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.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

2018 ◽  
Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin
Keyword(s):  
Water Injection ◽  
Severe Accident ◽  
Base Case ◽  
Source Terms ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Makeup Water ◽  
Accident Scenarios ◽  
The Impact

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

Risk-Informed Approach for the Regulation of Decommissioning of Nuclear Facilities

2009 ◽  
Author(s):  
Yukihiro Iguchi ◽  
Masami Kato
Keyword(s):  
Risk Assessment ◽  
High Activity ◽  
Risk Level ◽  
Spent Fuel ◽  
Nuclear Facilities ◽  
Uranium Enrichment ◽  
Periodic Safety ◽  
Distribution Of Resources ◽  
Safety Enhancement ◽  
Reprocessing Plant

During decommissioning of nuclear facilities, it is generally thought that the risk is relatively low after high activity inventory such as the spent fuel is removed. However, dismantlement works may be carried out with non-multiple barriers with a non-regular process depending mainly on human activities. Moreover, fire or gas incidents caused by conventional industry methods may lead to accidents with radioactivity release. This means more attention is necessary for safer dismantlement especially for nuclear reactors with high activity. Therefore, utilization of risk information based on risk assessment of the decommissioning was proposed. A method of risk assessment for decommissioning was developed and applied for the dismantlement of typical reactor facilities and nuclear fuel facilities (a uranium enrichment plant and a reprocessing plant). The results show that the consequences of such troubles are low, but their occurrences are still not negligible. This fact is supported by past trouble cases. Taking into account of the risk assessment results, a methodology to secure the safety of decommissioning was proposed. It consists of four steps i.e. 1) risk-informed approach, 2) graded approach, 3) phased approach and 4) layered approach and the results can be reflected to the management and regulation. The regulation means are for example, review of the decommissioning plan or the fitness-for-safety, the periodic safety inspections and usual monitoring. The methodology can evaluate the risk level of decommissioning more objectively and enable reasonable regulation based on the risk level. This leads to the appropriate distribution of resources with safety enhancement.

Risk-Informed Approach for the Regulation of Decommissioning of Nuclear Facilities

2010 ◽  
Vol 132 (10) ◽  
Author(s):  
Yukihiro Iguchi ◽  
Masami Kato
Keyword(s):  
Risk Assessment ◽  
High Activity ◽  
Risk Level ◽  
Spent Fuel ◽  
Nuclear Facilities ◽  
Uranium Enrichment ◽  
Periodic Safety ◽  
Distribution Of Resources ◽  
Safety Enhancement ◽  
Reprocessing Plant

During decommissioning of nuclear facilities, it is generally thought that the risk is relatively low after high activity inventory such as the spent fuel is removed. However, dismantlement works may be carried out with nonmultiple barriers with a nonregular process depending mainly on human activities. Moreover, fire or gas incidents caused by conventional industry methods may lead to accidents with radioactivity release. This means more attention is necessary for safer dismantlement, especially for nuclear reactors with high activity. Therefore, utilization of risk information based on risk assessment of the decommissioning was proposed. A method of risk assessment for decommissioning was developed and applied for the dismantlement of typical reactor facilities and nuclear fuel facilities (a uranium enrichment plant and a reprocessing plant). The results show that the consequences of such troubles are low but their occurrences are still not negligible. This fact is supported by past trouble cases. Taking into account the risk assessment results, a methodology to secure the safety of decommissioning was proposed. It consists of four steps, i.e., (1) risk-informed approach, (2) graded approach, (3) phased approach, and (4) layered approach and the results can be reflected to the management and regulation. The regulation means, for example, the review of decommissioning plan or the operational safety program, the periodic safety inspections and usual monitoring. The methodology can evaluate the risk level of decommissioning more objectively and enable reasonable regulation based on the risk level. This leads to the appropriate distribution of resources for safety enhancement.

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 Trends for Performance Assessment of HLW Glass and Spent Fuel as Waste Forms

1997 ◽  
Vol 506 ◽  
Author(s):  
B. Grambow
Keyword(s):  
Mass Transfer ◽  
Performance Assessment ◽  
Source Term ◽  
Reaction Rates ◽  
Dissolution Mechanism ◽  
Experimental Measurements ◽  
Source Terms ◽  
Spent Fuel ◽  
Fuel Oxidation ◽  
Waste Forms

ABSTRACTwith respect to the state of validation for source term development. Consequences of the various mechanism on mass half lives of the waste forms are calculated with analytical equations. For glass the largest uncertainty stems from the yet unclear dissolution mechanism under silica saturated conditions. Source terms based on silica solubility coupled to Si-mass transfer are probably neither conservative nor realistic. For spent fuel the largest uncertainty is in the extrapolation of radiolytic fuel oxidation for long periods of time. Considering the uncertainties involved, reaction rates cannot yet be extrapolated reliably to values much lower than the lowest reliable experimental measurements.

scholarly journals The EC MUSA Project on Management and Uncertainty of Severe Accidents: Main Pillars and Status

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4473
Author(s):  
Luis Enrique Herranz ◽  
Sara Beck ◽  
Victor Hugo Sánchez-Espinoza ◽  
Fulvio Mascari ◽  
Stephan Brumm ◽  
...  
Keyword(s):  
Source Term ◽  
Initial Boundary ◽  
Severe Accident ◽  
First Year ◽  
Model Parameters ◽  
Spent Fuel ◽  
Horizon 2020 ◽  
Severe Accidents ◽  
Current State ◽  
Accident Scenarios

In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation.

scholarly journals MELCOR Validation Study on Multi-Room Fire

2020 ◽  
Author(s):  
David L. Y. Louie ◽  
Samir El-Darazi ◽  
Lyndsey M. Fyffe ◽  
James L. Clark
Keyword(s):  
Validation Study ◽  
Source Term ◽  
Sensitivity Study ◽  
Radionuclide Transport ◽  
Aerosol Transport ◽  
Nuclear Facilities ◽  
Code Comparison ◽  
Fire Experiment ◽  
Accident Scenarios ◽  
Fire Accident

Abstract Estimation of radionuclide aerosol release to the environment, from fire accident scenarios, are one of the most dominant accident evaluations at the U.S. Department of Energy’s (DOE’s) nuclear facilities. Of particular interest to safety analysts, is estimating the radionuclide aerosol release, the Source Term (ST), based on aerosol transport from a fire room to a corridor and from the corridor to the environment. However, no existing literature has been found on estimating ST from this multi-room facility configuration. This paper contributes the following to aerosol transport modeling body of work: a validation study on a multiroom fire experiment (this includes a code-to-code comparison between MELCOR and Consolidated Fire and Smoke Transport, a specialized fire code without radionuclide transport capabilities), a sensitivity study to provide insight on the effect of smoke on ST, and a sensitivity study on the effect of aerosol entrainment in the atmosphere (puff and continuous rate) on ST.

Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 38-53
Author(s):  
M. J. Leotlela ◽  
I. Petr ◽  
A. Mathye
Keyword(s):  
Gradual Increase ◽  
Primary Objective ◽  
Fuel Pellet ◽  
Main Body ◽  
Spent Fuel ◽  
Reactivity Effect ◽  
Water Ingress ◽  
Accident Scenarios ◽  
The Impact ◽  
Gain Access

Abstract An essential component of safety analyses is the investigation of accident scenarios. In this paper water ingress scenarios of spent fuel containers, as they may occur during transport or storage, are examined. In the main body of this paper, a number of paths are studied through which water can gain access to the spent fuel cask and eventually reach the fuel pellet, potentially resulting in an increase in reactivity as a result of over-moderation. The primary objective of this project was to perform an assessment of what, in the unlikely event of a Fukushima- type accident, the impact would be on the reactivity of the cask by analyzing a gradual increase in water level in the spent fuel casks. In addition, the way the keff of the system responds to such an increase is discussed. The paper also provides the results of an assessment of the reactivity effect of water ingress via various pathways/channels.

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