• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.886-892
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• 2024

The cladding temperatures and axial mass distribution computed by MAAP5 were compared with their measured values in the test bundle of the Phebus Test FPT0. The computed cladding temperatures were in good agreed with the measured values in the pre-transient phase. In the transient heat-up phase, the computed temperatures were overestimated by the Baker-Just correlation in MAAP5, but the computed temperatures could simulate the subsequently measured values. The computed mass distribution in the axial direction was in qualitative agreement with the measured one for post-test fuel damage observations. The calculated flow areas of inner and outer regions in the test bundle were compared with the photographic observations. MAAP5 computed them at the height of 0.2 m where the molten pool formed was in qualitative agreement with the photographic observations. It was found that the remaining steam flow paths might be caused by the gas-liquid two-phase flow counter-current flow limitation.

• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• Nuclear Engineering and Technology
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• v.56 no.11
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• pp.4880-4886
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• 2024

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size. The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop. This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

• Journal of Radiation Industry
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• v.18 no.3
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• pp.217-221
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• 2024

This study aims to evaluate the radiation dose received by workers within the auxiliary building of the Saeul Units 1 and 2 during a severe accident. To achieve this, representative accident scenarios were selected, and operator actions required by the severe accident management guidelines were derived to present a methodology for dose assessment. The study utilized MAAP5.06 to analyze severe accidents and employed MAAP DOSE to evaluate worker radiation exposure. Among the three operator actions considered, the direct spray action on the reactor building outer wall-side penetration resulted in the highest estimated radiation dose. This is likely because the workers are deployed near the reactor building penetration, exposing them to higher radiation levels. Future plans include the optimization of dose performance by comparing these findings with evaluations conducted using MCNP, and the development of a data-driven ALARA decision support system for predicting and diagnosing radiation exposure on nuclear sites to ensure worker safety during severe accidents.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.1
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• pp.17-24
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• 2014

We measured black carbon concentration in rural area to understand the characteristic of atmospheric aerosol by comparing the black carbon concentration and meteorological factors such as PM10 concentration, relative humidity, temperature and wind velocity. A MAAP (Multi Angle Absorption Photometer) which is one of filter based equipments was used to measure black carbon concentration. Black carbon concentration was measured to be high from April to May and low from June to September. Black carbon concentration was proportional to PM10 concentration. Black carbon concentration was correlated to relative humidity. Black carbon concentration was inversely proportional to wind velocity and temperature. Finally, we suggest that the volume fraction of black carbon in the atmosphere can be estimated from the size, number concentration and absorption coefficient measured using the MAAP.

• Particle and aerosol research
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• v.13 no.2
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• pp.97-104
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• 2017

In order to characterize atmospheric aerosols in Chungcheong area, black carbon concentration, which is known to be closely related to global warming, was measured and compared with $PM_{10}$, $PM_{2.5}$ concentrations and various meteorological parameters such as wind velocity and wind direction. Multi Angle Absorption Photometer (MAAP), a filter-based equipment, was used for the black carbon measurement, and the $PM_{10}$, $PM_{2.5}$ concentrations, wind velocity and wind direction were provided by the local monitoring stations. Black carbon concentration was monitored to be high in spring and winter but low in fall. $PM_{10}$ concentration was observed to be high when westerly wind was strong.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.547-548
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• 2004

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Journal of Korean Society for Atmospheric Environment
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• v.31 no.5
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• pp.472-481
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• 2015

Black carbon (BC) aerosols were monitored at the KIST site ($37.603^{\circ}N$, $127.046^{\circ}E$) and Cheonan-KOREATECH site ($36.766^{\circ}N$, $127.281^{\circ}E$) during the pre KORea-US Air Quality Study (KORUS-AQ) campaign using a couple of Muliti Angle Absorption Photometers (MAAP). BC mass concentrations were presented as $2.14{\pm}1.06{\mu}g/m^3$ and $0.94{\pm}0.60{\mu}g/m^3$ at KIST site (Seoul) and KOREATECH site (Cheonan), respectively. BC mass concentrations measured at KIST and KOREATECH sites from 22:00 on May 22 to 12:00 on May 23, 2015 showed 80% and 72% higher than average BC mass concentrations measured during campaign period, respectively. It indicates both sites could be influenced by a remote source. Similar patterns of BC concentrations between two sites from 20:00 to 24:00 on June 6, 2015 implies that the BC could be transported into both sites and then be stagnant inside the Korean Peninsula. Diurnal variation of BC in weekdays and weekends were also presented for the KIST and KOREATECH sites. Morning rush hour peak was observed at KIST site located in metropolitan area though no distinct morning rush hour peak was not observed at KOREATECH site located in a suburban area. This study revealed transport pathways of BC near the Korean Peninsula using back-trajectory analysis of BC measured both in a metropolitan area and in a suburban area.

• Journal of Korea Society of Industrial Information Systems
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• v.27 no.6
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• pp.95-103
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• 2022

Quick and accurate understanding of the situation in a severe accident is essential for conducting the appropriate accident management and response using the accident diagnosis information. This study employed deep learning technology to diagnose severe accidents through the major safety parameters transferred from a nuclear power plant (NPP) to AtomCARE. After selecting the major accident scenarios to consider, a learning database was established for particular scenarios affiliated with major scenarios by performing a large number of severe accident analyses using MAAP5 code. The severe accident diagnosis technology, which classifies detailed accident scenarios using the major safety parameters from NPPs, was developed by training it with the established database . Verification and validation were conducted by blind test and principal component analysis. The technology developed in this study is expected to be extended and applied to all severe accident scenarios and be utilized as a base technology for quick and accurate severe accident diagnosis.