A Backward Method to Estimate the Dai-ichi Reactor Core Damage Using Radiation Exposure in the Environment

Udiyani, Pmu Pande Made; Kuntjoro, SK Sri; Widodo, SW Surip

doi:10.17146/aij.2016.497

Cited by 5 publications

(5 citation statements)

References 14 publications

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“…Filter efficiency in the cleaning system and Chemical and Volume Control System (CVCS) for iodine is 95%, other fission products are 90%, and noble gases are 0%. Filter efficiency in the reactor chimney taken for noble gases are 0%, iodine (organic) is 90%, and other nuclides (Br, Te, Cs, Rb) are 99% [4,13,25]. The normal operating source term calculation mechanism is shown in Fig.…”

Section: Methodsmentioning

confidence: 99%

“…Reduction in containment for nuclides I is 0.05%. Filter efficiency in the reactor chimney taken for noble gas is 0%, iodine (organic) is 90%; and other nuclides (Br, Te, Cs, Rb) are 99% [4,13,25]. The source term calculation mechanism in LOCA accident conditions is shown in Fig.…”

Section: Calculation Of the Characteristics Of The Pwr-mentioning

confidence: 99%

“…After Fukushima Daiichi accident, pessimistic source term was preferred to estimate radiological consequences in the site environment [16][17][18][19]. Source term is used for estimation of consequences [20][21][22][23][24], as it is more precise in determining the nuclear emergency zone as evidenced from the research of Udiyani et al [25]. Source term calculations for large nuclear power plants have been carried out because the fleets are already existing.…”

Section: Introduction *mentioning

confidence: 99%

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Source Term Assessment for 100 Mwe Pressurized Water Reactor

Udiyani

Setiawan

2020

Tri Dasa Mega

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One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

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Section: Methodsmentioning

confidence: 99%

Section: Calculation Of the Characteristics Of The Pwr-mentioning

confidence: 99%

Section: Introduction *mentioning

confidence: 99%

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Source Term Assessment for 100 Mwe Pressurized Water Reactor

Udiyani

Setiawan

2020

Tri Dasa Mega

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“…The dose calculation followed the Gaussian plume atmospheric model [5,8,[17][18][19]. The critical group selected consisted of adults.…”

Section: Dose Calculationmentioning

confidence: 99%

Atmospheric Dispersion Analysis for Expected Radiation Dose due to Normal Operation of RSG-GAS and RDE Reactors

et al. 2018

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BATAN is planning to build an experimental power reactor, the RDE, to complement the RSG-GAS reactor that is already operating in the Serpong Nuclear Zone (KNS). The experimental power reactor is an HTGR (high-temperature gas-cooled reactor) with 10 MWt power, while the RSG-GAS is a pool-type water-cooled reactor with 30 MWt power. According to standard regulatory practices, under normal operating conditions of the plant, radiological assessment of atmospheric releases to the environment and assessment of public exposures are considered essential. The purpose of this study is to estimate the dose acceptance in Serpong Nuclear Zone (KNS) after operate the RDE operates in KNS-2. To assess the doses, the PC-CREAM 08 computer code was used. It uses a standard Gaussian plume dispersion model and composes a suite of models and data for estimation of the radiological impact assessments of routine and continual discharges from a nuclear reactor. The input data include sourceterm from the RDE and the RSG-GAS, a stack the height of 60 m annual radionuclides release, meteorological data from the Serpong local meteorological station, and agricultural products data from Serpong site. Results show that the highest individual dose in the area around KNS for adults is 6.16×10-3 mSv/y in the S (South) direction and 300 m distance from the stack of RSG. The highest collective dose around KNS within 3 km radius is 6.37×10-3 man-Sv/yr. The results show that the radiological impact of the KNS on the critical groups of public and the individual effective doses satisfy the limits given by the Nuclear Regulatory Agency of Indonesia (BAPETEN). The operation of RDE in KNS-2 does not add significantly to acceptance radiation dose in the environment in KNS. It can also be concluded that the estimated effective doses are lower than the dose constraint of 0.3 mSv/y associated with this plant.

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“…For this initial stage, validation to a developed code program is done by comparing the result with manual calculation or using Excel spreadsheet, used in SimPact version 01. Furthermore, a comparative study will be carried out with various calculation results using other applicable commercial software such as PC-Cosyma implemented for research reactors as well as power reactors in various locations [3][4][5][6].…”

Section: Introduction mentioning

confidence: 99%

Preliminary Development of Radionuclides Release of Individual Dose Code Program for Radiation Monitoring Purposes

Pane

Udiyani²,

Setiawan³

et al. 2021

Tri Dasa Mega

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Environmental radiation monitoring is one of the important efforts in protecting society and the environment from radiation hazards, both natural and artificial. The presence of three nuclear research reactors and plans to build a nuclear power plant reactor prompted Indonesia to prepare a radiation monitoring system for safety and security (SPRKK). The goal of the study is to provide an appropriate method for developing radiation monitoring system to support the development of nuclear power plant in the near future. For this preliminary study, the author developed a code program using Gaussian distribution model approach for predicting radionuclide release and individual dose acceptancy by human being within 16 wind directions sectors and up to 50 km distance. The model includes estimation of source term from the nuclear installation, release of radionuclides source into air following Gaussian diffusion model, some of the release deposit to the land and entering human being through inhalation, direct external exposure, and resuspension, and predicted its accepted individual dose. This model has been widely used in various code program such as SimPact and PC-Cosyma. For this study, the model will be validated using SimPact code program. The model has been successfully developed with less than 5% deviation. Further study will be done by evaluating the model with real measuring data from research reactor installation and prepare for interfacing with real time radiation data acquisition and monitoring as part of radiation monitoring system during normal and accident condition.

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A Backward Method to Estimate the Dai-ichi Reactor Core Damage Using Radiation Exposure in the Environment

Cited by 5 publications

References 14 publications

Source Term Assessment for 100 Mwe Pressurized Water Reactor

Source Term Assessment for 100 Mwe Pressurized Water Reactor

Atmospheric Dispersion Analysis for Expected Radiation Dose due to Normal Operation of RSG-GAS and RDE Reactors

Preliminary Development of Radionuclides Release of Individual Dose Code Program for Radiation Monitoring Purposes

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