• Nuclear Engineering and Technology
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• 제34권6호
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• pp.596-609
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• 2002

Korea Electric Power Research Institute has developed the in-house safety analysis methodologies for non-LOCA(Loss Of Coolant Accident) events based on codes and methodologies of vendors and Electric Power Research Institute . According to the new methodologies, analyses of system responses and calculation of DNBR(Departure from Nucleate Boiling Ratio) during the transient have been carried out with RETRAN code and a sub-channel analysis code, respectively. However, it takes too much time to calculate DNBR for each case using the two codes to search for the limiting case from sensitivity study. To simplify the search for the limiting case, accordingly, RETRAN code has been modified to roughly calculate DNBR using hot channel modeling. The W-3 correlation is already included in RETRAN as one of the auxiliary DNBR models. However, WRB-1 and WRB-2 correlations required to analyze some Westinghouse type fuels are not considered in RETRAN DNBR models. In this paper, the RETRAN DNBR models using the correlations have been developed and the partial and complete loss of forced reactor coolant flow events have been analyzed for Yonggwang units 1 and 2 with the new methodologies to validate the models. The results of the analyses have been compared with those mentioned in the chapter 15 of the Final Safety Analysis Report.

• Nuclear Engineering and Technology
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• 제40권3호
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2003년도 춘계 학술발표회 논문집
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• pp.371-376
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• 2003

원자력 발전소에서 격납건물 계통의 건전성 유지는 냉각재상실사고(Loss of Coolant Accident: LOCA) 및 주증기관 파단(Main Steam Line Break : MSLB) 사고와 같은 설계기준사고 시 격납건물의 최대 온도/압력을 평가하는 격납건물 성능 평가는 격납용기 내에 방사능 물질을 효율적으로 가두어 방사능 피해로부터 공공의 안전을 확보할 수 있느냐 하는 관건이다.(중략)

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4014-4021
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• 2021

If safety injection systems (SISs) do not work in the event of a loss-of-coolant accident (LOCA), the accident can progress to a severe accident in which the reactor core is exposed and the reactor vessel fails. Therefore, it is considered that a technology that provides recoverable maximum time for SIS actuation is necessary to prevent this progression. In this study, the corresponding time was defined as the golden time. To achieve the objective of accurately predicting the golden time, the prediction was performed using the deep fuzzy neural network (DFNN) with rule-dropout. The DFNN with rule-dropout has an architecture in which many of the fuzzy neural networks (FNNs) are connected and is a method in which the fuzzy rule numbers, which are directly related to the number of nodes in the FNN that affect inference performance, are properly adjusted by a genetic algorithm. The golden time prediction performance of the DFNN model with rule-dropout was better than that of the support vector regression model. By using the prediction result through the proposed DFNN with rule-dropout, it is expected to prevent the aggravation of the accidents by providing the maximum remaining time for SIS recovery, which failed in the LOCA situation.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2099-2112
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• 2024

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.193-202
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• 2010

In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2565-2571
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• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.