• International Journal of Safety
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• v.10 no.2
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• pp.6-9
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• 2011

Examples and experiences of risk management on radiation under prolonged exposure situation are shown. The accident of the Fukushima dai-ichi nuclear power plant after the great east Japan earthquake (11 March, 2011) elevates background level of environmental radiation around the east Japan. For example, ambient dose equivalent rate around Tohkatsu area next to Tokyo located about 200 km-south from the plant, is about 0.1-0.6 micro-Sv $h^{-1}$ mainly due to $^{134}Cs$ and $^{137}Cs$ falling on the ground soil. This level is about double or up to ten times higher than the genuine natural level around the area. International Commission on Radiological Protection (ICRP) recommends how to face the existing exposure situation; that is the prolonged exposure situation. Referring to ICRP's reports and/or related international/domestic documents, we have been discussing how to manage this situation and acting to gain safety and relief of public, who have a possibility to be exposed to prolonged lower-dose radiation. Here, we introduce our several experiences on risk management, especially focusing on risk communication, radiation education to public, and stakeholder involvements into decision making in local governments on radiation protection, relating to the accident.

• Journal of the Korean Society of Radiology
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• v.7 no.3
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• pp.221-225
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• 2013

The purpose of this study was to examine the impact of the Fukushima nuclear power plant accident of Japan in 2011 on kindergarten teachers' perception on young children's safety due to radioactive contamination. This study analysed the 134 questionnaire forms out of 197 kindergarten teachers in the region of Chungbuk. The findings of the study indicated that the Japanese unclear power plant accident exerted an influence on anxious of kindergarten teachers on young children's safety due to radioactive contamination. In conclusion, kindergarten teacher education are required to improve the safety knowledge of kindergarten teachers and alternative resolution method for young children's safety due to radioactive contamination.

• Nuclear Engineering and Technology
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• v.19 no.4
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• pp.292-309
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• 1987

A large dry PWR containment response analysis for postulated severe accidents was performed as part of the Zion Risk Rebaselining study for input to the U.S. NRC's "Reactor Risk Reference Document," NUREG-1150. The Methodologies used in the present work were developed as part of the Severe Accident Risk Reduction Program (SARRP) at Sandia National Laboratory specifically for the Surry Plant, but they were extrapolated to Zion. Major steps of the quantification of risk from a nuclear power plant are first outlined. Then, the methodologies of containment response analysis for severe accidents used for Zion are described in detail: major features of the containment event tree (CET) analysis codes and CET quantification procedures are summarized. In addition, plant specific features important to containment response analysis are presented along with the containment loading and performance issues included in the present uncertainty analysis. Finally, a brief summary of the results of deterministic and statistical containment event tree analysis is presented to provide a perspective on the large dry PWR containment response for postulated severe accidents.accidents.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.59-65
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• 2015

Background: After the Fukushima accident, the importance of hazard analysis for extreme external events was raised. Methods: To analyze typhoon-induced hazards, which are one of the significant disasters of East Asian countries, a statistical analysis using the extreme value theory, which is a method for estimating the annual exceedance frequency of a rare event, was conducted for an estimation of the occurrence intervals or hazard levels. For the four meteorological variables, maximum wind speed, instantaneous wind speed, hourly precipitation, and daily precipitation, the parameters of the predictive extreme value theory models were estimated. Results: The 100-year return levels for each variable were predicted using the developed models and compared with previously reported values. It was also found that there exist significant long-term climate changes of wind speed and precipitation. Conclusion: A fragility analysis should be conducted to ensure the safety levels of a nuclear power plant for high levels of wind speed and precipitation, which exceed the results of a previous analysis.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.692-699
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• 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the $E_o-R_e$ plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.738-745
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• 2019

Coprecipitation using hydrous ferric oxide (HFO) has been effectively used for the removal of radionuclides from radioactive wastewater. This work studied the dynamic behavior of HFO floc formation during the neutralization of acidic ferric iron in the presence of several radionuclides by using a photometric dispersion analyzer (PDA). Then the coagulation-flocculation system using HFO-anionic poly acrylamide (PAM) composite floc system was evaluated and compared in seawater and distilled water to find the effective condition to remove the target nuclides (Co-60, Mn-54, Sb-125, and Ru-106) present in wastewater generated in the severe accident of nuclear power plant like Fukushima Daiichi case. A ferric iron dosage of 10 ppm for the formation of HFO was suitable in terms of fast formation of HFO flocs without induction time, and maximum total removal yield of radioactivity from the wastewater. The settling time of HFO flocs was reduced by changing them to HFO-PAM composite floc. The optimal dosage of anionic PAM for HFO-anionic PAM floc system was approximately 1-10 ppm. The total removal yield of Mn-54, Co-60, Sb-125, Ru-106 radionuclides by the HFO-anionic PAM coagulation-flocculation system was higher in distilled water than in seawater and was more than 99%.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.917-925
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• 2022

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (𝜃) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified 𝜃 projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.