Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Ekariansyah, Ase Andy Sofrani; Widodo, SW Surip

doi:10.17146/aij.2016.511

Cited by 9 publications

(2 citation statements)

References 15 publications

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“…Also, the highest peak of pressure is similar to the Design Control Document (DCD) result for long-term analysis. Meanwhile, Ekariansyah and Widodo [13] evaluated the PXS performance to mitigate Direct vessel injection line break (DVI-LB) using RELAP5/SCDAP/Mod 3.4. It is confirmed that the failure of DVI lines can be anticipated using injected coolant, which is operated based on a passive system so that the reactor core is always safe.…”

Section: Esfta Verificationmentioning

confidence: 99%

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

SONY TJAHYANİ,

PURBA,

EKARİANSYAH

et al. 2023

Gazi University Journal of Science

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Probabilistic safety assessment has been widely used to evaluate nuclear power plant safety systems. It couples fault tree (FT) analysis and event tree (ET) analysis. FT analysis is to develop failure scenarios, to quantify the failure probability and to identify critical components. ET analysis is to develop event sequences (ESs) leading to reactor core damage and quantify core damage frequency. Currently, there is no approach in probabilistic safety assessment that can be used to identify minimal combinations of safety system failures leading to reactor core damage. This study proposes an event sequence based fault tree analysis, which integrates FT model with ET model. The ET model is to identify the ESs leading to the reactor core damage. Meanwhile, the FT model is to identify minimal combinations of those ESs found in the ET model. The motivation of this study is on how to identify critical safety systems in nuclear power plants to mitigate disturbances caused by a group of postulated initiating events to avoid core damage. To confirm the feasibility and the applicability of the proposed approach, the performance of the AP1000 passive core cooling system is evaluated. It is found that if in-containment refueling water storage tank, automatic depressurization system - full, accumulator, core make-up tank, and passive residual heat removal work properly, the reactor core will intact. These results confirmed that the proposed approach can be applicable to identify minimal combinations of safety systems to keep the reactor core intact.

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Section: Esfta Verificationmentioning

confidence: 99%

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

SONY TJAHYANİ,

PURBA,

EKARİANSYAH

et al. 2023

Gazi University Journal of Science

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“…Until now, any safety analysis conducted has been performed using deterministic best estimate approach only when the model nodalization was developed as close as possible with the real nuclear power plant and the results to evaluate the safety criteria also were compared with the best estimate simulation using different code. Such best estimate analyses can be found in several scientific papers related to the AP1000 nuclear power plant [16,17] without considering uncertainty arising during modelling and simulation. The ATLAS test facility is selected due to the availability of the experimental results and related analysis is needed to validate the best estimate simulation.…”

Section: Introductionmentioning

confidence: 99%

Preliminary Study on Relap5 Simulation of Dvi Line Break Accident in the Atlas Facility Using Best Estimate Plus Uncertainty Method

Ekariansyah

Widodo

2017

Tri Dasa Mega

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PRELIMINARY STUDY ON RELAP5 SIMULATION OF DVI LINE BREAK ACCIDENT IN THE ATLAS FACILITY USING BEST ESTIMATE PLUS UNCERTAINTY METHOD. The BestEstimate plus Uncertainty (BEPU) is a methodology, which was introduced in the deterministic safety analysis to evaluate limitations of codes in simulating realistic plant behavior by providing quantified uncertainty bands of calculation results. It has been already widely accepted in licensing nuclear power plant by regulatory bodies of United States (USNRC), Argentina, and Canada. The uncertainty evaluation in the BEPU method is performed by different approaches such as GRS, IRSN, ENUSA, AEAT, and UNIPI. Due to the complexity of other approaches, the purpose of this study is to present some key aspects of the BEPU process using the GRS methodology by selecting the ATLAS test facility to simulate 50% break of DVI line since any safety analysis performed so far was using deterministic best estimate approach only. As comparison of the best estimate simulation performed by RELAP5/SCDAP/Mod3.4, experimental data related to the event was used. After 100 simulations, the uncertainty bands of peak heater of clad temperature and primary pressure transient obtained were only in a close agreement with the experimental data in the earlier period and less than 250 seconds during the transient condition. Therefore the overall accuracy of the best estimate simulation plays a key role on the final results of the uncertainty analysis because the propagation of any discrepancy in the best estimate with the experimental data will occur throughout the simulation. After that, selecting the important parameters to be randomly generated needs to be performed carefully by studying the important phenomena related to the event analyzed and associated plant model. Keywords: best estimate plus uncertainty, DVI line break, ATLAS facility, RELAP5, simulation ABSTRAK STUDI AWAL SIMULASI KECELAKAAN PUTUSNYA JALUR DVI PADA FASILITAS ATLAS MENGGUNAKAN RELAP5 DENGAN METODE ESTIMASI TERBAIK DAN KETIDAKPASTIAN. Metode Best estimate plus uncertainty (BEPU) adalah metode analisis keselamatan deterministik yang bertujuan untuk melakukan evaluasi keterbatasan program perhitungan dalam mensimulasikan sifat-sifat fisis instalasi secara realistik dengan mengkuantifikasi rentang ketidakpastian dari hasil perhitungan. Metode tersebut telah diterima secara luas dalam perijinan PLTN oleh badan pengatur dunia seperti di

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Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

Włostowski,

Darnowski

2024

Annals of Nuclear Energy

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Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Cited by 9 publications

References 15 publications

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

Preliminary Study on Relap5 Simulation of Dvi Line Break Accident in the Atlas Facility Using Best Estimate Plus Uncertainty Method

Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

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