• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.858-859
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• 2005

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.745-754
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• 2012

The correlations between Large Early Release Frequency (LERF) and Early Fatality need to be investigated for risk-informed application and regulation. In Regulatory Guide (RG) -1.174, while there are decision-making criteria using the measures of Core Damage Frequency (CDF) and LERF, there are no specific criteria on LERF. Since there are both huge uncertainties and large costs needed in off-site consequence calculation, a LERF assessment methodology needs to be developed, and its correlation factor needs to be identified, for risk-informed decision-making. A new method for estimating off-site consequence has been presented and performed for assessing health effects caused by radioisotopes released from severe accidents of nuclear power plants in this study. The MACCS2 code is used for validating the source term quantitatively regarding health effects, depending on the release characteristics of radioisotopes during severe accidents. This study developed a method for identifying correlations between LERF and Early Fatality and validates the results of the model using the MACCS2 code. The results of this study may contribute to defining LERF and finding a measure for risk-informed regulations and risk-informed decision-making.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.99-104
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• 2011

An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.215-223
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• 2003

Against major release of radioactive material in nuclear power plant, Emergency Planning Zone(EPZ)s are typically established around nuclear power plants to effectively perform the public protective measures. The domestic methodology to determine the size of the EPZ is similar to that of Japan established in 1980, where calculations were based on the conservative accident source term. The objective of this study is to re-evaluate the validity of established EPZ, the area within the radius of $8{\sim}10km$ around domestic nuclear power plants, using the source terms covering full spectrum of accidents obtained from PSA study of ULJIN 3&4. To evaluate the risks of health effects, the computer code MACCS2(MELCOR Accident Consequence Code System2) was used. The result shows that the existing EPZ can reduce the probability of early fatality adequately for most of the source term categories(STCs) except for STC-14 and STC-19. In case of STC-14 and 19, the evacuation distance of 16km and 13km, respectively, are required. These distances can be reduced by improving emergency preparedness since the sensitivity studies for the public protective actions show that the magnitude of early fatality is largely affected by the time delays in notification and evacuation.

• Journal of Radiation Protection and Research
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• v.38 no.2
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• pp.60-67
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• 2013

A diffusion coefficient is an important parameter in the prediction of atmospheric dispersion using a Gaussian plume model, and its modelling approach varies. In this study, dispersion coefficients recommended by the U. S. Nuclear Regulatory Commission's (U. S. NRC's) regulatory guide and the Canadian Nuclear Safety Commission's (CNSC's) regulatory guide, and used in probabilistic accident consequence analysis codes MACCS and MACCS2 have been investigated. Based on the atmospheric dispersion model for a hypothetical accidental release recommended by the U. S. NRC, its influence to atmospheric dispersion factor was discussed. It was found that diffusion coefficients are basically predicted from a Pasquill- Gifford curve, but various curve fitting equations are recommended or used. A lateral dispersion coefficient is corrected with consideration for the additional spread due to plume meandering in all models, however its modelling approach showed a distinctive difference. Moreover, a vertical dispersion coefficient is corrected with consideration for the additional plume spread due to surface roughness in all models, except for the U. S. NRC's recommendation. For a specified surface roughness, the atmospheric dispersion factors showed differences up to approximately 4 times depending on the modelling approach of a dispersion coefficient. For the same model, the atmospheric dispersion factors showed differences by 2 to 3 times depending on surface roughness.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.244-251
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• 2015

The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1&2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0~35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2~3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.

• Journal of Radiation Protection and Research
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• v.39 no.2
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• pp.109-117
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• 2014

Nuclear emergency planning is to plan sheltering, evacuation and iodine prophylaxis for the residents living in the area where the emergency plan is needed, the area is confirmed based on the dose assessment using the source-term through an accident analysis and the data measured from meteorological tower. In this study, the does change before and after protective measures was assessed stochastically based on the one year meteorological data in the condition of the maximum hypothetical accident which can be considered at the research reactor 'HANARO', and the optimized protective measures were derived based on the reference levels defined as a residual dose by ICRP 2007 recommendation which can be applied in a emergency exposure situation. The optimized protective measures for the HANARO in the maximum hypothetical accident were the evacuation to radius 300 m, the sheltering from 300 m to 800 m, the iodine prophylaxis only for the emergency workers under the protective measures for non emergency workers.

• Journal of the Korean Society of Safety
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• v.33 no.6
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• pp.144-156
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• 2018

Level 3 Probabilistic Safety Assessment (PSA) is performed for the risk assessment that calculates radioactive material dispersion to the environment. This risk assessment is performed with a tool of MELCOR Accident Consequence Code System (MACCS2 or WinMACCS). For the off-site consequence analysis of multi-unit nuclear power plant (NPP) accident, the single location (Center Of Mass, COM) method has been usually adopted with the assumption that all the NPPs in the nuclear site are located at the same COM point. It was well known that this COM calculation can lead to underestimated or overestimated radionuclide concentration. In order to overcome this underestimation or overestimation of radionuclide concentrations in the COM method, Multiple Location (ML) method was developed in this study. The radionuclide concentrations for the individual NPPs are separately calculated, and they are summed at every location in the nuclear site by the post-processing of radionuclide concentrations that is based on two-dimensional Gaussian Plume equations. In order to demonstrate the efficiency of the ML method, radionuclide concentrations were calculated for the six-unit NPP site, radionuclide concentrations of the ML method were compared with those by COM method. This comparison was performed for conditions of constant weather, yearly weather in Korea, and four seasons, and the results were discussed. This new ML method (1) improves accuracy of radionuclide concentrations when multi-unit NPP accident occurs, (2) calculates realistic atmospheric dispersion of radionuclides under various weather conditions, and finally (3) supports off-site emergency plan optimization. It is recommended that this new method be applied to the risk assessment of multi-unit NPP accident. This new method drastically improves the accuracy of radionuclide concentrations at the locations adjacent to or very close to NPPs. This ML method has a great strength over the COM method when people live near nuclear site, since it provides accurate radionuclide concentrations or radiation doses.

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

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.187-196
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• 2004

The health effects resulting from severe accidents of typical 1,000MWe KSNP(Korea Standard Nuclear Plant) PWR and typical 600MWe CANDU(CANada Deuterium Uranium) plants were estimated and compared. The population distribution of the site extending to 80km for both site were considered. The releaese fraction for various source term categories(STC) and core inventories were used in the estimation of the health effects risks by using the MACCS2(MELCOR Accident Consequence Code System2) code. Individuals are assumed to evacuate beyond 16km from the site. The health effects considered in this comparative study are early and cancer fatality risk, and the results are presented as CCDF(Complementary Cumulative Distribution Function) curves considering the occurrence probability of each STC's. According to the results, the early and cancer fatality risks of PHWR plants we lower than those of PWR plants. This is attributed the fact that the amount of radioactive mateials that released to the atmosphere resulting from the postulated severe accidents of PHWR plants are smaller than that of PWR plants. And, the dominating initiating event of STC that shows maximum early and cancer fatality risk is SGTR(Steam Generator Tube Rupture) for both plants. Therefore, the appropriated actions must be taken to reduce the occurrence probability and the amounts of radioactive materials released to the environment in order to protect the public for both PWR and PHWR plants.