scholarly journals Study on the DLOFC accident of the GEMINI+ conceptual design of HTGR reactor with MELCOR and SPECTRA

2021 ◽  
Vol 2048 (1) ◽  
pp. 012043
Author(s):  
M Skrzypek ◽  
E Skrzypek ◽  
M Stempniewicz ◽  
J Malesa
Keyword(s):  
Cooling System ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Nuclear Research ◽  
Severe Accident ◽  
Fuel Temperature ◽  
Forced Circulation ◽  
Horizon 2020 ◽  
Plant Behavior ◽  
Fast Running

Abstract The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI Plus project. Behavior of the HTGR reactor under severe accident conditions was investigated and the maximum fuel temperature was observed. Due to application of the TRISO particles and SiC layers in the fuel element, no damage of the fuel is expected up to 1600°C. Under the cooperation in the project between Nuclear Research Group (NRG) and National Centre for Nuclear Research (NCBJ) a code-to-code calculations were carried out between the SPECTRA and MELCOR codes. SPECTRA code, developed by the NRG is a thermal hydraulic analysis code and MELCOR 2.1.6342 used by NCBJ developed by SANDIA National Laboratory is fast running severe accident code. Both codes have already HTGR specific models build in. The following accident was analyzed and will be presented: Depressurized Loss of Forced Circulation (DLOFC) with 65 mm break at the top of reactor vessel. The scenario was calculated applying following sets of assumptions: best estimate and conservative. Plant behavior was analyzed including primary and secondary side of the reactor. As the results of applying conservative assumptions, it was found that fuel temperature excides the acceptable limit of 1620°C. Therefore, changes in the core design were proposed by project participants. Analyses of the new core showed acceptable temperatures. In the paper the results of code-to-code comparison are presented. Both codes have shown a good agreement of presented following characteristics on maximum fuel temperature, relative power and Reactor Cavity Cooling System power, primary pressure and break flow.

scholarly journals Evaluation Of Radionuclides Release Estimation Of Power Reactor Using Scdap / Relap

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Jupiter Sitorus Pane
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Fuel Temperature ◽  
Primary Coolant ◽  
Calculation Results ◽  
Computer Codes ◽  
Radionuclide Release ◽  
Cladding Material

<p>Incident of radiation release to the environment is important event in reactor safety analysis. Numerous studies have been conducted using various computer codes, including SCDAP/RELAP, to calculate radionuclide releases into the reactor coolant during severe accident. This paper contains description of calculation results of radionuclide release from reactor core to primary coolant system in a1000 MW PWR reactor with the aim to study behavior of radionuclide releases during severe accident. The calculations using SCDAP/RELAP was done by assuming that there has been a station black out which ends up with some vapor released into the containment. As a result, the water level in core was reduced up to a level where the core is no longer covered by water. The uncovered core heats up to certain temperature where the oxidation of the cladding started to occur.  Afterwards the oxidation generated heat made fuel melting temperature reached and as consequences the release of radionuclide to the primary coolant.  The calculations show that in parallel with the increasing of fuel temperature, the radionuclide releases into the gap through diffusion started at time of 2000seconds after initial simulation but with a neglected concentration. Subsequently at the time of 29200seconds, the temperature reached more than 1000 K and the oxidation of the Zr-cladding material occurred which accelerated the fuel temperature increase and as well as radionuclide release. At34000seconds, maximum temperature of core reached 2800 K and radionuclide release into the primary cooling system started. At this time, accumulated dissolve fission product reached amount of 74.5 kg, while the non-condensable radionuclide reached 122 kg. However, these value need to be investigated further.</p>

scholarly journals Computational Analysis of the Performance Characteristics of a Supercritical CO2 Centrifugal Compressor

Computation ◽  
2018 ◽  
Vol 6 (4) ◽  
pp. 54 ◽  
Author(s):  
Senthil Raman ◽  
Heuy Kim
Keyword(s):  
Centrifugal Compressor ◽  
Stokes Equations ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Ideal Gas ◽  
Operating Conditions ◽  
Tip Clearance ◽  
Performance Characteristics ◽  
Supercritical Regime ◽  
Vaned Diffuser

A centrifugal compressor working with supercritical CO 2 (S-CO 2 ) has several advantages over other supercritical and conventional compressors. S-CO 2 is as dense as the liquid CO 2 and becomes difficult to compress. Thus, during the operation, the S-CO 2 centrifugal compressor requires lesser compression work than the gaseous CO 2 . The performance of S-CO 2 compressors is highly varying with tip clearance and vanes in the diffuser. To improve the performance of the S-CO 2 centrifugal compressor, knowledge about the influence of individual components on the performance characteristics is necessary. This present study considers an S-CO 2 compressor designed with traditional engineering design tools based on ideal gas behaviour and tested by SANDIA national laboratory. Three-dimensional, steady, viscous flow through the S-CO 2 compressor was analysed with computational fluid dynamics solver based on the finite volume method. Navier-Stokes equations are solved with K- ω (SST) turbulence model at operating conditions in the supercritical regime. Performance of the impeller, the main component of the centrifugal compressor is compared with the impeller with vaneless diffuser and vaned diffuser configurations. The flow characteristics of the shrouded impeller are also studied to analyse the tip-leakage effect.

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 Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS injection line using MELCOR code.

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Seung Min Lee ◽  
Nelbia Da Silva Lapa ◽  
Gaianê Sabundjian
Keyword(s):  
Cooling System ◽  
Severe Accident ◽  
Time Intervals ◽  
Severe Accidents ◽  
Independent Analysis ◽  
Accident Management ◽  
Injection Line ◽  
Core Cooling ◽  
Control Volumes ◽  
Physical Phenomena

The aim of this work was to simulate a severe accident at a typical PWR, initiated with a break in Emergency Core Cooling System line of a hot leg, using the MELCOR code. The model of this typical PWR was elaborated by the Global Research for Safety and provided to the CNEN for independent analysis of the severe accidents at Angra 2, which is similar to this typical PWR. Although both of them are not identical, the results obtained of that typical PWR may be valuable because of the lack of officially published simulation of severe accident at Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes, after the break at the hot leg, were calculated as well as degree of core degradation and hydrogen production within the containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management by implementing each measure in this model.

A History of Cognitive Engineering Research at Risø (1962–1979)

1997 ◽  
Vol 41 (1) ◽  
pp. 210-214 ◽  
Author(s):  
Kim J. Vicente
Keyword(s):  
Human Error ◽  
National Laboratory ◽  
Nuclear Research ◽  
Field Studies ◽  
Cognitive Engineering ◽  
Operational Experience ◽  
Engineering Research ◽  
The Future ◽  
Power Plant Control ◽  
History Of

Following the theme for this year's conference, this paper contributes to ongoing discussions defining the future of cognitive engineering research by examining a part of its past. The history of one particular line of research, that of the Electronics Department at Risø National Laboratory, is reviewed. A number of important studies, conducted between 1962 and 1979, are briefly described. Among these are operational experience acquired from the introduction of a prototype digital console in a nuclear research reactor, two field studies of professional operators conducting representative tasks in representative settings (electronic trouble-shooting and conventional power plant control), and analyses of over 645 human error reports in the nuclear and aviation industries. Some of the themes characterizing the Risø research program in cognitive engineering are briefly summarized. These themes help define what cognitive engineering is, and what it might be concerned with in the future.

Severe Accident Analysis for BWR With Containment Outer Pool

2012 ◽  
Author(s):  
Koki Yoshimura ◽  
Kohei Hisamochi
Keyword(s):  
Cooling System ◽  
Water Injection ◽  
Heat Removal ◽  
Severe Accident ◽  
Gas Temperature ◽  
Passive System ◽  
Active Components ◽  
Active System ◽  
Station Blackout ◽  
Materials Used

Newly designed plants, e.g., next-generation light water reactor or ESBWR, employ a passive containment cooling system and have an enhanced safety with RHRs (Residual Heat Removal system) including active components. Passive containment cooling systems have the advantage of a simple mechanism, while materials used for the systems are too large to employ these systems to existing plants. Combination of passive system and active system is considered to decrease amount of material for existing plants. In this study, alternatives of applying containment outer pool as a passive system have been developed for existing BWRs, and effects of outer pool on BDBA (Beyond Design Basis Accident) have been evaluated. For the evaluation of containment outer pool, it is assumed that there would be no on-site power at the loss of off-site power event, so called “SBO (Station BlackOut)”. Then, the core of this plant would be uncovered, heated up, and damaged. Finally, the reactor pressure vessel would be breached. Containment gas temperature reached the containment failure temperature criteria without water injection. With water injection, containment pressure reached the failure pressure criteria. With this situation, using outer pool is one of the candidates to mitigate the accident. Several case studies for the outer pool have been carried out considering several parts of containment surface area, which are PCV (Pressure Containment vessel) head, W/W (Wet Well), and PCV shell. As a result of these studies, the characteristics of each containment outer pool strategies have become clear. Cooling PCV head can protect it from over-temperature, although its effect is limited and W/W venting can not be delayed. Cooling suppression pool has an effect of pressure suppressing effect when RPV is intact. Cooling PCV shell has both effect of decreasing gas temperature and suppressing pressure.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

2014 ◽  
Author(s):  
Zhang Dabin ◽  
Zhiwei Zhou ◽  
Heng Xie ◽  
Tang Yang
Keyword(s):  
Cooling System ◽  
Safety Performance ◽  
Severe Accident ◽  
Hybrid Reactor ◽  
Accident Analysis ◽  
Analysis Model ◽  
Decay Heat ◽  
Calculation Results ◽  
Core Meltdown ◽  
Dry Out

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

Performance Evaluation of DRACS System for FHTR and Time Assessment of Operation Procedure

2013 ◽  
Author(s):  
Junya Nakata ◽  
Mikihiro Wakui ◽  
Michitsugu Mori ◽  
Hiroto Sakashita ◽  
Charles Forsberg
Keyword(s):  
Performance Evaluation ◽  
High Temperature ◽  
Cooling System ◽  
Heat Removal ◽  
Severe Accident ◽  
Air Cooling ◽  
Fluoride Salt ◽  
Nuclear Power Reactor ◽  
Decay Heat ◽  
Operation Procedure

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a new concept of nuclear power reactor being investigated mainly in U.S. and China. The coolant is a liquid salt with a melting point of about 460°C and a boiling point of over 1400°C. As the baseline decay heat removal system, a passive Direct Reactor Air Cooling System (DRACS) is utilized. Though DRACS system has been developed in Sodium Fast reactors (SFR), there are some differences between both. For example, the system in FHR must decrease heat removal when temperatures are low to avoid freezing of the salt and blocking the flow of liquid. Therefore, considering its characteristics, a numerical investigation of DRACS system is needed to evaluate whether FHR has proper ability to remove decay heat and to be robust for a long-time cooling operation after even a severe accident. Furthermore, in addition to its performance evaluation, it is required to make up the operation plan of FHR considering features of this system. It is highly important, with the view of avoiding severe accident, to determine by when the system should be started up. In both countries mentioned above, Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently in progress to build. Reviewing its design and system is a crucial step needed to develop the FHR technology. In this research, a performance of DRACS system under some thermal-hydraulic basic events was evaluated by numerical simulation. This paper also suggested the adequate operation procedure suitable for FHTR to avoid a severe accident.

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