Development of Statistical Safety Analysis for Best-Estimate Code for Loss of Coolant Accident Using Relap5/Mod 3.3

2020 ◽  
Vol 2020.57 (0) ◽  
pp. D013
Author(s):  
Thanh Tung DUONG ◽  
Yuichi OTSUKA
Keyword(s):  
Safety Analysis ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

2012 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Computer Code ◽  
Safety Analysis ◽  
Fuel Rod ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Computer Codes ◽  
Flow Blockage

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.

Preliminary Safety Analysis of CSR1000

2013 ◽  
Author(s):  
Pan Wu ◽  
Junli Gou ◽  
Jianqiang Shan ◽  
Bo Zhang ◽  
Xiang Li
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Safety Analysis ◽  
Safety System ◽  
Sensitivity Analyses ◽  
Passive Safety ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Loss Of Coolant ◽  
Break Area

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.

A Preliminary Safety Analysis of the Whole Assembly Seed-Blanket Heterogeneous PWR Core

2004 ◽  
Author(s):  
Jeongik Lee ◽  
Pradip Saha ◽  
Mujid S. Kazimi ◽  
Won-Jae Lee
Keyword(s):  
Temperature Rise ◽  
Safety Analysis ◽  
Complete Loss ◽  
Pressurized Water Reactors ◽  
Primary Flow ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Safety Limit

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.

Best Estimate Analysis of Maanshan PWR LBLOCA by TRACE Coupling With DAKOTA

2014 ◽  
Author(s):  
Jung-Hua Yang ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Chunkuan Shih
Keyword(s):  
Power Plant ◽  
Input Data ◽  
Rank Correlation ◽  
Correlation Coefficients ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Probability Calculation ◽  
Partial Rank Correlation ◽  
Limiting Values

This research is focused on the Large Break Loss of Coolant Accident (LBLOCA) analysis of the Maanshan power plant by TRACE-DAKOTA code. In the acceptance criteria for Loss of Coolant Accidents (LOCAs), there are two accepted analysis methods: conservative methodology and best estimate methodology. Compared with conservative methodology, the best estimate and realistic input data with uncertainties to quantify the limiting values i.e., Peak Cladding Temperature (PCT) for LOCAs analysis. By the conservative methodology, the PCTCM (PCT calculated by conservative methodology) of Maanshan power plant LBLOCA calculated is 1422K. On the other hand, there are six key parameters taken into account in the uncertainty analysis in this study. In PCT95/95 (PCT of 95/95 confidence level and probability) calculation, the PCT95/95 is 1369K lower than the PCTCM (1422K). In addition, the partial rank correlation coefficients between input parameters and PCT indicate that accumulator pressure is the most sensitive parameter in this study.

The BEMUSE Programme: Results of the First Part Concerning the LOFT L2-5 Test

2006 ◽  
Author(s):  
Agne`s de Cre`cy ◽  
Pascal Bazin ◽  
Francesco D’Auria ◽  
Alessandro Petruzzi ◽  
Yong-Ho Ryu
Keyword(s):  
Probabilistic Approach ◽  
Sensitivity Analyses ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Sensitivity Evaluation ◽  
Loss Of Coolant ◽  
Regression Techniques ◽  
Uncertain Input ◽  
The University ◽  
Range Of Variation

This paper is aimed at describing results of the first part of the BEMUSE (Best Estimate Methods – Uncertainty and Sensitivity Evaluation) programme. The purpose of BEMUSE is the comparison of best-estimate calculations, followed by the comparison of uncertainty and sensitivity analyses for a Large Break Loss of Coolant Accident (LB-LOCA). The first part of the programme is devoted to the study of the LOFT L2-5 experiment. After a general presentation of the programme, which implies more than ten participants, this paper describes the qualification process and the results of the best-estimate calculations. The results are significantly less dispersed than those of the ISP-13, concerning already LOFT L2-5 more than 20 years ago. Then, it presents extensively the methods and the results of uncertainty and sensitivity analyses. All the participants, apart from the University of Pisa with the CIAU method, use a fully probabilistic approach, based on Wilks’ formula. However, differences appear for the choice of the uncertain input parameters to be considered and for their associated range of variation. Sensitivity analysis is performed with regression techniques, and the results are also compared. As a conclusion, main lessons learnt from BEMUSE and recommendations are presented.

Application of the UMAE Methodology for the Uncertainty Evaluation in Support to the EOP Background Analysis

2010 ◽  
Author(s):  
Tomislav Bajs ◽  
Damir Konjarek ◽  
Ilijana Ivekovic´
Keyword(s):  
Time Window ◽  
Uncertainty Evaluation ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Sensitivity Studies ◽  
Operator Action ◽  
Core Cooling ◽  
Reactor Pressure ◽  
Emergency Operating

Validation of EOPs (Emergency Operating Procedures) relies on the best-estimate analysis of the transient scenarios. In order to cover associated uncertainties, usually limited number of sensitivity studies is performed for the development of the EOPs in order to identify possible plant states and associated parameters relevant for operator actions. Recently, developed methodologies for the uncertainty evaluation made it possible to evaluate directly uncertainties with the respect to the scenarios analyzed. UMAE (Uncertainty Methodology based on Accuracy Extrapolation) uncertainty methodology has been applied for development of function restoration EOPs. More specifically, Inadequate Core Cooling (ICC) LOCA (Loss of Coolant Accident) scenario has been analyzed using best estimate transient analysis code RELAP5/SCDAPSIM code. Time window for successful operator action has been evaluated following 4.0″ cold leg break near the Reactor Pressure Vessel (RPV) in a 2-loop PWR plant.

Best-Estimate Evaluation of Large-Break Loss-of-Coolant Accident for Advanced Natural Circulation Nuclear Reactor

2008 ◽  
Vol 160 (3) ◽  
pp. 318-333 ◽  
Author(s):  
Vivek Bhasin ◽  
A. Srivastava ◽  
R. Rastogi ◽  
H. G. Lele ◽  
K. K. Vaze ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Uncertainty Analysis of SB-LOCA in AP1000 Nuclear Power Plant

2017 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen ◽  
Lizhi Jiang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Safety Margin ◽  
Analysis Model ◽  
The Core ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Best estimate plus uncertainty methods (BEPUs) could be used to eliminate the over-conservatism and gain more safety margin in the analysis of thermal-hydraulic transient process at nuclear power plant. Based on the Best estimate thermal-hydraulic system code RELAP5/MOD3.2 platform, the best estimate plus uncertainty methods (BEPUs) proposed by GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) are presented together with applications to a small break loss of coolant accident (SB-LOCA) on the AP1000 Nuclear Power Plant best estimate analysis model. According to the results of uncertainty calculations, the dispersion bands of maximum cladding temperature and the core outlet void fraction are displayed and assessed.

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