• Journal of the Korean Society of Safety
• /
• v.27 no.6
• /
• pp.192-204
• /
• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.2
• /
• pp.193-204
• /
• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• Journal of computational fluids engineering
• /
• v.19 no.1
• /
• pp.15-20
• /
• 2014

Accurate simulation of free-surface wave flows around a ship is very important for better hull-form design. In this paper, a computational fluid dynamics (CFD) code, termed SNUFOAM, which is based on the open source libraries, OpenFOAM, was developed to predict the wave patterns around a ship. Additional anti-diffusion source term for minimizing a numerical diffusion, which was caused by convection differencing scheme, was considered in the volume-fraction transport equation. The influence of the anti-diffusion source term was tested by applying it to free-surface wave flow around the Wigley model ship. In results, the band width of the volume fraction contours between 0.1 to 0.9 at the hull surface was narrowed by considering the anti-diffusion term.

• Nuclear Engineering and Technology
• /
• v.30 no.3
• /
• pp.235-244
• /
• 1998

The parametric model method determines the accident source term which is Presented by a set of source term parameters. In this method, the cumulative distribution of each source term parameter should be derived for its uncertainty analysis. This paper introduces a method of generating the parameters in the form of cumulative distribution using MAAP version 4.0. In MAAP, there are model parameters which could incorporate uncertain physical and/or chemical phenomena. In general, the model parameters do not have a point value but a range. In this paper, considering that, the input values of model parameters influencing each parameter are sampled using LHS. Then, the computation results are shown in cumulative distribution form. For a case study, the CDFs of FCOR and WES of Kori Unit 1 are derived. The target scenarios for the computation are the ones whose initial events are large LOCA, small LOCA and transient, respectively. It is found that the computed CDF's in this study are consistent to those of NUREG-1150 and the use of MAAP is proven to be adequate in assessing the parameters of the severe accident source term.

• Journal of the Korean Society of Safety
• /
• v.27 no.5
• /
• pp.219-223
• /
• 2012

For an off-site consequence analysis at nuclear power plant, MELCOR Accident Consequence Code System(MACCS) II code is widely used as a software tool. In this study, the algorithm of web-based off-site consequence analysis program(OSCAP) using the MACCS II code was developed for an Integrated Leak Rate Test (ILRT) interval extension and Level 3 probabilistic safety assessment(PSA), and verification and validation(V&V) of the program was performed. The main input data for the MACCS II code are meteorological, population distribution and source term information. However, it requires lots of time and efforts to generate the main input data for an off-site consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each site. To reduce the efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. For the source term data, the program includes the release fraction of each source term category resulting from modular accident analysis program(MAAP) code analysis and the core inventory data from ORIGEN. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so the user can select the defaulted source term data of each plant without handling source term input data.

• Nuclear Engineering and Technology
• /
• v.20 no.2
• /
• pp.120-131
• /
• 1988

A souce term model describes mathematically the source of radionuclides as they begin slow migration and decay in deep groundwater. Various source term models based on mass-transfer analysis and measurement-based source term models are reviewed. Ganerally, two processes are involved in leaching or dissolution: (1) chemical reactions and (2) mass transfer by diffusion. The chemical reaction controls the dissolution rates only during the early stage of exposure to groundwater. The exterior-field mass transfer may control the long term dissolution rates from the waste solid in a geologic repository. Mass-transfer analyses re3y on detailed and careful application of the governing equations that describe the mechanistic processes of transport of material between and within phases. If used correctly, source term models based on mass-transfer theory are valuable and necessary tools for developing reliable predictions.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4722-4730
• /
• 2022

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation - Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the depletion matrix are excluded in the corrector, and hence the corrector calculation involves only the reaction components, which can be efficiently solved with the Talyor Expansion Method (TEM). The numerical test shows that the new scheme substantially reduces computing time without loss of accuracy in comparison with the conventional scheme using CRAM (Chebyshev Rational Approximation Method), especially when the substep calculations are applied. The depletion calculation and source term estimation capability of BESNA are verified and validated through several problems, where results from BESNA are compared with those calculated by other codes as well as measured data. The analysis results show the computational efficiency of the new modified scheme and the reliability of BESNA in both isotopic predictions and source term estimations.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.40 no.7
• /
• pp.471-475
• /
• 2016

In the present study, temperature evaluations on long-term storage of radioactive waste produced in the process of isotope production were performed using two different methods. Three-dimensional analysis was carried out assuming a volumetric heat source, while two-dimensional studies were performed assuming a point source. The maximum temperature difference between the predictions of the volumetric and point source models was approximately $5^{\circ}C$. For the conceptual design level, a point source model may be suitable to obtain the overall temperature characteristics of different loading locations. For more detailed analysis, the model with the volumetric source may be applicable to optimize the loading pattern in order to obtain minimum temperatures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.13 no.1
• /
• pp.21-34
• /
• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2690-2701
• /
• 2022

In this paper, we select important fission products for the estimation of the source term during a severe accident of a PWR. The selection is based on the numerical results obtained from depletion calculations for the typical PWR fuel via the in-house code named DEGETION (Depletion, Generation, and Transmutation of Isotopes on Nuclear Application), release fractions of the fission products derived from NUREG-1465, and effective dose conversion coefficients from ICRP 119. Then, for the selected fission products, we obtain the adjoint solutions of the Bateman equations for radioactive decay in order to determine the importance of precursors producing the aforementioned fission products via radioactive decay, which would provide insights into the assumption used in MACCS 2 for a level 3 PSA analysis in which up to six precursors are considered in the calculations of radioactive decays for the fission product after release from the reactor.