Methodology for analyzing accidents with radioactive material release with code EPZDose

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 217-223
Author(s):  
J.-R. Wang ◽  
S.-S. Chen ◽  
Y. Chiang ◽  
C. Shih ◽  
J.-H. Yang ◽  
...  
Keyword(s):  
Source Term ◽  
Radioactive Material ◽  
Station Blackout ◽  
Accident Scenarios ◽  
Term Data

Abstract A methodology for analyzing accidents with radioactive material release with EPZDose code was established. This code assesses doses and it is designed and developed by NTHU (National Tsing Hua University). To confirm the capacity of EPZDose, three postulated accident scenarios Taiwanese NPPs Chinshan (BWR/4) and Maanshan (PWR) are analyzed. All these scenarios are SBO (station blackout) transients because it is assumed that they result in a release of radioactive material. In this study, the source term data for EPZDose are taken from MELCOR or RASCAL calculations. In addition, calculated results of RASCAL code are compared with the results of EPZDose for these scenarios. The comparison show that the EPZDose predictions are consistent with the data of RASCAL. This indicates that the EPZDose has a respectable accuracy in the analysis of radioactive material release accidents.

Scientists Debate Nuclear Source Terms

1985 ◽  
Vol 28 (6) ◽  
pp. 17-23
Author(s):  
John Graham
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Large Body ◽  
Radioactive Material ◽  
Source Terms ◽  
American Nuclear Society ◽  
Life Threatening ◽  
Accident Scenarios ◽  
American Physical ◽  
Term Data

The nuclear source term, defined as the quantity, timing, and characteristic of the release of radioactive material to the environment following a core-melt accident, was thoroughly debated in 1985. This debate, summarized here, turns on the Nuclear Regulatory Commission's (NRC) source term for radioactive iodine, which is postulated as potentially the most life-threatening radionuclide that might escape in a nuclear power-plant accident. A generic radioiodine source term has been used by NRC as the surrogate for all others; thus, it has become one of the bases on which nuclear-safety regulations are founded. Following the Three Mile Island (TMI) accident, from which only traces of radioiodine escaped, scientists began arguing that nuclear regulations based on source-term calculations are erroneous and should be modified. The American Nuclear Society (ANS) and industry researchers have concluded that warranted reductions in the NRC source terms could range from a factor of ten to several factors of ten in most accident scenarios. The American Physical Society (APS), after agreeing with a large body of the conclusions from the other research groups, has told NRC that its source-term data base is still inadequate because of the existence of a number of uncertainties it found therein. Although APS presented no such conclusion, its findings made clear to NRC that an early reduction of all source terms is not warranted. The anti-nuclear lobby agrees with APS. The NRC has taken a cautious, conservative approach to the revision of its regulations based on new source-term data, although it too concedes that its old methodologies and conclusions must be revised and ultimately superceded.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Thermal-Hydraulic Simulations of Guard Containment in Depressurization Accidents of a Gas-Cooled Fast Reactor

2013 ◽  
Author(s):  
Zhanjie Xu ◽  
Thomas Jordan
Keyword(s):  
Fast Reactor ◽  
Nuclear Reactor ◽  
Source Term ◽  
Reference Value ◽  
Design Stage ◽  
Primary System ◽  
Simulation Results ◽  
Accident Scenarios ◽  
Design Pressure ◽  
The Eu

A gas-cooled fast reactor is designed as an advanced nuclear reactor in next generation in the EU. In depressurization accident scenarios, pressurization caused by a release of helium from the primary system with a higher pressure into the guard containment would endanger the integrity of the containment. In the design stage, the released source term is analyzed theoretically, and is applied as a boundary condition in the 3D CFD code simulation to the transient pressurization process. The simulation results supply a reference value about the design pressure of the containment.

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

2006 ◽  
Author(s):  
Gert Sdouz
Keyword(s):  
Source Term ◽  
Management Strategies ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Management Measures ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Accident Sequences ◽  
Comprehensive Study

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the untightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the “Station Blackout”-sequence and the “Large Break LOCA”. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a “Large Break LOCA”-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a “Station Blackout”-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors.

scholarly journals Source term model of radioactive liquid spills for actual decision support systems

2021 ◽  
Vol 280 ◽  
pp. 09001
Author(s):  
Yurii Kyrylenko ◽  
Iryna Kameneva ◽  
Oleksandr Popov ◽  
Andrii Iatsyshyn ◽  
Iryna Matvieieva ◽  
...  
Keyword(s):  
Decision Support ◽  
Initial Data ◽  
Source Term ◽  
Radioactive Material ◽  
Atmospheric Models ◽  
The Public ◽  
Term Model ◽  
Radiological Impact ◽  
Emergency Scenarios ◽  
Source Term Model

Spills of liquid radioactive material are reviewed as potential event that can be associated with release into the atmosphere. Existing approaches to radiological impact assessment for onsite as well as offsite of facility are presented. The example of using the actual Java version of the European RODOS system as prototype of the decision support system shows the general implementation of the analysis and preparation of initial data in order to model the radiological impact on the public, personnel and environment. Given the specifics of the occurrence of emergency scenarios of this type, features of atmospheric models application, description of the source term model, software integration features, ventilation task solving, completeness and format of the initial data required for radiological consequence modelling.

scholarly journals Integrating four-dimensional geographical information and mobile techniques into radiological accident emergency response training

2012 ◽  
Vol 27 (1) ◽  
pp. 84-92 ◽  
Author(s):  
Ming-Kuan Tsai ◽  
Yung-Ching Lee ◽  
Chung-Hsin Lu ◽  
Mei-Hsin Chen ◽  
Tien-Yin Chou ◽  
...  
Keyword(s):  
Emergency Response ◽  
Large Scale ◽  
Training Programs ◽  
Data Representation ◽  
Radioactive Material ◽  
Geographical Information ◽  
Response Training ◽  
Effectiveness Of Training ◽  
Accident Scenarios

When radiological accidents occur, radioactive material may spread into the atmosphere, causing large-scale and long-term contamination. To diminish the effects of such accidents, researchers from many countries have investigated training programs in emergency response to radiological accidents, especially in the wake of several serious radiological accidents. Although many training programs have been proposed, this study identifies two problems: the lack of effective data representation and the lack of complete training records. Therefore, by considering various requirements for relief and evacuation work at radiological accident sites, it integrates four-dimensional geographical information and mobile techniques to construct a training platform for radiological accident emergency response. During training, groups of participants learn to respond to simulated radiological accident scenarios. Moreover, participants can use the training platform to review and discuss training details. Judging by the results, the training platform has not only increased the effectiveness of training programs, but also complied with standard operating procedures for radiological accident emergency response in Taiwan. In conclusion, this study could serve as a useful reference for similar studies and applications.

scholarly journals MELCOR Demonstration Analysis of Accident Scenarios at a Spent Nuclear Reprocessing Plant

2020 ◽  
Author(s):  
Kenneth C. Wagner ◽  
David L. Y. Louie
Keyword(s):  
Source Term ◽  
Model Development ◽  
Source Terms ◽  
Spent Fuel ◽  
Nuclear Facilities ◽  
Demonstration Analysis ◽  
Nuclear Reprocessing ◽  
Probabilistic Risk Assessments ◽  
Accident Scenarios ◽  
Reprocessing Plant

Abstract The work presented in this paper applies the MELCOR code developed at Sandia National Laboratories to evaluate the source terms from potential accidents in non-reactor nuclear facilities. The present approach provides an integrated source term approach that would be well-suited for uncertainty analysis and probabilistic risk assessments. MELCOR is used to predict the thermal-hydraulic conditions during fires or explosions that includes a release of radionuclides. The radionuclides are tracked throughout the facility from the initiating event to predict the time-dependent source term to the environment for subsequent dose or consequence evaluations. In this paper, we discuss the MELCOR input model development and the evaluation of the potential source terms from the dominated fire and explosion scenarios for a spent fuel nuclear reprocessing plant.

scholarly journals Ranking uncertainties in atmospheric dispersion modelling following the accidental release of radioactive material

2020 ◽  
Vol 55 ◽  
pp. S51-S55 ◽  
Author(s):  
S.J. Leadbetter ◽  
S. Andronopoulos ◽  
P. Bedwell ◽  
K. Chevalier-Jabet ◽  
G. Geertsema ◽  
...  
Keyword(s):  
Early Phase ◽  
Source Term ◽  
Dispersion Model ◽  
Atmospheric Dispersion ◽  
Radioactive Material ◽  
Dispersion Modelling ◽  
Dispersion Models ◽  
Sources Of Uncertainty ◽  
Accidental Release ◽  
Modelling Uncertainty

During the pre-release and early phase of an accidental release of radionuclides into the atmosphere there are few or no measurements, and dispersion models are used to assess the consequences and assist in determining appropriate countermeasures. However, uncertainties are high during this early phase and it is important to characterise these uncertainties and, if possible, include them in any dispersion modelling. In this paper we examine three sources of uncertainty in dispersion modelling; uncertainty in the source term, uncertainty in the meteorological information used to drive the dispersion model and intrinsic uncertainty within the dispersion model. We also explore the possibility of ranking these uncertainties dependent on their impact on the dispersion model outputs.

Dynamic Finite Element Analysis of Cask Handling Accidents at Storage Sites

2015 ◽  
Author(s):  
Uwe Zencker ◽  
Linan Qiao ◽  
Holger Völzke
Keyword(s):  
Finite Element ◽  
Small Variation ◽  
Limit Load ◽  
Impact Angle ◽  
Radioactive Material ◽  
Element Analysis ◽  
Dynamic Finite Element ◽  
Calculation Results ◽  
Accident Scenarios ◽  
Interim Storage

The safety assessment of casks for radioactive material at interim storage facilities or in final repositories includes the investigation of possible handling accidents if clearly defined test conditions are not available from the regulations. Specific handling accidents usually are the drop of a cask onto the transport vehicle or the floor as well as the collision with the wall of the storage building or another cask. For such load cases an experimental demonstration of cask safety would be difficult. Therefore, numerical analyses of the entire load scenario are preferred. The lessons learnt from dynamic finite element analyses of accident scenarios with thick-walled cubical containers or cylindrical casks are presented. The dependency of calculation results on initial and boundary conditions, material models, and contact conditions is discussed. Parameter sets used should be verified by numerical simulation of experimentally investigated similar test scenarios. On the other hand, decisions have to be made whether a parameter or property is modeled in a realistic or conservative manner. For example, a very small variation of the initial impact angle of a container can cause significantly different stresses and strains. In sophisticated cases an investigation of simpler limit load scenarios could be advantageous instead of analyzing a very complicated load scenario.

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