputer are dynamically updated on colour pictures generated by the terminal. Experimenters select system functions by touching simulated switches on a transparent touch-sensitive overlay, mounted directly over the face of the colour screen, eliminating the need for a keyboard. Switch labels and colours are changed on the screen by the terminal software as different functions are selected. Interaction is self-prompting and can be learned quickly. System operation for a complete set of 20 tests has demonstrated the convenience of interactive touchsensitive colour graphics

Nuclear Facility Accident (NFAC) Unit Test Report For HPAC Version 6.3

International Nuclear Information System (INIS)

Lee, Ronald W.; Morris, Robert W.; Sulfredge, Charles David

2015-01-01

This is a unit test report for the Nuclear Facility Accident (NFAC) model for the Hazard Prediction and Assessment Capability (HPAC) version 6.3. NFAC's responsibility as an HPAC component is three-fold. First, it must present an interactive graphical user interface (GUI) by which users can view and edit the definition of an NFAC incident. Second, for each incident defined, NFAC must interact with RTH to create activity table inputs and associate them with pseudo materials to be transported via SCIPUFF. Third, NFAC must create SCIPUFF releases with the associated pseudo materials for transport and dispersion. The goal of NFAC unit testing is to verify that the inputs it produces are correct for the source term or model definition as specified by the user via the GUI.

Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

1977-01-01

With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

The philosophy of severe accident management in the US

International Nuclear Information System (INIS)

Baratta, A.J.

1990-01-01

The US NRC has put forth the initial steps in what is viewed as the resolution of the severe accident issue. Underlying this process is a fundamental philosophy that if followed will likely lead to an order of magnitude reduction in the risk of severe accidents. Thus far, this philosophy has proven cost effective through improved performance. This paper briefly examines this philosophy and the next step in closure of the severe accident issue, the IPE. An example of the authors experience with determinist. (author)

Accident sequences and causes analysis in a hydrogen production process

Energy Technology Data Exchange (ETDEWEB)

Jae, Moo Sung; Hwang, Seok Won; Kang, Kyong Min; Ryu, Jung Hyun; Kim, Min Soo; Cho, Nam Chul; Jeon, Ho Jun; Jung, Gun Hyo; Han, Kyu Min; Lee, Seng Woo [Hanyang Univ., Seoul (Korea, Republic of)

2006-03-15

Since hydrogen production facility using IS process requires high temperature of nuclear power plant, safety assessment should be performed to guarantee the safety of facility. First of all, accident cases of hydrogen production and utilization has been surveyed. Based on the results, risk factors which can be derived from hydrogen production facility were identified. Besides the correlation between risk factors are schematized using influence diagram. Also initiating events of hydrogen production facility were identified and accident scenario development and quantification were performed. PSA methodology was used for identification of initiating event and master logic diagram was used for selection method of initiating event. Event tree analysis was used for quantification of accident scenario. The sum of all the leakage frequencies is 1.22x10{sup -4} which is similar value (1.0x10{sup -4}) for core damage frequency that International Nuclear Safety Advisory Group of IAEA suggested as a criteria.

Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

International Nuclear Information System (INIS)

Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

1993-01-01

Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

Energy Technology Data Exchange (ETDEWEB)

Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

2005-08-15

This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

PSA modeling of long-term accident sequences

International Nuclear Information System (INIS)

Georgescu, Gabriel; Corenwinder, Francois; Lanore, Jeanne-Marie

2014-01-01

In the context of the extension of PSA scope to include external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the accident sequences induced by initiators which affect a whole site containing several nuclear units (reactors, fuel pools,...). These methodological improvements represent an essential prerequisite for the development of external hazards PSA. However, it has to be noted that in French PSA, even before Fukushima, long term accident sequences were taken into account: many insight were therefore used, as complementary information, to enhance the safety level of the plants. IRSN proposed an external events PSA development program. One of the first steps of the program is the development of methods to model in the PSA the long term accident sequences, based on the experience gained. At short term IRSN intends to enhance the modeling of the 'long term' accident sequences induced by the loss of the heat sink or/and the loss of external power supply. The experience gained by IRSN and EDF from the development of several probabilistic studies treating long term accident sequences shows that the simple extension of the mission time of the mitigation systems from 24 hours to longer times is not sufficient to realistically quantify the risk and to obtain a correct ranking of the risk contributions and that treatment of recoveries is also necessary. IRSN intends to develop a generic study which can be used as a general methodology for the assessment of the long term accident sequences, mainly generated by external hazards and their combinations. This first attempt to develop this generic study allowed identifying some aspects, which may be hazard (or combinations of hazards) or related to initial boundary conditions, which should be taken into account for further developments. (authors)

Fukushima accident - reasons and impacts

International Nuclear Information System (INIS)

Slugen, V.

2011-01-01

The Fukushima accident influenced dramatically the current view on safety of nuclear facilities. Consideration about possible impacts of natural catastrophe in design of nuclear facilities seems to be much more important than before. European commission is focused on the stress-tests at nuclear power plants. His paper will go more in details having in mind reasons and impacts of Fukushima accident (Author)

Modelling Reactivity-Initiated-Accident Experiments With Falcon And SCANAIR: A Comparison Exercise

International Nuclear Information System (INIS)

Romano, A.; Wallin, H.; Zimmermann, M.A.

2005-01-01

A critical assessment is made of the state-of-the-art fuel performance code FALCON in the context of selected Reactivity Initiated Accident (RIA) experiments from the CABRI REP Na series, and contrasts its predictions against those of the extensively benchmarked SCANAIR (Version 3.2) code. The thermal fields in the fuel and cladding, the clad mechanical deformation, and the Fission Gas Release (FGR) are adopted as 'Figures of Merit' by which to judge code performance. Particular attention is paid to the importance of fission-gas-induced clad deformation (which is modelled in SCANAIR, but not in FALCON), relative to that driven by the fuel thermal expansion (which is modelled by both codes). The thermal fields calculated by the codes are in good agreement with each other, especially during the initial stages of the transients --- the adiabatic phase. Larger discrepancies are observed at later times, and are due to the different models applied to calculate the gap conductance. FALCON predicts clad permanent deformations at the end of the transients with a maximum deviation from the experimental measurements of about 20%. Generally, the code always tends to underpredict the measurements. SCANAIR performs similarly, but grossly overpredicts the permanent clad strain for the case involving a very energetic pulse. The fission-gas-driven clad deformation is only relevant for very fast pulse energy injection cases, which are not prototypical of the RIA transients expected in PWRs. The FGR models in FALCON do not capture the mechanism of 'burst-release' in the RIA transients, having been developed for steady-state irradiation conditions. This also explains why they performed poorly when applied to the fast-transient cases analyzed here. In contrast, the FGR results from SCANAIR are in satisfactory agreement with the experimental results. (author)

Full-length fuel rod behavior under severe accident conditions

International Nuclear Information System (INIS)

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

1992-12-01

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

Environmental monitoring activities in JAERI at JCO accident

International Nuclear Information System (INIS)

Yamaguchi, Takenori

2001-01-01

The Japan Atomic Energy Research Institute (JAERI) was involved in a various environmental monitoring activities, such as environmental radiation monitoring by monitoring post and monitoring car, air dust, soil, green vegetables and water sampling and measurements, neutron and gamma radiation dose rate survey around the JCO site for emergency response actions to the JCO accident on September 30, 1999. These activities were performed from Sep. 30 to Oct. 2, and were the initial and first stage activities in the emergency environmental monitoring activities. JAERI has been assigned to the public organization to support the government by Disaster Prevention Fundamental Law. These activities were performed to ensure the public safety to avoid the effluent of the accident. Through the environmental monitoring activities, I recognized that the importance of the accident information to make the best use for the initial environmental monitoring, and the monitoring information exchange is important to perform the effective monitoring activities for taking the early countermeasures such as evacuation to the public. (author)

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

International Nuclear Information System (INIS)

Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-03-01

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

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Joy L. Rempe; Darrell L. Knudson

2014-05-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Joy L. Rempe; Darrell L. Knudson

2013-03-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

SCC Initiation Testing of Alloy 600 in High Temperature Water

Science.gov (United States)

Etien, Robert A.; Richey, Edward; Morton, David S.; Eager, Julie

Stress corrosion cracking (SCC) initiation tests have been conducted on Alloy 600 at temperatures from 304 to 367°C. Tests were conducted with in-situ monitored smooth tensile specimens under a constant load in hydrogenated environments. A reversing direct current electric potential drop (EPD) system was used for all of the tests to detect SCC initiation. Tests were conducted to examine the effects of stress (and strain), coolant hydrogen, and temperature on SCC initiation time. The thermal activation energy of SCC initiation was measured as 103 ± 18 kJ/mol in hydrogenated water, which is similar to the thermal activation energy for SCC growth. Results suggest that the fundamental mechanical parameter which controls SCC initiation is plastic strain not stress. SCC initiation was shown to have a different sensitivity than SCC growth to dissolved hydrogen level. Specifically, SCC initiation time appears to be relatively insensitive to hydrogen level in the nickel stability region.

Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

International Nuclear Information System (INIS)

Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

2000-03-01

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

Prediction of failure of highly irradiated Zircaloy clad tubes under reactivity initiated accidents

International Nuclear Information System (INIS)

Jernkvist, L.O.

2003-01-01

This paper deals with failure of irradiated Zircaloy tubes under the heat-up stage of a reactivity initiated accident (RIA). More precisely, by use of a model for plastic strain localization and necking failure, we theoretically analyse the effects of local surface defects on clad ductility and survivability under RIA. The results show that even very shallow surface defects, e.g. arising from a non-uniform or partially spilled oxide layer, have a strong limiting effect on clad ductility. Moreover, in presence of surface defects, the ability of the clad tube to expand radially without necking failure is found to be extremely sensitive to the stress biaxiality ratio σ zz /σ θθ , which is here assumed to be in the range from 0 to 1. The results of our analysis are compared with clad ductility data available in literature, and their consequences for clad failure prediction under RIA are discussed. In particular, the results raise serious concerns regarding the applicability of failure criteria, which are based on clad strain energy density. These criteria do not capture the observed sensitivity to stress biaxiality on clad failure propensity. (author)

Pedestrian injury causation study (pedestrian accident typing)

Science.gov (United States)

1982-08-01

A new computerized pedestrian accident typing procedure was tested on 1,997 cases from the Pedestrian Injury Causation Study (PICS). Two coding procedures were used to determine the effects of quantity and quality of information on accident typing ac...

Severe fuel-damage scoping test performance

International Nuclear Information System (INIS)

Gruen, G.E.; Buescher, B.J.

1983-01-01

As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

The Chernobyl reactor accident source term: Development of a consensus view

International Nuclear Information System (INIS)

Guntay, S.; Powers, D.A.; Devell, L.

1997-01-01

In August 1986, scientists from the former Soviet Union provided the nuclear safety community with an impressively detailed account of what was then known about the Chernobyl accident. This included assessments of the magnitudes, rates, and compositions of radionuclide releases during the ten days following initiation of the accident. A summary report based on the Soviet report, the oral presentations, and the discussions with scientists from various countries was issued by the International Atomic Energy Agency shortly thereafter. Ten years have elapsed since the reactor accident at Chernobyl. A great deal more data is now available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for about ten days. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved our understanding of the Chernobyl source term. Because of the special features of the reactor design and the pecularities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability of the safety analysis of other types of reactors

Modernized CDTN's air-water experimental test circuit: initial results

International Nuclear Information System (INIS)

Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos

2017-01-01

The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

Modernized CDTN's air-water experimental test circuit: initial results

Energy Technology Data Exchange (ETDEWEB)

Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos, E-mail: marcioaraujopessoa@gmail.com, E-mail: mauricio.sobrinho223@gmail.com, E-mail: e.a.salomao@gmail.com, E-mail: arthur1303@gmail.com, E-mail: moysesnavarro@yahoo.com.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

Effects of the Chernobyl accident on public perceptions of nuclear plant accident risks

International Nuclear Information System (INIS)

Lindell, M.K.; Perry, R.W.

1990-01-01

Assessments of public perceptions of the characteristics of a nuclear power plant accident and affective responses to its likelihood were conducted 5 months before and 1 month after the Chernobyl accident. Analyses of data from 69 residents of southwestern Washington showed significant test-retest correlations for only 10 of 18 variables--accident likelihood, three measures of impact characteristics, three measures of affective reactions, and hazard knowledge by governmental sources. Of these variables, only two had significant changes in mean ratings; frequency of thought and frequency of discussion about a nearby nuclear power plant both increased. While there were significant changes only for two personal consequences (expectations of cancer and genetic effects), both of these decreased. The results of this study indicate that more attention should be given to assessing the stability of risk perceptions over time. Moreover, the data demonstrate that experience with a major accident can actually decrease rather than increase perceptions of threat

Reconstruction of dose loads on population in the initial period of the Chernobyl accident and estimation of thyroid cancer risk in Belarus

International Nuclear Information System (INIS)

Krivoruchko, K.; Naumov, A.

1997-01-01

The Chernobyl accident caused significant long-term consequences to the environment, public health, and economic status of Belarus. The contamination from short-lived radionuclides, in particular iodine 131, was so high that the subsequent exposure of millions of people has been termed 'iodine shock'. During the first days of the accident, the majority of the dose of radiation received by the residents of Belarus was to the thyroid gland. This will affect the health of the population for a long time to come. The resulting epidemic of childhood thyroid cancer is the first indisputable health after-effect of the Chernobyl accident. Thyroid cancer morbidity among children increased more than 10 fold in the post-Chernobyl period. Maps of cesium 137, which has a half life of 37 years, have been published, but it is evident, that the distribution of thyroid cancer morbidity differs from the known distribution of cesium 137 in soil. Territorial distribution of thyroid cancer morbidity is often compared to distribution of cesium 137 in the soil. This practice is inaccurate but often utilized since no maps of iodine 131 contamination exist, due to its short half life of 8.04 days. Reconstruction of the spatial distribution of short-lived isotopes in the first days after the accident, could clarify the impact of radiation on human health and allow for a spatial and temporal prognosis of the development of the cancer epidemic, particularly, thyroid cancer. Due to the unfortunate fact that the measuring equipment was inadequate to properly monitor the scale of radiation exposure during the early period of the accident, detailed direct information on the deposition of the short-lived radionuclides and the doses to the population has been irretrievably lost. Now the only way to reconstruct the dynamics of the radioecological situation of the initial period of the Chernobyl accident is to make a retrospective assessment of radiation exposures related to the short

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

Postulated accidents

International Nuclear Information System (INIS)

Ullrich, W.

1980-01-01

This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

Severe accident management guidelines tool

International Nuclear Information System (INIS)

Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

2014-01-01

Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

HIV surveillance in needlestick accidents with health workers

Directory of Open Access Journals (Sweden)

Janete Lane Amadei

2010-12-01

Full Text Available Objective: To characterize the occurrence of needlestick accidents with health professionals submitted to rapid HIV tests. Methods: A descriptive, epidemiological study, carried out by notification of the occurrence of needlestick accidents in the Epidemiology Sector of the State Health Secretariat, in 2008. The following variables were assessed: gender, age, exposed biological material, type of exposure, source patient, and injured patient, progression of the case, accident situation, and use of personal protective equipment (PPE, 180 days serology and occupational area. Results: There have been reports of 143 accidents, prevailing in nursing, female, 20 to 30 years, involving the blood and biological material by percutaneous puncture. We found no standardization in the use of PPE. The HIV test revealed no positive cases. Conclusion: This study helped to characterize the occurrence of accidents reported in health care professionals and evaluate the protocol of care given. It also revealed non-contamination by HIV.

Severe accident risks: An assessment for five US nuclear power plants: Appendices A, B, and C

International Nuclear Information System (INIS)

1990-12-01

This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United States. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two or the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide release and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. Volume 2 of this report contains three appendices, providing greater detail on the methods used, an example risk calculation, and more detailed discussion of particular technical issues found important in the risk studies

Investigation of an accident in a resins manufacturing site: The role of accelerator on polymerisation of methyl methacrylate

Energy Technology Data Exchange (ETDEWEB)

Casson, Valeria, E-mail: valeria.casson.moreno@gmail.com [Alma Mater Studiorum—Università di Bologna, Dipartimento di Ingegneria Chimica, Mineraria e delle Tecnologie Ambientali, Bologna (Italy); Dipartimento di Ingegneria Industriale, Via Marzolo 9, 35131 Padova (Italy); Snee, Tim, E-mail: Tim.Snee@hsl.gsi.gov.uk [Health and Safety Laboratory, Harpur Hill, Buxton, Derbyshire SK 179 JN (United Kingdom); Maschio, Giuseppe, E-mail: giuseppe.maschio@unipd.it [Dipartimento di Ingegneria Industriale, Via Marzolo 9, 35131 Padova (Italy)

2014-04-01

Highlights: • The accelerator produces an increase in the initial rate of polymerisation. • The accelerator increases the extent of polymerisation in certain conditions. • The accelerator decreases the induction time due to the presence of inhibitor. • Runaway reaction is more likely to occur in presence of the accelerator. • The experimental data support the hypothesis about the accident. - Abstract: This paper analyzes the effect of an accelerator on the polymerisation of methyl methacrylate (MMA). This study is based on the results of an investigation of an accident in a manufacturing site for resins located in the United Kingdom. As sequence of event to cause the accident the following was assumed: during an unattended batch process a runaway undesired polymerisation of methyl methacrylate occurred, generating rapid vaporisation of monomer, which in contact with an ignition source, led to an explosion followed by a fire. Since no initiator for the polymerisation reaction had been jet added to the blend, it was supposed that the accelerator contributed to the onset of the undesired polymerisation. The accelerator involved in the accident t has therefore been tested by differential scanning calorimetry and adiabatic calorimetry. The experimental data allowed the authors to prove the hypothesis made and to define safety ranges for the polymerisation reaction.

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.E.; Sype, T.T.; Camp, A.L.

1990-01-01

The risk from five nuclear power plants was examined during the NUREG-1150 program. When the analyses of the plants were complete, an effort was undertaken to examine the implications of NUREG-1150 for accident management initiatives. The framework provided by the NUREG-1150 analysis presented a means within which current accident management strategies could be evaluated and future accident management strategies could be developed and assessed. Five separate but interrelated phases of risk management were considered: (1) prevention of accident initiators, (2) prevention of core damage, (3) implementation of an effective emergency response, (4) prevention of vessel breach and mitigation of radionuclide releases from the reactor coolant system, and (5) retention of fission products in the containment and other surrounding buildings. A risk-based methodology was developed to identify and evaluate risk management options for each of these five phases. The methodology was demonstrated through quantitative examples for the first two phases of risk management listed above. In addition, the reduction in risk for several currently implemented risk management strategies at operating plants was quantified

Accident management

International Nuclear Information System (INIS)

Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

1989-01-01

The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

Consequences of the Chernobyl accident in Lithuania

International Nuclear Information System (INIS)

Mastauskas, A.; Nedvecktaite, T.; Filistovic, V.

1997-01-01

After the Chernobyl accident of 26 April, 1986, population dose assessment favours the view that the radiation risk of population effected by the early fallout would be different from that in regions contaminated later. Taking into account the short half-time of the most important radioactive iodine isotopes, thyroid disorders would be expected mainly to follow the early fallout distribution. At the time of accident at Unite 4 of the Chernobyl NPP, surface winds were from the Southeast. The initial explosions and heat carried volatile radioactive materials to the 1,5 km height, from where they were transported over the Western part of Belarus, Southern and Western part of Lithuania toward Scandinavian countries. Thus the volatile radioiodine and some other radionuclides were detected in Lithuania on the very first days after the accident. The main task of the work - to conduct short Half-time radioiodine and long half-time radiocesium dose assessment of Lithuanian inhabitants a result of the early Chernobyl accident fallout

Synergy effect in accident simulation

International Nuclear Information System (INIS)

Alba, C.; Carlin, F.; Chenion, J.; Gaussens, G.; Le Meur, M.; Petitjean, M.

1984-05-01

Accidental breaking of PWR coolant canalization would entail water vaporization into confinement enclosure. Equipments would be simultaneously subjected to temperature and pressure increase, chemical spray, and radiation action of reactor core products. Some equipments have to work after accident in order to stop reactor running and blow out water calories. Usually, in France, accident simulation tests are carried out sequentialy: irradiation followed by thermodynamical and chemical tests. Equipments working is essentially due to those polymer materials behaviour. Is the polymers behaviour the same when they are either subjected to sequential test, or an accident (simultaneous action of irradiation and thermodynamical and chemical sequence). In order to answer to this question, nine polymer materials were subjected to simultaneous and sequential test in CESAR cell. Experiments were carried out in CESAR device with thermodynamical chocks and a temperature and pressure decrease profil in presence or without irradiation. So, the test is either simultaneous or sequential. Mechanical properties change are determined for the following polymeric materials. Two polyamide-imide varnishes used in motors and coils; one epoxydic resin, glass fiber charged (electrical insulating); polyphenylene sulfide, glass fiber charged, the Ryton R4 (electrical insulating); three elastomeric materials: Hypalon, fire proof by bromine or by alumina EPDM (cables jacket); VAMAC which is a polyethylene methyl polymethacrylate copolymer; then a silicon thermoset material glass fiber charged (electrical insulating). After test, usually, mechanical and electrical properties change of polymer materials show sequential experiment is more severe than simultaneous test however, Hypalon does not follow this law. For this polymer simultaneous test appears more severe than sequential experiment [fr

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

HTGR accident initiation and progression analysis status report. Volume VIII. Responses to comments on AIPA status report

Energy Technology Data Exchange (ETDEWEB)

Raabe, P.H.

1977-01-01

The first seven volumes of the report series provide formal documentation of the status of the ERDA-sponsored Accident Initiation and Progression Analysis (AIPA) study as of the end of FY75. That portion of the report was given broad distribution to government agencies, industrial organizations, and academic institutions. Comments on the Status Report have been actively solicited from these and other organizations. The volume presented (the eighth in the AIPA Status Report) documents all of the formal written comments that have been received as of September 30, 1976, together with the responses to those comments. The comments as presented are direct quotations from the manuscripts as submitted by the reviewers; none have been paraphrased. The comments are presented in the same order as submitted by the reviewers and are generally addressed individually.

HTGR accident initiation and progression analysis status report. Volume VIII. Responses to comments on AIPA status report

International Nuclear Information System (INIS)

Raabe, P.H.

1977-01-01

The first seven volumes of the report series provide formal documentation of the status of the ERDA-sponsored Accident Initiation and Progression Analysis (AIPA) study as of the end of FY75. That portion of the report was given broad distribution to government agencies, industrial organizations, and academic institutions. Comments on the Status Report have been actively solicited from these and other organizations. The volume presented (the eighth in the AIPA Status Report) documents all of the formal written comments that have been received as of September 30, 1976, together with the responses to those comments. The comments as presented are direct quotations from the manuscripts as submitted by the reviewers; none have been paraphrased. The comments are presented in the same order as submitted by the reviewers and are generally addressed individually

Preliminary report about nuclear accident of Chernobylsk reactor

International Nuclear Information System (INIS)

Oliveira, A.R. de.

1986-07-01

The preliminary report of nuclear accident at Chernobyl, in URSS is presented. The Chernobyl site is located geographically and the RBMK type reactors - initials of russian words which mean high power pressure tube reactors are described. The conditions of reactor operation in beginning of accident, the events which lead to reactor destruction, the means to finish the fire, the measurements adopted by Russian in the accident location, the estimative of radioactive wastes, the meteorological conditions during the accident, the victims and medical assistence, the sanitary aspects and consequences for population, the evaluation of radiation doses received at small and medium distance and the estimative of reffered doses by population attained are presented. The official communication of Russian Minister Council and the declaration of IAEA general manager during a collective interview in Moscou are annexed. (M.C.K.) [pt

Macro Data Analysis of Traffic Accidents in Indonesia

Directory of Open Access Journals (Sweden)

Annisa Jusuf

2017-04-01

Full Text Available This paper presents a macro data analysis of Indonesian road accidents in the form of statistical data. Traffic accidents and their subsequent fatalities bring enormous social and economic consequences. A good understanding of the problem is expected to initiate major action toward the improvement of road and vehicle safety. One important milestone is the collection and analysis of road accident data. The results from this study portray the ‘tangled threads’ problem of traffic in Indonesia. The population number and number of vehicles have increased steadily, as has been accurately predicted by experts. Meanwhile, there is not enough infrastructure growth. Motorcycles are the main contributor to traffic accidents and fatalities due to their popularity as an effective vehicle to jump traffic jams. The ‘tangled threads’ need an extremely creative and comprehensive solution.

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

Strategy generator in computerized accident management support system

International Nuclear Information System (INIS)

Sirola, M.

1994-02-01

An increased interest for research in the field of accident management of nuclear power plants can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accidents in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The idea of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information from the plant will help the strategy planning. (orig.). (40 refs., 20 figs.)

Long-term effects of ionizing radiation after the Chernobyl accident: Possible contribution of historic dose.

Science.gov (United States)

Omar-Nazir, Laila; Shi, Xiaopei; Moller, Anders; Mousseau, Timothy; Byun, Soohyun; Hancock, Samuel; Seymour, Colin; Mothersill, Carmel

2018-08-01

The impact of the Chernobyl NPP accident on the environment is documented to be greater than expected, with higher mutation rates than expected at the current, chronic low dose rate. In this paper we suggest that the historic acute exposure and resulting non-targeted effects (NTE) such as delayed mutations and genomic instability could account at least in part for currently measured mutation rates and provide an initial test of this concept. Data from Møller and Mousseau on the phenotypic mutation rates of Chernobyl birds 9-11 generations post the Chernobyl accident were used and the reconstructed dose response for mutations was compared with delayed reproductive death dose responses (as a measure of genomic instability) in cell cultures exposed to a similar range of doses. The dose to birds present during the Chernobyl NPP accident was reconstructed through the external pathway due to Cs-137 with an estimate of the uncertainty associated with such reconstruction. The percentage of Chernobyl birds several generations after the accident without mutations followed the general shape of the clonogenic survival percentage of the progeny of irradiated cells, and it plateaued at low doses. This is the expected result if NTE of radiation are involved. We suggest therefore, that NTE induced by the historic dose may play a role in generating mutations in progeny many generations following the initial disaster. Copyright © 2018 Elsevier Inc. All rights reserved.

Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

International Nuclear Information System (INIS)

Journeau, Ch.

2008-01-01

The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

Directory of Open Access Journals (Sweden)

Robertas Alzbutas

2015-01-01

Full Text Available The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate.

Severe accident recriticality analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. E-mail: wiktor.frid@ski.se; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H

2001-11-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s{sup -1} injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

Severe accident recriticality analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H.

2001-01-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s -1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

Current status of accident analysis for Korean HCCR TBS

Energy Technology Data Exchange (ETDEWEB)

Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Current status of accident analysis for Korean HCCR TBS

International Nuclear Information System (INIS)

Ahn, Mu-Young; Jin, Hyung Gon; Cho, Seungyon; Lee, Dong Won; Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin

2014-01-01

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements

A review of accidents, prevention and mitigation options related to hazardous gases

International Nuclear Information System (INIS)

Fthenakis, V.M.

1993-05-01

Statistics on industrial accidents are incomplete due to lack of specific criteria on what constitutes a release or accident. In this country, most major industrial accidents were related to explosions and fires of flammable materials, not to releases of chemicals into the environment. The EPA in a study of 6,928 accidental releases of toxic chemicals revealed that accidents at stationary facilities accounted for 75% of the total number of releases, and transportation accidents for the other 25%. About 7% of all reported accidents (468 cases) resulted in 138 deaths and 4,717 injuries ranging from temporary respiratory problems to critical injuries. In-plant accidents accounted for 65% of the casualties. The most efficient strategy to reduce hazards is to choose technologies which do not require the use of large quantities of hazardous gases. For new technologies this approach can be implemented early in development, before large financial resources and efforts are committed to specific options. Once specific materials and options have been selected, strategies to prevent accident initiating events need to be evaluated and implemented. The next step is to implement safety options which suppress a hazard when an accident initiating event occurs. Releases can be prevented or reduced with fail-safe equipment and valves, adequate warning systems and controls to reduce and interrupt gas leakage. If an accident occurs and safety systems fail to contain a hazardous gas release, then engineering control systems will be relied on to reduce/minimize environmental releases. As a final defensive barrier, the prevention of human exposure is needed if a hazardous gas is released, in spite of previous strategies. Prevention of consequences forms the final defensive barrier. Medical facilities close by that can accommodate victims of the worst accident can reduce the consequences of personnel exposure to hazardous gases

Identification of Initiating Events for PGSFR

International Nuclear Information System (INIS)

Kim, Jintae; Jae, Moosung

2016-01-01

The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150

Identification of Initiating Events for PGSFR

Energy Technology Data Exchange (ETDEWEB)

Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150.

Identification of the security threshold by logistic regression applied to fuel under accident conditions

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de, E-mail: dsgomes@ipen.br, E-mail: bdbfilho@ipen.br, E-mail: fabio@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

2015-07-01

A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

Identification of the security threshold by logistic regression applied to fuel under accident conditions

International Nuclear Information System (INIS)

Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de; Giovedi, Claudia

2015-01-01

A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

Radioactive release during nuclear accidents in Chernobyl and Fukushima

Science.gov (United States)

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

2018-01-01

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

SWR-1000 concept on control of severe accidents

International Nuclear Information System (INIS)

Meyer, P.J.

1998-01-01

It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

The Chernobyl nuclear accident and its consequences

International Nuclear Information System (INIS)

1986-01-01

An AAEC Task Group was set up shortly after the accident at the Chernobyl Nuclear Power Plant to monitor and evaluate initial reports and to assess the implications for Australia. The Task Group issued a preliminary report on 9 May 1986. On 25-29 August 1986, the USSR released details of the accident and its consequences and further information has become available from the Nuclear Energy Agency of OECD and the World Health Organisation. The Task Group now presents a revised report summarising this information and commenting on the consequences from the Australian viewpoint

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

1974-01-01

The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2001-01-01

The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

A comparison of U.S. and European methods for accident scenario, identificaton, selection and quantification

International Nuclear Information System (INIS)

Cadwallader, L.C.; Djerassi, H.; Lampin, I.

1989-10-01

This paper presents a comparison of the varying methods used to identify and select accident-initiating events for safety analysis and probabilistic risk assessment (PRA). Initiating events are important in that they define the extent of a given safety analysis or PRA. Comprehensiveness in identification and selection of initiating events is necessary to ensure that a thorough analysis is being performed. While total completeness cannot ever be realized, inclusion of all safety significant events can be attained. The European approach to initiating event identification and selection arises from within a newly developed Safety Analysis methodology framework. This is a functional approach, with accident initiators based on events that will cause a system or facility loss of function. The US method divides accident initiators into two groups, internal accident initiators into two groups, internal and external events. Since traditional US PRA techniques are applied to fusion facilities, the recommended PRA-based approach is a review of historical safety documents coupled with a facility-level Master Logic Diagram. The US and European methods are described, and both are applied to a proposed International Thermonuclear Experiment Reactor (ITER) Magnet System in a sample problem. Contrasts in the US and European methods are discussed. Within their respective frameworks, each method can provide the comprehensiveness of safety-significant events needed for a thorough analysis. 4 refs., 8 figs., 11 tabs

Assessment of CRBR core disruptive accident energetics

International Nuclear Information System (INIS)

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

1984-03-01

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

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

1991-01-01

The use of NUREG-1150 and similar probabilistic risk assessments in the Nuclear Regulatory Commission (NRC) and industry risk management programs is discussed. Risk management is more comprehensive than the commonly used term accident management. Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

1990-01-01

The use of NUREG-1150 and similar Probabilistic Risk Assessments in NRC and industry risk management programs is discussed. ''Risk management'' is more comprehensive than the commonly used term ''accident management.'' Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed. 2 refs., 3 figs

Deepwater Horizon Accident Investigation Report

International Nuclear Information System (INIS)

2010-09-01

from any investigation conducted by other companies involved in the accident, and it did not review its analyses, conclusions or recommendations with any other company or investigation team. Also, at the time this report was written, other investigations, such as the U.S. Coast Guard and Bureau of Ocean Energy Management, Regulation and Enforcement Joint Investigation and the President's National Commission were ongoing. While the understanding of this accident will continue to develop with time, the information in this report can support learning and the prevention of a recurrence. The accident on April 20, 2010, involved a well integrity failure, followed by a loss of hydrostatic control of the well. This was followed by a failure to control the flow from the well with the BOP equipment, which allowed the release and subsequent ignition of hydrocarbons. Ultimately, the BOP emergency functions failed to seal the well after the initial explosions. During the course of the investigation, the team used fault tree analysis to define and consider various scenarios, failure modes and possible contributing factors. Eight key findings related to the causes of the accident emerged: (1) The annulus cement barrier did not isolate the hydrocarbons; (2) The shoe track barriers did not isolate the hydrocarbons; (3) The negative-pressure test was accepted although well integrity had not been established; (4) Influx was not recognized until hydrocarbons were in the riser; (5) Well control response actions failed to regain control of the well; (6) Diversion to the mud gas separator resulted in gas venting onto the rig; (7) The fire and gas system did not prevent hydrocarbon ignition; (8) The BOP emergency mode did not seal the well.

Deepwater Horizon Accident Investigation Report

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-09-15

separately from any investigation conducted by other companies involved in the accident, and it did not review its analyses, conclusions or recommendations with any other company or investigation team. Also, at the time this report was written, other investigations, such as the U.S. Coast Guard and Bureau of Ocean Energy Management, Regulation and Enforcement Joint Investigation and the President's National Commission were ongoing. While the understanding of this accident will continue to develop with time, the information in this report can support learning and the prevention of a recurrence. The accident on April 20, 2010, involved a well integrity failure, followed by a loss of hydrostatic control of the well. This was followed by a failure to control the flow from the well with the BOP equipment, which allowed the release and subsequent ignition of hydrocarbons. Ultimately, the BOP emergency functions failed to seal the well after the initial explosions. During the course of the investigation, the team used fault tree analysis to define and consider various scenarios, failure modes and possible contributing factors. Eight key findings related to the causes of the accident emerged: (1) The annulus cement barrier did not isolate the hydrocarbons; (2) The shoe track barriers did not isolate the hydrocarbons; (3) The negative-pressure test was accepted although well integrity had not been established; (4) Influx was not recognized until hydrocarbons were in the riser; (5) Well control response actions failed to regain control of the well; (6) Diversion to the mud gas separator resulted in gas venting onto the rig; (7) The fire and gas system did not prevent hydrocarbon ignition; (8) The BOP emergency mode did not seal the well.

Occupational Accidents with Agricultural Machinery in Austria.

Science.gov (United States)

Kogler, Robert; Quendler, Elisabeth; Boxberger, Josef

2016-01-01

The number of recognized accidents with fatalities during agricultural and forestry work, despite better technology and coordinated prevention and trainings, is still very high in Austria. The accident scenarios in which people are injured are very different on farms. The common causes of accidents in agriculture and forestry are the loss of control of machine, means of transport or handling equipment, hand-held tool, and object or animal, followed by slipping, stumbling and falling, breakage, bursting, splitting, slipping, fall, and collapse of material agent. In the literature, a number of studies of general (machine- and animal-related accidents) and specific (machine-related accidents) agricultural and forestry accident situations can be found that refer to different databases. From the database Data of the Austrian Workers Compensation Board (AUVA) about occupational accidents with different agricultural machinery over the period 2008-2010 in Austria, main characteristics of the accident, the victim, and the employer as well as variables on causes and circumstances by frequency and contexts of parameters were statistically analyzed by employing the chi-square test and odds ratio. The aim of the study was to determine the information content and quality of the European Statistics on Accidents at Work (ESAW) variables to evaluate safety gaps and risks as well as the accidental man-machine interaction.

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.

1991-01-01

Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

Structural evaluation of electrosleeved tubes under severe accident transients

International Nuclear Information System (INIS)

Majumdar, S.

1999-01-01

A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients

Traffic Accidents Involving Cyclists Identifying Causal Factors Using Questionnaire Survey, Traffic Accident Data, and Real-World Observation.

Science.gov (United States)

Oikawa, Shoko; Hirose, Toshiya; Aomura, Shigeru; Matsui, Yasuhiro

2016-11-01

The purpose of this study is to clarify the mechanism of traffic accidents involving cyclists. The focus is on the characteristics of cyclist accidents and scenarios, because the number of traffic accidents involving cyclists in Tokyo is the highest in Japan. First, dangerous situations in traffic incidents were investigated by collecting data from 304 cyclists in one city in Tokyo using a questionnaire survey. The survey indicated that cyclists used their bicycles generally while commuting to work or school in the morning. Second, the study investigated the characteristics of 250 accident situations involving cyclists that happened in the city using real-world bicycle accident data. The results revealed that the traffic accidents occurred at intersections of local streets, where cyclists collided most often with vehicles during commute time in the morning. Third, cyclists' behavior was observed at a local street intersection in the morning in the city using video pictures. In one hour during the morning commute period, 250 bicycles passed through the intersection. The results indicated that one of the reasons for traffic accidents involving cyclists might be the combined effect of low visibility, caused by the presence of box-like building structures close to the intersections, and the cyclists' behavior in terms of their velocity and no confirming safety. It was observed that, on average, bicycle velocity was 3.1 m/s at the initial line of an intersection. The findings from this study could be useful in developing new technologies to improve cyclist safety, such as alert devices for cyclists and vehicle drivers, wireless communication systems between cyclists and vehicle drivers, or advanced vehicles with bicycle detection and collision mitigation systems.

Low level waste shipment accident lessons learned

International Nuclear Information System (INIS)

Rast, D.M.; Rowe, J.G.; Reichel, C.W.

1995-01-01

On October 1, 1994 a shipment of low-level waste from the Fernald Environmental Management Project, Fernald, Ohio, was involved in an accident near Rolla, Missouri. The accident did not result in the release of any radioactive material. The accident did generate important lessons learned primarily in the areas of driver and emergency response communications. The shipment was comprised of an International Standards Organization (ISO) container on a standard flatbed trailer. The accident caused the low-level waste package to separate from the trailer and come to rest on its top in the median. The impact of the container with the pavement and median inflicted relatively minor damage to the container. The damage was not substantial enough to cause failure of container integrity. The success of the package is attributable to the container design and the packaging procedures used at the Fernald Environmental Management Project for low-level waste shipments. Although the container survived the initial wreck, is was nearly breached when the first responders attempted to open the ISO container. Even though the container was clearly marked and the shipment documentation was technically correct, this information did not identify that the ISO container was the primary containment for the waste. The lessons learned from this accident have DOE complex wide applicability. This paper is intended to describe the accident, subsequent emergency response operations, and the lessons learned from this incident

Reactor accidents of four decades

International Nuclear Information System (INIS)

Szabo, Z.

1982-11-01

The report covers the period between 1942 and June 30, 1982. A detailed description and a comparative analysis of reactor accidents and chemical-processing-plant excursions are presented. The analysis takes into account the following points: causes (design, maintenance, operation); events (initiating event and sequence of events); consequences (environmental impacts, personnel effects and equipment damages). (author)

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

International Nuclear Information System (INIS)

Guntay, Salih; Dehbi, Abdel; Suckow, Detlef; Birchley, Jon

2002-01-01

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

Examining accident reports involving autonomous vehicles in California

Science.gov (United States)

Nader, Nazanin; Eurich, Sky O.; Tripp, Michelle; Varadaraju, Naresh

2017-01-01

Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017). The data provides important information on autonomous vehicles accidents’ dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama. PMID:28931022

[A monitoring system for work-related accidents in Piracicaba, São Paulo, Brazil].

Science.gov (United States)

Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

2005-01-01

The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives are triggered by identification of sentinel events; spatial distribution analysis of work-related accidents is a basic tool in designing accident awareness strategies and accident prevention policies. The system was implemented in November 2003 and by October 2004 had identified 5,320 work-related accidents, or a 3.8% annual proportional incidence of work-related accidents in the municipal area. We illustrate spatial analysis of registered work-related accidents and present a detailed investigation of one example of a serious accident.

Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding

Energy Technology Data Exchange (ETDEWEB)

None, None

2015-04-01

Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal

Accident and emergency management

International Nuclear Information System (INIS)

Andersen, V.; Moellenbach, K.; Heinonen, R.; Jakobsson, S.; Kukko, T.; Berg, Oe.; Larsen, J.S.; Westgaard, T.; Magnusson, B.; Andersson, H.; Holmstroem, C.; Brehmer, B.; Allard, R.

1988-06-01

There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

Testing to determine the leakage behavior of inflatable seals subject to severe accident loadings

International Nuclear Information System (INIS)

Parks, M.B.

1988-01-01

Under the sponsorship of the United States Nuclear Regulatory Commission, Sandia National Laboratories is currently developing test validated methods to predict the pressure capacity, at elevated temperatures, of light water reactor (LWR) nuclear containment vessels subject to loads well beyond their design basis - the so-called severe accident. Scale model tests of containments with the major penetrations represented have been carried to functional failure by internal pressurization. Also, combined pressure and elevated temperature tests of typical compression seals and gaskets, a full size personnel airlock, and of typical electrical penetration assemblies (EPAs), have been conducted in order to better understand the leakage behavior of containment penetrations. Because inflatable seals are also a part of the pressure boundary of some containments, it is important to understand their leakage behavior as well. This paper discusses the results of tests that were performed to better define the leakage behavior of inflatable seals when subjected to loads well beyond their design basis

Qualification of class 1e equipment: regulation, technological margins and test experience

International Nuclear Information System (INIS)

Pasco, Y.; Le Meur, M.; Henry, J.Y.; Droger, J.P.; Morange, E.; Roubault, J.

1986-10-01

French regulation requires licensee to qualify electrical equipment important to safety for service in nuclear power plants to ensure that the equipment can perform its safety function under the set of plausible operating conditions. The French regulatory texts entitled Fundamental safety rules have classified safety related electrical equipment in three main categories: k1, k2, k3, according to their location and operating conditions. The definition of a design basis accident test profile must account for margins applied to thermal hydraulic code outputs. Specific safety margins was added to cover uncertainties in qualification test representativity. Up to now, accidental sequence studies have shown the validity of such a qualification test profile. On the other hand, the results from post accident simulation tests have shown that it is useful not only to validate post accident operating life but also to reveal failures initiated during previous tests [fr

Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

International Nuclear Information System (INIS)

Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

2015-01-01

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

2014-01-01

Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

Management, administrative and operational causes of the accident: Chernobyl nuclear power station

International Nuclear Information System (INIS)

Anastas, G.

1996-01-01

Full text: The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products (including 70-100 P Bq of 137 Cs). The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'

Unavoidable Accident

OpenAIRE

Grady, Mark F.

2009-01-01

In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

International Nuclear Information System (INIS)

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.; Fischer, A.K.; Meek, C.C.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references

Severe Accident Recriticality Analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Severe Accident Recriticality Analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B 4 C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

The post-accident nuclear issue: the new crisis expertise challenges for the IRSN

International Nuclear Information System (INIS)

Champion, D.

2010-01-01

The author reports the work performed by two work groups conducted by the IRSN (the French Radioprotection and Nuclear Safety Institute), the first one on the issue of assessment of radiological and dosimetric consequences in a post-accident situation, and the second one on hypotheses to be used to perform predictive assessments of these consequences. First dealing with the end of the emergency phase, he describes how to anticipate actions of protection against immediate post-accident consequences: orientation of the expertise strategy based on the CODIRPA's doctrine, post-accident zoning based on predictive indicators, use of reasonably prudent hypotheses for the first predictive assessments, importance of initial radioactive deposits to perform predictive assessments. Then, the author presents an iterative method of assessment of post-accident consequences: organization of environment radioactivity measurement programmes, periodic update of mapping of initial deposit and of actual deposits at a given time

Analysis of severe core damage accident progression for the heavy water reactor

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

[Drugs and occupational accident].

Science.gov (United States)

Bratzke, H; Albers, C

1996-02-01

In a case of a fatal occupational accident (construction worker, fall from roof, urine test positive for cocaine and THC, e.g. cannabis) the question arised to what extent those drug-related occupational accidents occur. In the literature only few cases, mainly dealing with cannabis influence, have been reported, however, a higher number is suspected. Cocaine and other stimulating drugs (amphetamine) are more often used to increase physical fitness. By direct or indirect interference with vigilance these compounds may provoke accidents. Due to the lack of a legal basis proving of the influence of drugs at the working place is still very limited, although highly sensitive chemical-toxicological assay procedures are available to detect even the chronic abuse (in hair). In the general conditions of accident insurances a compensation is excluded when alcohol is involved, but drugs are not mentioned. It is indeed difficult to establish a concentration limit for drugs like that existing for alcohol (1.1%). In each case the assay of the drug involved and exact knowledge of its specific effects is in an essential prerequisite to prove the causal relationship.

Cardiac Injury After All-Terrain Vehicle Accidents in 2 Children and a Review of the Literature.

Science.gov (United States)

Ngo, Kimberly D; Pian, Phillip; Hanfland, Robert; Nichols, Christopher S; Merritt, Glenn R; Campbell, David; Ing, Richard J

2016-07-01

All-terrain vehicle (ATV) accidents leading to severe morbidity and mortality are common. At our institution, 2 children presented within weeks of each other after ATV accidents. Both children required cardiac valve surgery. The surgical management of these 2 children is discussed, and the literature is reviewed. On initial patient presentation, the diagnosis of a ruptured cardiac valve or ventricular septal defect (VSD) associated with these types of accidents is often delayed. We propose that patients presenting with evidence of high-energy blunt thoracic trauma after an ATV accident should undergo an electrocardiogram, cardiac enzyme assessment, and cardiac echocardiogram as part of the initial work-up to rule out significant myocardial injury.

The relationships between organizational and individual variables to on-the-job driver accidents and accident-free kilometres.

Science.gov (United States)

Caird, J K; Kline, T J

2004-12-01

Highway fatalities are the leading cause of fatal work injuries in the US, accounting for approximately 1 in 4 of the 5900 job-related deaths during 2001. The present study focused on the contribution of organizational factors and driver behaviours to on-the-job driving accidents in a large Western Canadian corporation. A structural equation modelling (SEM) approach was used which allows researchers to test a complex set of relationships within a global theoretical framework. A number of scales were used to assess organizational support, driver errors, and driver behaviours. The sample of professional drivers that participated allowed the recording of on-the-job accidents and accident-free kilometres from their personnel files. The pattern of relationships in the fitted model, after controlling for exposure and social desirability, provides insight into the role of organizational support, planning, environment adaptations, fatigue, speed, errors and moving citations to on-the-job accidents and accident-free kilometres. For example, organizational support affected the capacity to plan. Time to plan work-related driving was found to predict accidents, fatigue and adaptations to the environment. Other interesting model paths, SEM limitations, future research and recommendations are elaborated.

Transport of fresh MOX fuel assemblies for the Monju initial core

International Nuclear Information System (INIS)

Kurakami, J.; Ouchi, Y.; Usami, M.

1997-01-01

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

Domino effect in chemical accidents: main features and accident sequences

OpenAIRE

Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

2010-01-01

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes a...

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Crash tests for passenger cars and their relationship to the actual accident occurrence

International Nuclear Information System (INIS)

Appel, Hermann; Lutter, Gerhard; Sigmund, Thomas

1994-01-01

Current consensus about crash tests implies that, for verification of self-protection of a vehicle or its occupants, at least three full size tests with the following specifications are necessary:(1)frontal impact against a rigid, non-moving 0 -barrier with 100% overlap;(2)frontal offset impact against a rigid, non-moving 15 -barrier with 50% overlap (impact speed between 50 and 55kmh -1 );(3)side impact of a moving deformable barrier; preferably according to EEVC-method (impact speed 50kmh -1 ).From the social general view it is not sufficient to test only the self-protection of the vehicle and to give most importance to the front of the vehicle. The other factors of passive safety, partner protection and compatibility, respectively, have to be included, as two thirds of the cost of injuries originates from car-to-car accidents, and only one third from vehicle collisions against fixed objects. It follows that at least one additional test of compatibility has to be added to those mentioned above. It has to be investigated whether this compatibility test could be a frontal impact against a controllably deformable barrier and could substitute one or even two of the first-mentioned tests. ((orig.))

Plan for IER-443 Testing of the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor

Energy Technology Data Exchange (ETDEWEB)

Scorby, J. C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Garbett, S. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Auld, G. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Horrne, A. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Beller, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haught, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Woodrow, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-07-24

This document provides the scope and details of the “Plan for Testing the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor”. Due to the relative simplicity of the testing goals, scope, and methodology, the NCSP Manager approved execution of the test when ready. No preliminary CED-1 or final design CED-2 reports were required or issued. The test will subject Criticality Accident Alarm System (CAAS) detectors supplied by Y- 12 and AWE to very intense and short duration mixed neutron and gamma radiation fields. The goals of the test will be to (1) substantiate functionality, for both existing and newly acquired Y- 12 CAAS detectors, and (2) the ability of the AWE detectors to provide quality temporal dose information after a hypothetical criticality accident. ANSI/ANS-8.3.1997 states that the “system shall be sufficiently robust as to actuate an alarm signal when exposed to the maximum radiation expected”, which has been defined at Y-12, in Documented Safety Analyses (DSAs), to be a dose rate of 10 Rad/s. ANSI/ANS-8.3.1997 further states that “alarm actuation shall occur as a result of a minimum duration transient” which may be assumed to be 1 msec. The pulse widths and dose rates which will be achieved in this test will exceed these requirements. Pulsed radiation fields will be produced by the Godiva IV fast metal burst reactor at the National Criticality Experimental Research Center (NCERC) at the Nevada National Security Site (NNSS). The magnitude of the pulses and the relative distances to the detectors will be varied to afford a wide range of radiation fluence and pulse widths. The magnitude of the neutron and gamma fields will be determined by reactor temperature rise to fluence and dose conversions which have been previously established through extensive measurements performed under IER-147. The requirements for CAAS systems to detect and alarm under a “minimum accident of concern” as well as other

Strategies for the prevention and mitigation of severe accidents

International Nuclear Information System (INIS)

Ader, C.; Heusener, G.; Snell, V.G.

1999-01-01

The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

National and regional analysis of road accidents in Spain.

Science.gov (United States)

Tolón-Becerra, A; Lastra-Bravo, X; Flores-Parra, I

2013-01-01

In Spain, the absolute fatality figures decreased almost 50 percent between 1998 and 2009. Despite this great effort, road mortality is still of great concern to political authorities. Further progress requires efficient road safety policy based on an optimal set of measures and targets that consider the initial conditions and characteristics in each region. This study attempts to analyze road accidents in Spain and its provinces in time and space during 1998-2009. First, we analyzed daily, monthly, and nationwide (NUTS 0) development of road accidents, the correlation between logarithmic transformations of road accidents and territorial and socioeconomic variables, the causality by simple linear regression of road accidents and territorial and socioeconomic variables, and preliminary frequency by fast Fourier transform. Then we analyzed the annual trend in accidents in the Spanish provinces (NUTS 3) and found a correlation between the logarithmic transformations of the mortality rate, fatalities per fatal accident, and accidents resulting in injuries per inhabitant variables and population, population density, gross domestic product (GDP), length of road network, and area. Finally, causality was analyzed by simple linear regression. The most outstanding results were the negative correlation between mortality rate and population density in Spanish provinces, which has increased over time, and that road accidents in Spain have an approximate periodicity of 57 days. The fast Fourier transform analysis of road accident frequency in Spain was useful in identifying the periodic, harmonic components of accidents and casualties. The periodicity observed both for the period 1998-2009 and by year showed that the highest intensity in road accidents was bimonthly, despite the lower number of accidents and casualties in the spectra of amplitude and power and efforts to reduce the intensity and concentration during off-season travel (summer and December).

PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise

International Nuclear Information System (INIS)

Hill, Robin L.; Conrady, Matthew M.

2011-01-01

This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

Improving aircraft accident forecasting for an integrated plutonium storage facility

International Nuclear Information System (INIS)

Rock, J.C.; Kiffe, J.; McNerney, M.T.; Turen, T.A.

1998-06-01

Aircraft accidents pose a quantifiable threat to facilities used to store and process surplus weapon-grade plutonium. The Department of Energy (DOE) recently published its first aircraft accident analysis guidelines: Accident Analysis for Aircraft Crash into Hazardous Facilities. This document establishes a hierarchy of procedures for estimating the small annual frequency for aircraft accidents that impact Pantex facilities and the even smaller frequency of hazardous material released to the environment. The standard establishes a screening threshold of 10 -6 impacts per year; if the initial estimate of impact frequency for a facility is below this level, no further analysis is required. The Pantex Site-Wide Environmental Impact Statement (SWEIS) calculates the aircraft impact frequency to be above this screening level. The DOE Standard encourages more detailed analyses in such cases. This report presents three refinements, namely, removing retired small military aircraft from the accident rate database, correcting the conversion factor from military accident rates (accidents per 100,000 hours) to the rates used in the DOE model (accidents per flight phase), and adjusting the conditional probability of impact for general aviation to more accurately reflect pilot training and local conditions. This report documents a halving of the predicted frequency of an aircraft impact at Pantex and points toward further reductions

Risk assessment for long-term post-accident sequences

International Nuclear Information System (INIS)

Ellia-Hervy, A.; Ducamp, F.

1987-11-01

Probabilistic risk analysis, currently conducted by the CEA (French Atomic Energy Commission) for the French replicate series of 900 MWe power plants, has identified accident sequences requiring long-term operation of some systems after the initiating event. They have been named long-term sequences. Quantification of probabilities of such sequences cannot rely exclusively on equipment failure-on-demand data: it must also take into account operating failures, the probability of which increase with time. Specific studies have therefore been conducted for a number of plant systems actuated during these long-term sequences. This has required: - Definition of the most realistic equipment utilization strategies based on existing emergency procedures for 900 MWe French plants. - Evaluation of the potential to repair failed equipment, given accessibility, repair time, and specific radiation conditions for the given sequence. - Definition of the event bringing the long-term sequence to an end. - Establishment of an appropriate quantification method, capable of taking into account the evolution of assumptions concerning equipment utilization strategies or repair conditions over time. The accident sequence quantification method based on realistic scenarios has been used in the risk assessment of the initiating event loss of reactor coolant accident occurring at power and at shutdown. Compared with the results obtained from conventional methods, this method redistributes the relative weight of accident sequences and also demonstrates that the long term can be a significant contribution to the probability of core melt

From learning from accidents to teaching about accident causation and prevention: Multidisciplinary education and safety literacy for all engineering students

International Nuclear Information System (INIS)

Saleh, Joseph H.; Pendley, Cynthia C.

2012-01-01

In this work, we argue that system accident literacy and safety competence should be an essential part of the intellectual toolkit of all engineering students. We discuss why such competence should be taught and nurtured in engineering students, and provide one example for how this can be done. We first define the class of adverse events of interest as system accidents, distinct from occupational accidents, through their (1) temporal depth of causality and (2) diversity of agency or groups and individuals who influence or contribute to the accident occurrence/prevention. We then address the question of why the interest in this class of events and their prevention, and we expand on the importance of system safety literacy and the contributions that engineering students can make in the long-term towards accident prevention. Finally, we offer one model for an introductory course on accident causation and system safety, discuss the course logistics, material and delivery, and our experience teaching this subject. The course starts with the anatomy of accidents and is grounded in various case studies; these help illustrate the multidisciplinary nature of the subject, and provide the students with the important concepts to describe the phenomenology of accidents (e.g., initiating events, accident precursor or lead indicator, and accident pathogen). More importantly, the case studies invite a deep reflection on the underlying failure mechanisms, their generalizability, and the various safety levers for accident prevention. The course then proceeds to an exposition of defense-in-depth, safety barriers and principles, essential elements for an education in accident prevention, and it concludes with a presentation of basic concepts and tools for uncertainty and risk analysis. Educators will recognize the difficulties in designing a new course on such a broad subject. It is hoped that this work will invite comments and contributions from the readers, and that the journal will

Benchmark accident scenarios for nuclear powered warship visits to Australian ports

International Nuclear Information System (INIS)

Frikken, A.J.

1996-01-01

Full text: Port safety arrangements for visits of nuclear powered warships (NPWs) to Australian ports require compliance with a number of Conditions of Entry. One of these Conditions of Entry is the provision of 'an operating safety organisation, competent to produce a suitable radiation monitoring program and able to initiate actions and provide services necessary to safeguard the public in the event of a release of radioactivity following an accident'. The States and Territories which receive NPW visits have the responsibility for this contingency planning, although the Commonwealth provides assistance through the Visiting Ships Panel Nuclear (VSP(N)). The visit of a NPW to an Australian port may only proceed if the VSP(N) is satisfied that the port safety plan has been exercised in sufficient depth to demonstrate its adequacy and efficacy. Emergency exercises are held on a regular basis in Australian ports which have been validated for visits by NPWs to ensure compliance with the Conditions of Entry. Important aspects of these exercises are the procedures for estimating radiation doses to members of the public following an accident, and the process for making decisions on the need for countermeasures based on the results of dose estimates. To exercise these aspects of the emergency response, detailed emergency exercise scenarios, including simulated radiological monitoring data, are required. To date, emergency exercises have usually been based on a severe and highly improbable scenario, termed the Reference Accident, which is used to assess the suitability of ports for visits by NPW's. The repeated adoption of this scenario does not adequately test the flexibility of the emergency plans to cope with all possible accidents, particularly more likely, less severe accidents. At the request of the VSP(N), the Nuclear Safety Bureau has analysed a spectrum of NPW accident scenarios and developed a set of Bench Mark Accident (BMA) scenarios for emergency response

Occupational Radiation Protection in Severe Accident Management

International Nuclear Information System (INIS)

2015-01-01

As an early response to the Fukushima Daiichi NPP accident, the Information System on Occupational Exposure (ISOE) Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011: - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers /responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE official participants and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the-art ISOE report on best radiation protection management practices for proper radiation

Testing for Turkeys Faith-Based Community HIV Testing Initiative: An Update.

Science.gov (United States)

DeGrezia, Mary; Baker, Dorcas; McDowell, Ingrid

2018-06-04

Testing for Turkeys (TFT) HIV/hepatitis C virus (HCV) and sexually transmitted infection (STI) testing initiative is a joint effort between Older Women Embracing Life (OWEL), Inc., a nonprofit faith-based community HIV support and advocacy organization; the Johns Hopkins University Regional Partner MidAtlantic AIDS Education and Training Center (MAAETC); and the University of Maryland, Baltimore JACQUES Initiative (JI), and is now in its 11th year of providing HIV outreach, testing, and linkage to care. Since 2008, the annual TFT daylong community HIV testing and linkage to care initiative has been held 2 weeks before Thanksgiving at a faith-based center in Baltimore, Maryland, in a zip code where one in 26 adults and adolescents ages 13 years and older are living with HIV (Maryland Department of Health, Center for HIV Surveillance, Epidemiology, and Evaluation, 2017). TFT includes a health fair with vendors that supply an abundance of education information (handouts, videos, one-on-one counseling) and safer sex necessities, including male and female condoms, dental dams, and lube. Nutritious boxed lunches and beverages are provided to all attendees and volunteers. Everyone tested for HIV who stays to obtain their results is given a free frozen turkey as they exit. The Baltimore City Health Department is on hand with a confidential no-test list (persons in the state already known to have HIV) to diminish retesting of individuals previously diagnosed with HIV. However, linkage to care is available to everyone: newly diagnosed individuals and those previously diagnosed and currently out of care. Copyright © 2018 Association of Nurses in AIDS Care. Published by Elsevier Inc. All rights reserved.

Regulatory approach to enhanced human performance during accidents

International Nuclear Information System (INIS)

Palla, R.L. Jr.

1990-01-01

It has become increasingly clear in recent years that the risk associated with nuclear power is driven by human performance. Although human errors have contributed heavily to the two core-melt events that have occurred at power reactors, effective performance during an event can also prevent a degraded situation from progressing to a more serious accident, as in the loss-of-feedwater event at Davis-Besse. Sensitivity studies in which human error rates for various categories of errors in a probabilistic risk assessment (PRA) were varied confirm the importance of human performance. Moreover, these studies suggest that actions taken during an accident are at least as important as errors that occur prior to an initiating event. A program that will lead to enhanced accident management capabilities in the nuclear industry is being developed by the US Nuclear Regulatory Commission (NRC) and industry and is a key element in NRC's integration plan for closure of severe-accident issues. The focus of the accident management (AM) program is on human performance during accidents, with emphasis on in-plant response. The AM program extends the defense-in-depth principle to plant operating staff. The goal is to take advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents that are beyond the design basis. The purpose of this paper is to describe the NRC's objectives and approach in AM as well as to discuss several human performance issues that are central to AM

[a Monitoring System For Work-related Accidents In Piracicaba, São Paulo, Brazil].

OpenAIRE

Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

2015-01-01

The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives ar...

Longitudinal relationship between economic development and occupational accidents in China.

Science.gov (United States)

Song, Li; He, Xueqiu; Li, Chengwu

2011-01-01

The relativity between economic development and occupational accidents is a debated topic. Compared with the development courses of both economic development and occupational accidents in China during 1953-2008, this paper used statistic methods such as Granger causality test, cointegration test and impulse response function based on the vector autoregression model to investigate the relativity between economic development and occupational accidents in China from 1953 to 2008. Owing to fluctuation and growth scale characteristics of economic development, two dimensions including economic cycle and economic scale were divided. Results showed that there was no relationship between occupational accidents and economic scale during 1953-1978. Fatality rate per 10(5) workers was a conductive variable to gross domestic product per capita during 1979-2008. And economic cycle was an indicator to occupational accidents during 1979-2008. Variation of economic speed had important influence on occupational accidents in short term. Thus it is necessary to adjust Chinese occupational safety policy according to tempo variation of economic growth. Crown Copyright © 2010. Published by Elsevier Ltd. All rights reserved.

Severe accident considerations in Canadian nuclear power reactors

International Nuclear Information System (INIS)

Omar, A.M.; Measures, M.P.; Scott, C.K.; Lewis, M.J.

1990-08-01

This paper describes a current study on severe accidents being sponsored by the Atomic Energy Control Board (AECB) and provides background on other related Canadian work. Scoping calculations are performed in Phase I of the AECB study to establish the relative consequences of several permutations resulting from six postulated initiating events, nine containment states, and a selection of meteorological conditions and health effects mitigating criteria. In Phase II of the study, selected accidents sequences would be analyzed in detail using models suitable for the design features of the Canadian nuclear power reactors

A modular structure to accident identification using neural networks

International Nuclear Information System (INIS)

Duque Estrada, Cassius Rodrigo

2005-01-01

This work uses the accident identification method based on Artificial Neural Networks (ANN) as basic blocks of a modular structure, allowing the inclusion of new accidents to be identified without modifying the ANN already trained. This structure comprises several modules for accident identification and one module for analysis. Each identification module follows the structure of the basic block. The identification modules are responsible for the recognition of an accident belonging to the specific set of events for which it were trained. The analysis module processes the output from the identification module to determine the system response. In order to test this structure it was proposed a transient identification problem comprising fifty accidents distributed in five identification modules. The results have demonstrated that the accident identification method used as basic block of a modular structure allows the inclusion of new sets of accidents, or variations of a same accident, without modifying the ANN already trained. For this, it is enough to include into the system an specific module for this new set of accidents. (author)

Management, administrative and operational causes of the accident: Chernobyl nuclear power station

International Nuclear Information System (INIS)

Anastas, G.

1996-01-01

The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products including 70-100 PBq of 137 Cs. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'. This corrupted safety culture exacerbated the poor design of the reactor. The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. (authors)

Plant state identification using fuzzy logic in the framework of computerized accident management support (CAMS)

International Nuclear Information System (INIS)

Van Dyck, Claude

1997-05-01

CAMS (computerized accident management support) is a system that will provide assistance in case of accident in a nuclear power plant. In order to support the user in evaluating the plant state, it contains a state identification module. The state identification module provides high-level, qualitative information about the status of critical safety functions, about the availability of safety systems and about the occurrence of initiating events. This information is sent to the man-machine interface and to other CAMS modules. The state identification module is developed using a specific tool: GPS (Goal Processing System) which is based on the Goal Tree - Success Tree formalism. GPS is a tool designed to manage ''process related'' knowledge and aimed at process supervision via real-time acquisition of process variables. Fuzzy logic has been introduced in GPS in order to have smoother transitions between different states of critical safety functions and systems changes and to have a truth value associated to each piece of information provided to the user. The whole system has been tested, integrated with the rest of CAMS, on several accident scenarios. The test results are satisfactory. A brief comparison is made between the present work and previous related work at the HRP. (author)

Overview of main accident parameters in car-to-cyclist accidents for use in AEB-system test protocol

NARCIS (Netherlands)

Uittenbogaard, J.; Camp, O.M.G.C. op den; Montfort, S. van

2016-01-01

The number of fatalities in road traffic accidents in Europe is decreasing. Unfortunately, the number of fatalities among cyclists does not follow this trend with the same rate [1]. The au-tomotive industry is making a significant effort in the development and implementation of safety systems in

Self-reported accidents

DEFF Research Database (Denmark)

Møller, Katrine Meltofte; Andersen, Camilla Sloth

2016-01-01

The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

Meteorological data related to the Chernobyl accident

International Nuclear Information System (INIS)

Graziani, G.; Zarimpas, N.

1989-01-01

This report presents a detailed technical description of the JRC-Ispra comprehensive collection of meteorological information related to the Chernobyl accident and attempts an analysis of the data in order to perform an initial checking of their quality and facilitate a suitable and compact way of display

Strengthening Regulatory Effectiveness in India – Lessons Learnt from Fukushima Accident

International Nuclear Information System (INIS)

Solanki, R.

2016-01-01

Following the Fukushima Daiichi accident in Japan, one of the most important lessons learnt, among other things, was the issue of strengthening the effectiveness of the regulatory bodies. Immediately after the Fukushima accident, National level safety audits were conducted on all operating NPPs in India to review safety of NPPs in India. A national action plan has been prepared to implement the identified short term, midterm and long term measures. The assessment indicates that national response to the Fukushima Accident for safety assessment of NPPs and subsequent actions and initiatives taken for safety enhancement of the NPPs in India are in-line with the objectives of the IAEA Action plan. This paper highlights the actions taken by India in the light of Fukushima Daiichi accident in order to strengthen the regulatory effectiveness through improvements in the existing core processes, challenges faced, Insights gained from the recent initiatives on safety performance indicators and assessment of safety culture, relevant observations of IRRS mission report and Indian perspectives on the further cooperation among the member states for enhancing the regulatory effectiveness for nuclear oversight of regulated organizations. (author)

Simulation and verification studies of reactivity initiated accident by comparative approach of NK/TH coupling codes and RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Ud-Din Khan, Salah [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center; Peng, Minjun [Harbin Engineering Univ. (China). College of Nuclear Science and Technology; Yuntao, Song; Ud-Din Khan, Shahab [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; Haider, Sajjad [King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center

2017-02-15

The objective is to analyze the safety of small modular nuclear reactors of 220 MWe power. Reactivity initiated accidents (RIA) were investigated by neutron kinetic/thermal hydraulic (NK/TH) coupling approach and thermal hydraulic code i.e., RELAP5. The results obtained by these approaches were compared for validation and accuracy of simulation. In the NK/TH coupling technique, three codes (HELIOS, REMARK, THEATRe) were used. These codes calculate different parameters of the reactor core (fission power, reactivity, fuel temperature and inlet/outlet temperatures). The data exchanges between the codes were assessed by running the codes simultaneously. The results obtained from both (NK/TH coupling) and RELAP5 code analyses complement each other, hence confirming the accuracy of simulation.

Fuel temperature analysis method for channel-blockage accident in HTTR

International Nuclear Information System (INIS)

Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

1994-01-01

During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

Analysis of heavy radiological accidents in NPP and gamma-irradiators

Energy Technology Data Exchange (ETDEWEB)

Angelov, V [Civil Defence Administration, Sofia (Bulgaria); Semova, T; Bonchev, Ts [Sofia Univ. (Bulgaria). Fizicheski Fakultet

1996-12-31

A review of several heavy radiological accidents, their cause, character, radioactivity emission, victims and economical impact is presented in the form of uniform tables. Eleven cases of incidents in power plants and 4 cases of accidents involving powerful gamma irradiators are considered. Radiological accidents in Bulgaria, not connected with the Kozloduy NPP, are listed. The human factor has been identified as the main cause for most of the accidents. It is stressed that the probability of heavy accident increases at the time of reactor refuelling, repair or testing. Technical failures could be eliminated by improved check and diagnostics procedures. 2 tabs., 12 refs.

Analysis of heavy radiological accidents in NPP and gamma-irradiators

International Nuclear Information System (INIS)

Angelov, V.; Semova, T.; Bonchev, Ts.

1995-01-01

A review of several heavy radiological accidents, their cause, character, radioactivity emission, victims and economical impact is presented in the form of uniform tables. Eleven cases of incidents in power plants and 4 cases of accidents involving powerful gamma irradiators are considered. Radiological accidents in Bulgaria, not connected with the Kozloduy NPP, are listed. The human factor has been identified as the main cause for most of the accidents. It is stressed that the probability of heavy accident increases at the time of reactor refuelling, repair or testing. Technical failures could be eliminated by improved check and diagnostics procedures. 2 tabs., 12 refs

Cesium-137: psychological and social consequences of the Goiania's accident

International Nuclear Information System (INIS)

Helou, Suzana; Costa Neto, Sebastiao Benicio da

1995-01-01

The book care for radioactive accident occurred in 1987 in Goiania - brazilian city. The accident had origin by the hospitable equipment incorrect handling which contained a stainless steel capsule, in which interior there was cesium-137 chloride. The main boarded aspects are: psychological and social aspects verified after the accident; psychological and social analysis of population of Goiania three years after the accident; essay on the pertinence of Luscher's abbreviate test in psychological evaluation of the radioactive accident victims of Goiania; and psychological and mobile evaluation of intra-uterus children exposed to the radiation with cesium-137

Simulation of operator's actions during severe accident management

International Nuclear Information System (INIS)

Viktorov, A.

2015-01-01

Implementing accident management counter measures or actions to mitigate consequences of a severe accident is essential to reduce radiological risks to the public and environment. Station-specific severe accident management guidelines (SAMGs) have been developed and implemented at all Canadian nuclear power plants. Following the Fukushima Daiichi nuclear accident certain enhancements were introduced to the SAMG, namely consideration of multi-units accidents, events involving spent fuel pools, incorporation of capability offered by the portable emergency mitigating equipment, and so on. To evaluate the adequacy and usability of the SAMGs, CNSC staff initiated a number of activities including a desktop review of SAMG documentation, evaluation of SAMG implementation through exercises and interviews with station staff, and independent verification of SAMG action effectiveness. This paper focuses on the verification of SAMG actions through analytical simulations. The objectives of the work are two-folds: (a) to understand the effectiveness of SAMG-specified mitigation actions in addressing the safety challenges and (b) to check for potential negative effects of the action. Some sensitivity calculations were performed to help understanding of the impact from actions that rely on the partially effective equipment or limited material resources. The severe accident computer code MAAP4-CANDU is used as a tool in this verification. This paper will describe the methodology used in the verification of SAMG actions and some results obtained from simulations. (author)

National report on 'stress tests', NPP Dukovany and NPP Temelin, Czech Republic. Evaluation of safety and safety margins in the light of the accident of the NPP Fukushima. Rev. 1

International Nuclear Information System (INIS)

2012-03-01

The stress tests were performed based on European Commission requirement as a response to the Fukushima-Daiichi accident. The stress tests encompassed the Dukovany and Temelin nuclear power plants and concentrated on the potential impacts of earthquakes, flooding, extreme weather conditions, loss of electrical power and loss of ultimate heat sink, and severe accident management. (P.A.)

Effect of alternative aging and accident simulations on polymer properties

International Nuclear Information System (INIS)

Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

1985-05-01

The influence of accident irradiation, steam, and chemical spray exposures on the behavior of twenty-three age-preconditioned polymer sample sets (twenty-one different materials) has been investigated. The test program varied the following conditions: (1) Accident simulations of irradiation and thermodynamic (steam and chemical spray) conditions were performed both sequentially and simultaneously. (2) Accident thermodynamic (steam and chemical spray) exposures were performed both with and without air present during the exposures. (3) Sequential accident irradiations were performed both at 28 0 C and 70 0 C. (4) Age preconditioning was performed both sequentially and simultaneously. (5) Sequential aging irradiations were performed both at 27 0 C and 70 0 C. (6) Sequential aging exposures were performed using two sequences: (1) thermal followed by irradiation and (2) irradiation followed by thermal. We report both general trends applicable to a majority of the tested materials as well as specific results for each polymer. Our data base consists of ultimate tensile properties at the completion of the accident exposure for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. We also report bend test results at completion of the accident exposures for two TEFZEL materials and permanent set after compression results for three EPR, one VAMAC, one BUNA N, one SILICONE, and one VITON material

Full scale simulations of accidents on spent-nuclear-fuel shipping systems

International Nuclear Information System (INIS)

Yoshimura, H.R.

1978-01-01

In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

Dichotic auditory-verbal memory in adults with cerebro-vascular accident

Directory of Open Access Journals (Sweden)

Samaneh Yekta

2014-01-01

Full Text Available Background and Aim: Cerebrovascular accident is a neurological disorder involves central nervous system. Studies have shown that it affects the outputs of behavioral auditory tests such as dichotic auditory verbal memory test. The purpose of this study was to compare this memory test results between patients with cerebrovascular accident and normal subjects.Methods: This cross-sectional study was conducted on 20 patients with cerebrovascular accident aged 50-70 years and 20 controls matched for age and gender in Emam Khomeini Hospital, Tehran, Iran. Dichotic auditory verbal memory test was performed on each subject.Results: The mean score in the two groups was significantly different (p<0.0001. The results indicated that the right-ear score was significantly greater than the left-ear score in normal subjects (p<0.0001 and in patients with right hemisphere lesion (p<0.0001. The right-ear and left-ear scores were not significantly different in patients with left hemisphere lesion (p=0.0860.Conclusion: Among other methods, Dichotic auditory verbal memory test is a beneficial test in assessing the central auditory nervous system of patients with cerebrovascular accident. It seems that it is sensitive to the damages occur following temporal lobe strokes.

Accident Testing of High Temperature Reactor Fuel Elements with the KueFA Device

International Nuclear Information System (INIS)

Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Ferreira-Teixeira, A.E.; Van Winckel, S.; Rondinella, V.V.; Allelein, H.J.

2013-06-01

The High Temperature Reactor (HTR) is characterised by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with Tri-Isotropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 deg. C, remaining well below the melting point of the fuel. The Cold Finger Apparatus (KueFA) is used to observe the combined effects of Depressurization and Loss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in He atmosphere for several hundred hours, mimicking accident temperatures up to 1800 deg. C and realistic temperature transients. Non-gaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analysing plate deposits by means of HPGe gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. In order to achieve a good quantification of the release, a careful calibration of the setup is mandatory. An especially tailored collimator was designed to perform plate scanning with high spatial resolution, thus yielding information about the fission product distribution on the condensation plates. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Chemical dissolution has been performed for some condensation plates in order to assess beta nuclides of interest such as 90 Sr and possibly 129 I using an Inductively Coupled Plasma - Mass Spectrometer (ICP-MS) and to cross check the HPGe gamma spectroscopy measurements

Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

International Nuclear Information System (INIS)

Vasiliev, A.D.; Stuckert, J.

2013-01-01

Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

Reconstruction of the Chernobyl emergency and accident management

International Nuclear Information System (INIS)

Schinner, F.; Andreev, I.; Andreeva, I.; Fritsche, F.; Hofer, P.; Lettner, E.; Seidelberger, E.; Kromp-Kolb, H.; Kromp, W.

1998-01-01

Full text of publication follows: on April 26, 1986 the most serious civil technological accident in the history of mankind occurred of the Chernobyl Nuclear Power Plant (ChNPP) in the former Soviet Union. As a direct result of the accident, the reactor was severely destroyed and large quantities of radionuclides were released. Some 800000 persons, also called 'liquidators' - including plant operators, fire-fighters, scientists, technicians, construction workers, emergency managers, volunteers, as well as medical and military personnel - were part of emergency measurements and accident management efforts. Activities included measures to prevent the escalation of the accident, mitigation actions, help for victims as well as activities in order to provide a basic infrastructure for this unprecedented and overwhelming task. The overall goal of the 'Project Chernobyl' of the Institute of Risk Research of the University of Vienna was to preserve for mankind the experience and knowledge of the experts among the 'liquidators' before it is lost forever. One method used to reconstruct the emergency measures of Chernobyl was the direct cooperation with liquidators. Simple questionnaires were distributed among liquidators and a database of leading accident managers, engineers, medical experts etc. was established. During an initial struggle with a number of difficulties, the response was sparse. However, after an official permit had been issued, the questionnaires delivered a wealth of data. Furthermore a documentary archive was established, which provided additional information. The multidimensional problem in connection with the severe accident of Chernobyl, the clarification of the causes of the accident, as well as failures and successes and lessons to be learned from the Chernobyl emergency measures and accident management are discussed. (authors)

Major Accidents (Gray Swans) Likelihood Modeling Using Accident Precursors and Approximate Reasoning.

Science.gov (United States)

Khakzad, Nima; Khan, Faisal; Amyotte, Paul

2015-07-01

Compared to the remarkable progress in risk analysis of normal accidents, the risk analysis of major accidents has not been so well-established, partly due to the complexity of such accidents and partly due to low probabilities involved. The issue of low probabilities normally arises from the scarcity of major accidents' relevant data since such accidents are few and far between. In this work, knowing that major accidents are frequently preceded by accident precursors, a novel precursor-based methodology has been developed for likelihood modeling of major accidents in critical infrastructures based on a unique combination of accident precursor data, information theory, and approximate reasoning. For this purpose, we have introduced an innovative application of information analysis to identify the most informative near accident of a major accident. The observed data of the near accident were then used to establish predictive scenarios to foresee the occurrence of the major accident. We verified the methodology using offshore blowouts in the Gulf of Mexico, and then demonstrated its application to dam breaches in the United Sates. © 2015 Society for Risk Analysis.

Domino effect in chemical accidents: main features and accident sequences.

Science.gov (United States)

Darbra, R M; Palacios, Adriana; Casal, Joaquim

2010-11-15

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes are external events (31%) and mechanical failure (29%). Storage areas (35%) and process plants (28%) are by far the most common settings for domino accidents. Eighty-nine per cent of the accidents involved flammable materials, the most frequent of which was LPG. The domino effect sequences were analyzed using relative probability event trees. The most frequent sequences were explosion→fire (27.6%), fire→explosion (27.5%) and fire→fire (17.8%). Copyright © 2010 Elsevier B.V. All rights reserved.

Modelization of the initiator success of the accident of Fukushima Daiichi in the NPP of ZION by the code MAAP-5

International Nuclear Information System (INIS)

Kevin Fernandez, M.; Jimenez, G.; Batteira, P.

2013-01-01

The main objective of the project is the modeling of the accident with a code new in the industry, MAAP5, and in a different nuclear plant, as well as various parameters sensitivity analysis to assess their influence on the evolution of the accident. The paper presents the analysis of the evolution of the simulated accident, as well as the evaluation of different sensitivity analyses performed on different parameters influence on the evolution: pre-accident conditions, actions of operator, etc. Operator actions, not referred to in the emergency procedures, which could influence the behavior of the reactor vessel during severe accident progression were analyzed.

Accident information needs

International Nuclear Information System (INIS)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-12-31

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

A database system for the management of severe accident risk information, SARD

International Nuclear Information System (INIS)

Ahn, K. I.; Kim, D. H.

2003-01-01

The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies

A database system for the management of severe accident risk information, SARD

Energy Technology Data Exchange (ETDEWEB)

Ahn, K. I.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

2003-10-01

The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies.

Substance use among Iranian drivers involved in fatal road accidents

Directory of Open Access Journals (Sweden)

Shervin eAssari

2014-08-01

Full Text Available Background: Although the problem of substance use among drivers is not limited to a special part of the world, most published epidemiological reports on this topic is from industrial world.Aim: To determine drug use among Iranian adults who were imprisoned for vehicle accidents with fatality. Methods: This study enrolled 51 Iranian adults who were imprisoned for vehicle accidents with fatality. This sample came from a national survey of prisoners. Data was collected at entry to prisons during the last 4 months of 2008 in 7 prisons in different parts of the country. Self reported drug use was registered. Commercial substance use screening tests were also done. Results: Drug test was positive for opioids, cannabis and both in 37.3%, 2.0% and 13.7%, respectively. 29.4% tested positive for benzodiazepines. Using test introduced 23.5% of our sample as drug users, who had declined to report any drug use. Conclusion: Opioids are the most used illicit drug in the case of vehicle accidents with fatality, however, 20% of users do not declare their use. This high rate of drug use in vehicle accidents with fatality reflects the importance of drug use control as a part of injury prevention in Iran. There might be a need for drug screening after severe car accidents.

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Energy Technology Data Exchange (ETDEWEB)

Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

2015-08-15

Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

Reducing Side-Sweep Accidents with Vehicle-to-Vehicle Communication

Directory of Open Access Journals (Sweden)

Gamini Bulumulle

2016-12-01

Full Text Available Side-sweep accidents are one of the major causes of loss of life and property damage on highways. This type of accident is caused by a driver initiating a lane change while another vehicle is blocking the road in the target lane. In this article, we are trying to quantify the degree to which different implementations of vehicle-to-vehicle (V2V communication could reduce the occurrence of such accidents. We present the design of a simulator that takes into account common sources of lack of driver awareness such as blind-spots and lack of attention. Then, we study the impact of both traditional, non-technological communication means such as turning signals as well as unidirectional and bidirectional V2V communications.

Simple estimate of fission rate during JCO criticality accident

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, Kazuhiro [Faculty of Studies on Contemporary Society, Aichi Shukutoku Univ., Nagakute, Aichi (Japan)

2000-03-01

The fission rate during JCO criticality accident is estimated from fission-product (FP) radioactivities in a uranium solution sample taken from the preparation basin 20 days after the accident. The FP radioactivity data are taken from a report by JAERI released in the Accident Investigation Committee. The total fission number is found quite dependent on the FP radioactivities and estimated to be about 4x10{sup 16} per liter, or 2x10{sup 18} per 16 kgU (assuming uranium concentration 278.9 g/liter). On the contrary, the time dependence of the fission rate is rather insensitive to the FP radioactivities. Hence, it is difficult to determine the fission number in the initial burst from the radioactivity data. (author)

Simple estimate of fission rate during JCO criticality accident

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro

2000-01-01

The fission rate during JCO criticality accident is estimated from fission-product (FP) radioactivities in a uranium solution sample taken from the preparation basin 20 days after the accident. The FP radioactivity data are taken from a report by JAERI released in the Accident Investigation Committee. The total fission number is found quite dependent on the FP radioactivities and estimated to be about 4x10 16 per liter, or 2x10 18 per 16 kgU (assuming uranium concentration 278.9 g/liter). On the contrary, the time dependence of the fission rate is rather insensitive to the FP radioactivities. Hence, it is difficult to determine the fission number in the initial burst from the radioactivity data. (author)

Learning lessons from Natech accidents - the eNATECH accident database

Science.gov (United States)

Krausmann, Elisabeth; Girgin, Serkan

2016-04-01

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

Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

International Nuclear Information System (INIS)

Girault, N.; Dickinson, S.; Funke, F.; Auvinen, A.; Herranz, L.; Krausmann, E.

2006-01-01

The International Phebus Fission Product programme, initiated in 1988 and performed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building. The results of the first two tests, called FPT0 and Ftp, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for Ftp and fuel burned in a reactor at 23 GWdt -1 for Ftp, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides. To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical

Development of Parameter Network for Accident Management Applications

Energy Technology Data Exchange (ETDEWEB)

Pak, Sukyoung; Ahemd, Rizwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jung Taek; Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

When a severe accident happens, it is hard to obtain the necessary information to understand of internal status because of the failure or damage of instrumentation and control systems. We learned the lessons from Fukushima accident that internal instrumentation system should be secured and must have ability to react in serious conditions. While there might be a number of methods to reinforce the integrity of instrumentation systems, we focused on the use of redundant behavior of plant parameters without additional hardware installation. Specifically, the objective of this study is to estimate the replaced value which is able to identify internal status by using set of available signals when it is impossible to use instrumentation information in a severe accident, which is the continuation of the paper which was submitted at the last KNS meeting. The concept of the VPN was suggested to improve the quality of parameters particularly to be logged during severe accidents in NPPs using a software based approach, and quantize the importance of each parameter for further maintenance. In the future, we will continue to perform the same analysis to other accident scenarios and extend the spectrum of initial conditions so that we are able to get more sets of VPNs and ANN models to predict the behavior of accident scenarios. The suggested method has the uncertainty underlain in the analysis code for severe accidents. However, In case of failure to the safety critical instrumentation, the information from the VPN would be available to carry out safety management operation.

Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

2015-07-01

The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

Evaluation of severe accident safety system value based on averting financial risks

International Nuclear Information System (INIS)

Hatch, S.W.; Benjamin, A.S.; Bennett, P.R.

1983-01-01

The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions

Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

Energy Technology Data Exchange (ETDEWEB)

Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

1997-07-01

The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system {open_quotes}IMPACT{close_quotes} for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT`s distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data.

Impact of Pre-Initiators on PSA in Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Ochirbat, Chimedtseren [KAIST, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor.

Impact of Pre-Initiators on PSA in Research Reactor

International Nuclear Information System (INIS)

Ochirbat, Chimedtseren; Kim, Sok Chul

2014-01-01

Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor

Statistical Analysis And Treatment Of Accident Black Spots: A Case Study Of Nandyal Mandal

Science.gov (United States)

Sudharshan Reddy, B.; Vishnu Vardhan Reddy, L.; Sreenivasa Reddy, G., Dr

2017-08-01

Background: Increased, economic activity raised the consumption levels of the people across the country. This created scope for increase in travel and transportation. The increase in the vehicles since last 10 years has put lot of pressure on the existing roads and ultimately resulting in road accidents. Nandyal Mandal is located in the Kurnool district of Andhra Pradesh and well developed in both agricultural and industrial sectors after Kurnool. 567 accidents occurred in the last seven years at 143 locations shows the severity of the accidents in the Nandyal Mandal. There is a need to carry out some work in the Nandyal Mandal to improve the accidents black spots for reducing the accidents. Methods: Last seven years (2010-2016) of accident data collected from Police Stations. Weighted Severity Index (WSI), a scientific method is used for identifying the accident black spots. Statistical analysis has carried out for the collected data using Chi-Square Test to determine the independence of accidents with other attributes. Chi-Square Goodness of fit test conducted for test whether the accidents are occurring by chance or following any pattern. Results: WSI values are determined for the 143 locations. The Locations with high WSI are treated as accident black spots. Five black spots are taken for field study. After field observations and interaction with the public, some improvements are suggested for improving the accident black spots. There is no relationship between the severity of accidents and the other attributes like month, season, day, hours in day and the age group except type of vehicle. Road accidents are distributed throughout the Year, Month and Season. Road accidents are not distributed throughout the day.

[Recreational boating accidents--Part 1: Catamnestic study].

Science.gov (United States)

Lignitz, Eberhard; Lustig, Martina; Scheibe, Ernst

2014-01-01

Deaths on the water are common in the autopsy material of medicolegal institutes situated on the coast or big rivers and lakes (illustrated by the example of the Institute of Legal Medicine of Greifswald University). They mostly occur during recreational boating activities. Apart from hydro-meteorological influences, human error is the main cause of accidents. Often it is not sufficiently kept in mind whether the boat crew is fit for sailing and proper seamanship is ensured. Drowning (following initial hypothermia) is the most frequent cause of death. Medicolegal aspects are not decisive for ordering a forensic autopsy. As statistics are not compiled in a uniform way, a comparison of the data of different institutions engaged in investigating deaths at sea and during water sports activities is hardly possible, neither on a national nor an international basis--and the reconstruction of aquatic accidents is generally difficult. Fatal accidents can only be prevented by completely clarifying their causes.

Aspects of risk analysis application to estimation of nuclear accidents and tests consequences and intervention management

International Nuclear Information System (INIS)

Demin, V.F.; Hedemann-Jensen, P.; Rolevich, I.V.; Schneider, T.S.; Sobolev, B.G.

1996-01-01

For assessment of accident consequences and a post-accident management a risk analysis methodology and data bank (BARD) with allowance for radiation and non-radiation risk causes should be developed and used. Aspects of these needs and developments are considered. Some illustrative results of health risk estimation made with BARD for the Bryansk region territory with relatively high radioactive contamination from the Chernobyl accident are presented

Phase Startup Initiative Phases 3 and 4 Test Plan and Test Specification ( OCRWM)

Energy Technology Data Exchange (ETDEWEB)

PAJUNEN, A.L.; LANGEVIN, M.J.

2000-08-07

Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: Define the Phase 3 and 4 test scope for the FRS and IWTS; Provide detailed test requirements that can be used to write the specific test procedures; Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and Define specific test objectives and acceptance criteria.

Phase Startup Initiative Phases 3 and 4 Test Plan and Test Specification (OCRWM)

International Nuclear Information System (INIS)

PAJUNEN, A.L.; LANGEVIN, M.J.

2000-01-01

Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: Define the Phase 3 and 4 test scope for the FRS and IWTS; Provide detailed test requirements that can be used to write the specific test procedures; Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and Define specific test objectives and acceptance criteria

Evaluation of strategies for severe accident prevention and mitigation

International Nuclear Information System (INIS)

Tokarz, R.

1989-01-01

The NRC is planning to establish regulatory oversight on severe accident management capability in the US nuclear reactor industry. Accident management includes certain preparatory and recovery measures that can be taken by the plant operating and technical personnel to prevent or mitigate the consequences of a severe accident. Following an initiating event, accident management strategies include measures to (1) prevent core damage, (2) arrest the core damage if it begins and retain the core inside the vessel, (3) maintain containment integrity if the vessel is breached, and (4) minimize offsite releases. Objectives of the NRC Severe Accident Management Program are to assure that technically sound strategies are identified and guidance to implement these strategies is provided to utilities. This paper will describe work performed to date by Pacific Northwest Laboratory (PNL) and Battelle Memorial Institute (BMI) relative to severe accident strategy evaluation, as well as work to be performed and expected results. Working with Brookhaven National Laboratory, PNL evaluated a series of NRC suggested accident management strategies. The evaluation of these strategies was divided between PNL and Brookhaven National Laboratory and a similar paper will be presented by Brookhaven regarding their strategy evaluation. This paper will stress the overall safety issues related to the research and emphasize the strategies that are applicable to major safety issues. The relationship of these research activities to other projects is discussed, as well as planning for future changes in the direction of work to be undertaken

RBMK-1500 accident management for loss of long-term core cooling

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.

2001-01-01

Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

Severe accidents at nuclear power plants. Their risk assessment and accident management

International Nuclear Information System (INIS)

Abe, Kiyoharu.

1995-05-01

This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

The impact of the Chernobyl accident on Norway

International Nuclear Information System (INIS)

Christensen, G.C.

1988-01-01

As the fallout from the atmospheric nuclear weapons tests gradually decreased during the 1970s, the national preparedness and analytical capacity in Norway gradually disintegrated as well. The Chernobyl accident was therefore met without any overall contingency preparedness plan. The affected governmental bodies and other institutions had to improvise their first steps, including information to the public, until necessary coordination had been established. A complicating factor was the change of government during the first days of May 1986, the reasons for this had however nothing to do with the reactor accident. A great deal of uncertainty prevailed about the accident and its consequences especially during the first days after the accident. The Ministry of Health and Social Affairs and the Ministry of the Environment in May 1986 both appointed committees to report on the accident and its impacts and on a future preparedness system, although their terms of reference were not identical. A third committee was appointed in June by the Ministry of Health and Social Affairs to report on the information crises in connection with the accident

Report of the Fukushima nuclear accident by the National Academy of Science. Lessons learned from the Fukushima nuclear accident for improving safety of U.S. nuclear plants

International Nuclear Information System (INIS)

Nariai, Hideki

2014-01-01

U.S. National Academy of Science investigated the accident at the Fukushima Daiichi nuclear plant initiated by the Great East Japan Earthquake for two years and published a draft report in July 24, 2014. Investigation results were summarized in nine new findings and made ten recommendations in a wide horizon; (1) hardware countermeasures against severe accidents and training of operators, (2) upgrade of risk assessment capability for beyond design basis accident, (3) incorporation of new information about hazards in safety regulations, (4) needed improvement of off-site emergency preparedness, and (5) improvements of nuclear safety culture. New information about hazards related with tsunami assessment, new risk assessment for beyond design basis accident, advice of foreigner resident evacuations, regulatory capture, and safety culture and regulator's specialty were discussed as Japanese issues. (T. Tanaka)

A probabilistic safety assessment of in-pile test loop in HWRR

International Nuclear Information System (INIS)

Cao Xuewu; Li Zhaohuan

1991-07-01

The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

Safety related studies on the accident behaviour of the HTR-100

International Nuclear Information System (INIS)

Wolters, J.; Mertens, J.; Altes, J.; Bongartz, R.; Breitbach, G.; David, P.H.; Degen, G.; Ehrlich, H.G.; Escherich, K.H.; Frank, E.; Hennings, W.; Jahn, W.; Koschmieder, R.; Marx, J.; Meister, G.; Moormann, R.; Rehm, W.; Verfondern, K.

1991-10-01

The aim of investigations was to verify the safety concept of the plant for balance and to quantify the radiological risk to be expected in operating an HTR-100 double unit system. Moreover, aspects of the investment risk were considered. The spectrum of initiating events ranged from so-called transients to leaks in the primary circuit and steam generator and even included earthquakes. Some of the event trees derived were highly complex and extensive due to the situation of the steam generator above the core and with regard to the double unit plant concept with increased possibilities of accident control, but also with respect to potential accident propagation. Correspondingly sophisticated analyses were required to identify risk-relevant event sequences. Environmental exposure for all risk-relevant accidents is so low that accident consequence calculations do not reveal any lethal radiation doses and practically no stochastic fatal injuries. These calculations neither assumed acute protective measures nor long-term resettlement or decontamination. The radiological risk caused by an HTR-100 plant is therefore to be classified as very low. The initiating events selected as representative and the event sequences studied in detail cover the risk-relevant event spectrum well into the hypothetical range. (orig./HP) [de

Study on air ingress during an early stage of a primary-pipe rupture accident of a high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Hishida, M.; Takeda, T.

1991-01-01

A primary-pipe rupture accident is one of the design-based accidents of the HTTR. As the first step of our final goal of predicting the multicomponent gas flow in a reactor during the early stages of the accident, the present paper aims at studying experimentally and analytically, the basic features of air ingress and gas transportation by transient molecular diffusion and the transient natural convection of a two-component gas mixture. The present paper comprises two main parts. The first part deals with analytical and experimental studies on N 2 ingress (corresponding to air ingress) and gas transportation by molecular diffusion and the one-dimensional natural convection of an He-N 2 two-component gas mixture in a reverse-U-shaped tube. Analytical and experimental results are discussed on the N 2 mole fraction change with time after the simulated pipe rupture and on the initation time of the natural circulation of pure N 2 . The second part deals with a preliminary simulation test of air ingress during the early stages of the accident. The test is performed with a very simple model of the reactor. The experimental results are discussed on the change in mole fraction of air with time and on the initiation time of the natural circulation of pure air. (orig.)

The Impact of Driver Cell Phone Use on Accidents