The Development of the Advanced Method for the Source Term Evaluation Applicable to the Dynamic PRA

2018 ◽  
Author(s):  
Koichi Nakamura ◽  
Sunghyon Jang ◽  
Akira Yamaguchi
Keyword(s):  
Power Plants ◽  
Evaluation Method ◽  
Source Term ◽  
Fission Products ◽  
Accident Risk ◽  
Branch Points ◽  
Event Tree ◽  
Term Evaluation ◽  
Accident Scenarios ◽  
Level 2

Useful insights on nuclear safety are provided by the level 2 probabilistic risk assessment (PRA), which evaluates the risk of fission products (FP) released in the environment during an accident in nuclear power plants (NPPs). The containment event tree method is generally employed for level 2 PRA. In this method, the accident scenarios are expressed by the combination of a number of branch points. The possible accident scenarios are approximately or representatively dealt with in this method. Dynamic PRA can evaluate the accident risk with complex changes or composition of various accident events dynamically. A reasonable source term evaluation method, which can be conducted with a small computational load, is developed for the establishment of dynamic PRA method focused on the risk of the release of FP in the environment. The proposed study aims to develop a reasonable source term evaluation method by applying phenomenological relationship diagram (PRD) method applicable to the dynamic PRA is developed.

scholarly journals SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

2020 ◽  
Vol 22 (2) ◽  
pp. 61
Author(s):  
Pande Made Udiyani ◽  
Muhammad Budi Setiawan
Keyword(s):  
Power Plants ◽  
Public Perception ◽  
Source Term ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Starting Point

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

scholarly journals Development of Source Term Evaluation Method Applicable to Dynamic PRA

2020 ◽  
Vol 19 (1) ◽  
pp. 34-46
Author(s):  
Koichi NAKAMURA ◽  
Sunghyon JANG ◽  
Akira YAMAGUCHI
Keyword(s):  
Evaluation Method ◽  
Source Term ◽  
Term Evaluation

Evaluation of Iodine Source Term in PWR under Loss of Coolant Accident

2012 ◽  
Vol 482-484 ◽  
pp. 1115-1119 ◽  
Author(s):  
Khurram Mehboob ◽  
Xin Rong Cao
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Fission Products ◽  
Retention Factor ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Term Evaluation ◽  
Computer Based ◽  
Instantaneous Release

During the severe accident in nuclear power plant (NPP), large amounts of fission products are released with accident progression, including In-vessel and Ex-vessel release. Thus, the Source term evaluation is essential for the probability risk assessment (PRA) and is still imperative for the licensing and operation of NPPs. Iodine is one of the most reactive fission products emitting in a large amount to containment and have a severe impact on health and sounding environment. Therefore, the iodine source term has been evaluated for 1000MW Reactor, by considering the TMI-2 as the reference reactor. The modeling and simulation of released radioactivity have been carried out by developing a MATLAB computer-based program. For post 1100 operation days, with the instantaneous release of radioactivity to the containment has been studied under LOCA. The dependency of radioiodine on ventilation exhaust rates has been studied in normal, emergency and isolation mode of containment. Moreover, the containment retention factor is also evaluated in said states of containment.

Development of the Methodology for Application of Revised Source Term to Operating Nuclear Power Plants in Korea

2004 ◽  
Author(s):  
Moon Soo Park ◽  
Chang-Sun Kang ◽  
Joo-Hyun Moon
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Source Term ◽  
Current Trend ◽  
Fission Products ◽  
Nuclear Plants ◽  
Limited Application ◽  
Technical Basis ◽  
Release Timing

Considering the current trend in applying the revised source term proposed by NUREG-1465 to the nuclear power plants in the U. S., it is expected that the revised source term will be applied to the Korean operating nuclear power plants in the near future, even though the exact time can not be estimated. To meet the future technical demands, it is necessary to prepare the technical system including the related regulatory requirements in advance. In this research, therefore, it is intended to develop the methodology to apply the revised source term to operating nuclear power plants in Korea. Several principles were established to develop the application methodologies. First, it is not necessary to modify the existing regulations about source term (i.e., any back-fitting to operating nuclear plants is not necessary). Second, if the pertinent margin of safety is guaranteed, the revised source term suggested by NUREG-1465 may be useful to full application. Finally, a part of revised source term could be selected to application based on the technical feasibility. As the results of this research, several methodologies to apply the revised source term to the Korean operating nuclear power plants have been developed, which include 1) the selective (or limited) application to use only some of all the characteristics of the revised source term, such as release timing of fission products and chemical form of radio-iodine and 2) the full application to use all the characteristics of the revised source term. The developed methodologies are actually applied to Ulchin 9&4 units and their application feasibilities are reviewed. The results of this research are used as either a manual in establishing the plan and the procedure for applying the revised source term to the domestic nuclear plant from the utility’s viewpoint; or a technical basis of revising the related regulations from the regulatory body’s viewpoint.

Stand-Alone Containment Analysis of the Phébus Fission Products Test 1 With the ASTEC and the MELCOR Codes

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Source Term ◽  
State Of The Art ◽  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Primary Coolant ◽  
Term Evaluation ◽  
Control Volumes ◽  
Coolant System

The estimation of fission products (FPs) release from the containment system of a nuclear plant to the external environment during a severe accident (SA) is a quite complex task. In the last 30–40 yr, several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state of the art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, also a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus fission products test 1 (FPT-1) test employing the accident source term evaluation code (ASTEC) and MELCOR codes (respectively, ASTEC v.2.0 revision 3 patch 3 and MELCOR V2.1.6840 version). The analysis focuses on the stand-alone containment aspects of the test, and three different modelizations of the containment vessel have been developed showing that at least 15/20 control volumes (CVs) are necessary for the spatial schematization to correctly predict the test thermal hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Development of Ultimate Strength Evaluation Method for Piping Subjected to Seismic Loads

2013 ◽  
Author(s):  
Hideo Machida ◽  
Hiromasa Chitose ◽  
Tatsuhiro Yamazaki
Keyword(s):  
Power Plants ◽  
Evaluation Method ◽  
Failure Modes ◽  
Limit State ◽  
Seismic Loading ◽  
Earthquake Resistance ◽  
State Function ◽  
Limit State Function ◽  
Input Acceleration ◽  
Resistance Tests

This paper reports the results of the study on the failure modes and limit loads of piping in nuclear power plants subjected to cyclic seismic loading. By investigating the past fracture tests and earthquake resistance tests, it became clear that dominant failure mode of piping was fatigue, and the effect of ratchet strain was negligible. Until now, the stress generated with the acceleration of an earthquake was classified into the primary stress. However, the relationship between the input acceleration and the seismic response displacement of the pipe observed from earthquake resistance tests is non-linear, and increasing rate of displacement is lower than that of input acceleration in elastic-plastic stress condition. Therefore, the seismic loading can be treated as displacement controlled loading. To evaluate the reliability-based critical acceleration, a limit state function was defined taking the variations in the fatigue strength or some parameters into consideration. By using the limit state function, the reliability was evaluated for the typical piping of boiling water reactor (BWR) plants subjected to cyclic seismic loading, and a partial safety factors were calculated. Based on these results, a fatigue curve corresponding to the target reliability was proposed.

Evaluation Method for Seismic Fatigue Damage of Plant Pipeline

2019 ◽  
Vol 141 (2) ◽  
Author(s):  
Fumio Inada ◽  
Michiya Sakai ◽  
Ryo Morita ◽  
Ichiro Tamura ◽  
Shin-ichi Matsuura ◽  
...  
Keyword(s):  
Fatigue Damage ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Response Spectrum ◽  
Damage Mechanism ◽  
Peak Frequency ◽  
Cumulative Damage ◽  
Cumulative Absolute Velocity ◽  
Seismic Index

Although acceleration and cumulative absolute velocity (CAV) are used as seismic indexes, their relationship with the damage mechanism is not yet understood. In this paper, a simplified evaluation method for seismic fatigue damage, which can be used as a seismic index for screening, is derived from the stress amplitude obtained from CAV for one cycle in accordance with the velocity criterion in ASME Operation and Maintenance of Nuclear Power Plants 2012, and the linear cumulative damage due to fatigue can be obtained from the linear cumulative damage rule. To verify the performance of the method, the vibration response of a cantilever pipe is calculated for four earthquake waves, and the cumulative fatigue damage is evaluated using the rain flow method. The result is in good agreement with the value obtained by the method based on the relative response. When the response spectrum obtained by the evaluation method is considered, the value obtained by the evaluation method has a peak at the peak frequency of the ground motion, and the value decreases with increasing natural frequency above the peak frequency. A higher peak frequency of the base leads to a higher value obtained by the evaluation method.

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