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

Transient Analysis for the Case of Changes in Coolant Inventory for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
H. G. Lele ◽  
A. Srivastava ◽  
B. Chatterjee ◽  
A. J. Gaikwad ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Two Phase ◽  
Decay Heat ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Core Cooling

Safety of nuclear reactor needs to be assessed against different categories of Postulated initiating events. Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Inventory of the system is important parameter in determination of flow characteristics of this natural circulation reactor. In view of this, various events that cause changes in PHT system inventory are analysed in this paper. One of the reason for decrease in coolant inventory is hypothetical Loss of coolant accident (LOCA) This event is of very low probability but important from designing engineered safeguard system of a reactor. Loss of coolant accident in a nuclear reactor can cause voiding of the reactor core due to expulsion of primary coolant from break. In such, a situation the reactor core experiences very low heat removal rate from the nuclear fuel though the decay heat generation continues even after tripping of the reactor. Heat generation in the reactor core is due to various sources such as decay heat, stored heat etc, can lead to heating of fuel elements. However, Emergency core cooling systems of the reactor are actuated and prevent undesirable temperature rise. These events are called design basis events and focus is on adequacy of Emergency Core Cooling System (ECCS) and fuel integrity. The scenarios, phenomena encountered and consequences depend upon size and location of break, system characteristics, and actuation and capability of different protection and engineered safeguard systems of the reactor system. Moreover, this reactor has several passive features to ensure safety of this reactor. which are considered in analyzing these events. Events under category of decrease in coolant inventory includes loss of coolant accidents due to break at different locations of different sizes. Various locations considered in this paper are steam line, inlet header, inlet feeder, ECCS header, downcomer, pressure tube, Isolation condenser inlet header, instrument line break at inlet header and steam drum. The paper also considers scenario emerging due to malfunctions like relief valve stuck open. Causes for events under category of increase in coolant inventory are Increase in Drum level controller set point, Inadvertent valving in of Accumulators and Inadvertent valving in of Gravity driven water pool (GDWP). Last two events are not analysed as they are not possible. The analysis for the above events is complex due to various complex and wide ranges of phenomena involved during different pies under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, two-phase natural circulation under depleted inventory conditions. Coupled neutronics and thermal hydraulics behaviour, Phenomena under LOCA, phenomena during ECCS injection, direct injection into fuel rod, advanced accumulator injection., vapour pull through and coupled controller and thermal hydraulics. Modelling of these phenomena for each event is discussed in this paper. In this paper summary of analyses for representive event is presented.

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

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

scholarly journals Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
A. Del Nevo ◽  
M. Adorni ◽  
F. D'Auria ◽  
O. I. Melikhov ◽  
I. V. Elkin ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Computer Codes ◽  
Set Up ◽  
Advanced Computer

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

scholarly journals Small break loss of coolant accident of 200 cm² in cold leg of primary loop of Angra2 nuclear power reactor evaluation

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
EDUARDO MADEIRA BORGES ◽  
GAIANÊ SABUNDJIAN
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Power Reactor ◽  
Reactor Core ◽  
Flow Regimes ◽  
Nuclear Power Reactor ◽  
Primary Loop ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Mukesh Kumar ◽  
A. K. Nayak ◽  
Sumit V. Prasad ◽  
P. K. Verma ◽  
R. K. Singh ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Low Flow ◽  
Experimental Demonstration ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Channel Conditions ◽  
Channel Type

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

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.

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