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>

Analysis of Accident Progression of Fukushima Daiichi NPPs With SAMPSON Code

2013 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Severe Accident ◽  
Injection System ◽  
Part Load ◽  
Calculation Results ◽  
Fukushima Daiichi ◽  
Measured Pressure

The Fukushima Daiichi Nuclear Power Plant units 1, 2, and 3 had serious damages due to the huge earthquake and tsunami which occurred on March 11th 2011. Pressure transients in the reactor pressure vessels (RPVs) of the units 1, 2, and 3 were analyzed with the severe accident analysis code, SAMPSON for a few days from the scram until occurrence of depressurization. Since preliminary analysis results with the original SAMPSON showed difference from the measured data, the following phenomena were newly considered in the current analyses. For unit 1: Damage of a source range monitor, which is one of in-core monitors. For unit 2: Part load operation of the reactor core isolation cooling system. For unit 3: Part load operation of the high pressure coolant injection system. The calculation results showed fairly good agreements with the measured pressure data and showed RPV bottom damage for all the units resulting in falling of debris in the core region into the pedestal of the drywell.

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.

Best Estimate Analysis of PHEBUS FPT1 Test Using RELAP/SCDAPSIM Code

2012 ◽  
Author(s):  
Tadas Kaliatka ◽  
Eugenijus Ušpuras ◽  
Virginijus Vileiniškis
Keyword(s):  
Reactor Core ◽  
Fission Products ◽  
Severe Accident ◽  
Cooperative Research ◽  
Best Estimate ◽  
Fuel Bundle ◽  
Calculation Results ◽  
Estimate Method ◽  
Analysis Computer ◽  
Accident Conditions

The PHEBUS-FP program is an outstanding example of an international cooperative research program that is yielding valuable data for validating severe accident analysis computer codes. The main objective of the PHEBUS FPT1 experiment was to study the processes in the overheated reactor core, release of fission products and their subsequent transport and deposition under conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test could be divided in the bundle degradation, aerosol, washing and chemistry phases. The objective of this article is the best estimate analysis of the bundle degradation phase. GRS (Germany) best estimate method with the statistic tool SUSA used for uncertainty and sensitivity analysis of calculation results and RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during severe accident conditions, was used for the simulation of this test. The RELAP/SCDAPSIM calculation results were compared with the experimental measurements and calculations results, received by employing ICARE module of ASTEC V2 code. The performed analysis demonstrated, that the best estimate method, employing RELAP/SCDAPSIM and SUSA codes, is capable to model main severe accidents phenomena in the fuel bundle during the overheating and melting of reactor core.

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.

The Development of the Evolutionary BWR

2006 ◽  
Author(s):  
A. Murase ◽  
M. Nakamaru ◽  
M. Kuroki ◽  
Y. Kojima ◽  
S. Yokoyama
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Safety Performance ◽  
Severe Accident ◽  
Main Role ◽  
Isolation System ◽  
Near Term ◽  
Core Cooling ◽  
Control Rods

Considering the delay of the fast breeding reactor (FBR) development, it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accident (SA). At the same time, we have made the simplification of the overall plant systems in order to improve economy. The simplification of the AB1600 is based on the proven technologies. To retain the safety performance superior or equivalent to the current ABWR and to strengthen the countermeasure against SA, the AB1600 has introduced the passive systems such as the passive containment cooling system (PCCS), the gravity driven core cooling system (GDCS) and the isolation condenser (IC). While we retain the safety performance superior or equivalent to the current ABWR, we have made the simplification of the safety systems. We could eliminate the high pressure core flooder system (HPCF) and the reactor core isolation system (RCIC) by extending the height of reactor pressure vessel (RPV) two meters. To achieve simplification of reactor systems, we have reduced the number of fuel bundles and the number of control rods by adopting large bundle that has a bundle pitch 1.2 times wider than that of the current ABWR. In the 1600MWe class, the number of fuel bundles could be reduced to 600 from 872 of the current ABWR, and the number of control rods could be reduced to 137 from 205 of the current ABWR. Because the reactor internal pump (RIP) of the current ABWR has sufficient performance capacity and the improvement of fuel characteristics from the current fuel enables the operation at lower core flow, the number of RIPs could be decreased from ten to eight. Furthermore, we have reduced the number of divisions of emergency core cooling system (ECCS)/heat removal system to two from three of the current ABWR. This configuration change contributes to reduce the amount of resources of not only reactor systems but also auxiliary systems. In the previous paper, the AB1600 had four low pressure flooder systems (LPFLs). We have studied about the possibility of reduction of LPFLs to two from four by providing the LPFL with alternative injection lines. This change is expected to contribute to reduce the total number of ECCS pumps and the capacity of emergency AC power.

Variant Porosity Pebble Bed Reactor Core Thermal Hydraulic Simulation

2010 ◽  
Author(s):  
Gang Zhao ◽  
Ping Ye ◽  
Toru Obala
Keyword(s):  
Heat Transfer ◽  
Cooling System ◽  
Reactor Core ◽  
Maximum Temperature ◽  
Porous Media Flow ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Passive Cooling System ◽  
Porosity Model ◽  
Random Porosity

Spherical fuel elements are distributed randomly in the pebble bed reactor core and helium flow through the pebble bed to remove nuclear reaction heat. Pebble bed reactor core is usually treated as a uniform porous media flow in thermal hydraulic research. However the porosity distribution is nonuniform and the porosity near the wall increase sharply. A new random model is developed in this paper to investigate thermal hydraulic characteristics of pebble bed reactor core. Porosity assumption is based on porosity measurement of other research. Porosity simulation is divided into three parts according to the distance from wall. In the center of core, porosity is assumed to obey normal distribution, where average porosity is from the experimental relation based on statistical results. The mean and standard deviation of porosity distribution near the wall will increase because of the wall effect, where the distance from the wall is about three times of fuel ball’s diameter. The third part is zone from three times to five times of ball’s diameter departed from the wall. The wall effect of this zone is between center and the wall zone. Based on above assumption, a random porosity simulation is completed to apply in this research. COMSOL Multiphysics 3.5a software is used in this research. COMSOL Multiphysics are a calculation platform using proven Finite Elements Methods (FEM). In this research, Brinkman equation for porous media flow is applied in the simulation. Non-thermal Balance model is used in local heat transfer research between gas and pebble bed. A geometry model is built to simulate HTR-10. Temperature profile of variant porosity is gained from stationary analysis and comparison with uniform porosity is also discussed in the paper. For transient analysis, four cases simulation is carried out in the research. Case 1 and 2 simulate heat transfer phenomena with forced cooling system and with passive cooling system after reactor shut down. Way-Wigner-curve is used in Case 1 and Case 2 to simulate decay heat in the calculation. Case 3 and Case 4 simulate ATWS phenomena with natural convection and without natural convection system when blower is trip off in normal operation. Simulation results also are compared with some ATWS experiments and some discussion is done in the paper. From the results, it can be seen that random porosity will affect temperature distribution near the wall and make outlet temperature non-uniform greatly. The maximum temperature of variant porosity is much greater than the maximum temperature of uniform porosity at the same condition. Transient analyses of variant model show that passive cooling system can remove residual heat even in accident conditions when the blower trip off whether reactor shut down or not and the analyses results correspond substantially with experimental results. In general, variant porosity should be considered in the thermal hydraulic research of pebble bed reactor core. Variant porosity model can provide good prediction of heat transfer phenomena than uniform porosity model. Especially it can explain some transient analysis results.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

scholarly journals Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core

2020 ◽  
Vol 35 (3) ◽  
pp. 201-207
Author(s):  
Surian Pinem ◽  
Tagor Sembiring ◽  
Tukiran Surbakti
Keyword(s):  
Steady State ◽  
Reactor Core ◽  
Maximum Temperature ◽  
Diffusion Method ◽  
Accident Analysis ◽  
Calculation Results ◽  
Foil Target ◽  
Enriched Uranium ◽  
Initial Power ◽  
Reactivity Insertion

Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity insertion, the coolant, fuel, and clad temperature are observed. The calculation results for the initial power of 1 W showed that the maximum temperature of the coolant, clad, and fuel are 49.76?C, 65.01?C, and 65.26?C, respectively. Meanwhile, when the reactivity insertion at the initial power of 1 MW, the maximum temperature of the coolant, clad, and fuel are 72.23?C, 140.79?C, and 141.97?C, respectively. Based on those calculation results during irradiation low enriched uranium foil target, the temperature in the steady-state and reactivity insertion accident does not exceed the allowable safety limit.

Design of a Discharge Header and a Working Platform for a Research Reactor With a CFD Model

2013 ◽  
Author(s):  
Kyoungwoo Seo ◽  
Hyungi Yoon ◽  
Dae-young Chi ◽  
Seonghoon Kim ◽  
Juhyeon Yoon
Keyword(s):  
Research Reactor ◽  
Cold Water ◽  
Cooling System ◽  
Reactor Core ◽  
Water Layer ◽  
Hot Water ◽  
Primary Coolant ◽  
Radiation Level ◽  
Reactor Pool ◽  
Pool Type

Most research reactors are designed as an open-pool type and the reactor is located on the bottom of the open-pool. The reactor in the pool is connected to the primary cooling system, which is designed for adequate cooling of the heat generated from the reactor core. One of the characteristics of an open-pool type research reactor is that the primary coolant after passing through the reactor core and the primary cooling system (PCS) is returned to the reactor pool. Because the primary coolant contains many kinds of radionuclides, the research reactor should be designed to protect the radionuclides from being released outside the pool by a stratified stable water layer, which is formed between a hot water layer and cold water near the reactor and prevents the natural circulation of water in the pool. In this study, additional components such as a discharge header and a working platform inside the pool were developed to help diminish the radiation level to the pool top. To discharge coolant stably inside the reactor pool, a discharge header was installed at the end of the pool inlet pipe. Many holes were made in the discharge header to discharge the coolant slowly and minimize the disturbance of the hot water layer by the flow inside the pool. The working platform was also equipped inside the reactor pool to remove the convective flow near the pool top. The commercially available CFD code, ANSYS CFD-FLEUNT, was used to specifically design the discharge header and working platform for satisfying the requirement of the pool top radiation level. The computations were conducted to analyze the flow and temperature characteristics inside the pool for several geometries using an SST k-ω turbulent model and cell modeling, which were conducted to isolate the root cause of these differences and the given inlet conditions. The discharge header and working platform were designed using the CFD results.

scholarly journals COMPUTATIONAL FLUID DYNAMICS SIMULATION OF KARTINI REACTOR FUELED PLATE

2019 ◽  
Vol 4 (2) ◽  
pp. 33-38
Author(s):  
Tri Nugroho Hadi Susanto
Keyword(s):  
Fluid Dynamics ◽  
Heat Flux ◽  
Computational Fluid Dynamics ◽  
Cooling System ◽  
Reactor Core ◽  
Dynamics Simulation ◽  
Maximum Temperature ◽  
Steady State Condition ◽  
Rectangular Channels ◽  
Made In

The purpose of this study is to determine the characteristics of the cooling system on the new design of the Kartini Reactor plate fuel based on numerical calculations (Computational Fluid Dynamics). The fuel plate model was simplified and made in 3D. The model dimensions are 17.3 mm x 68 mm x 900 mm. The space between the two plates called the narrow rectangular channels has a gap of 2 mm. On these simulations a heat flux of 10612,7 watt/m2 was used which was obtained from the MCNP calculation program. Simulations were conducted in a steady state condition and single-phase model laminar flow of an incompressible fluid through the gap between the two fuel plates. This simulation uses UDF (User Define Function) to approach heat flux behaviour that follows the neutron distribution in the reactor core. The simulation results show that the maximum temperature that occur at a flow rate of 0.01 m/s was 43.5 °C.

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