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.

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

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.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Comparison of CORA and MELCOR Core Degradation Simulation and MELCOR Oxidation Model

2014 ◽  
Author(s):  
Jun Wang ◽  
Michael L. Corradini ◽  
Wen Fu ◽  
Troy Haskin ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Test Data ◽  
Flow Rate ◽  
Hydrogen Generation ◽  
Severe Accident ◽  
Generation Rate ◽  
Accident Analysis ◽  
University Of Wisconsin ◽  
Calculation Results ◽  
Oxidation Model ◽  
The University

MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin – Madison. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed.

iB1350: Part 2 — Level1 PRA Considering Optimization of Safety Systems for the iB1350

2018 ◽  
Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Lessons Learned ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Active Safety ◽  
Decay Heat ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

Analysis of 3-Dimensional Hydrogen Behavior With Passive Containment Cooling

2017 ◽  
Author(s):  
Xingguan Huang ◽  
Tingzao Fu ◽  
Gaofeng Huang
Keyword(s):  
Cooling System ◽  
Water Film ◽  
Severe Accident ◽  
Accident Analysis ◽  
3 Dimensional ◽  
Film Evaporation ◽  
Lumped Parameter ◽  
Development Procedure ◽  
Hydrogen Analysis ◽  
Hydrogen Safety

Hydrogen safety is one of the most important issues for severe accident analysis. Comparing with the lumped parameter code, 3-D CFD code GASFLOW has more advantages for analyzing hydrogen related phenomena [1]. However, due to the lacking of passive containment cooling system (PCS) model, GASFLOW is inapplicable for hydrogen analysis of PWR with PCS. This paper shows the development procedure of PCS model, including the model introduction and validation. The main functions of the PCS model are the calculation of falling water film thickness, heat transfer between wall and film, and film evaporation. Then, the hydrogen safety analysis by GASFLOW with PCS model is performed for CAP1400.

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.

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>

Study on Control Strategy of Hydrogen Risk in Small Containment Under Severe Accident

2018 ◽  
Author(s):  
Zou Zhiqiang ◽  
Zhang Ming ◽  
Peng Huanhuan ◽  
Hou Liqiang ◽  
Deng Chunrui ◽  
...  
Keyword(s):  
Oxygen Concentration ◽  
Hydrogen Concentration ◽  
Control Strategy ◽  
Cooling System ◽  
Severe Accident ◽  
Condensation Process ◽  
Analysis Model ◽  
Gas Distribution ◽  
High Hydrogen ◽  
Small Steel

Hydrogen combustion or detonation happened in the containment within the process of the small reactor severe accident may threaten the integrity of the containment. In this paper, based on systemic design of the Small Modular Reactor (SMR) surrounded by the steel containment, an innovatory combustible gas control strategy which using the passive containment cooling system (PCCS) and passive autocatalytic recombiners (PARs) is made to control the hydrogen risk in the small steel containment. A severe accident hydrogen risk analysis model is built by the integrative severe accident analysis program MELCOR, the validity of the strategy is analyzed at a typical severe accident. With this understanding, a three-dimensional computed fluid dynamics hydrogen behavior analysis model of the small steel containment is established by GASFLOW code, and the gas distribution non-uniformity in the containment is analyzed. The result shows that the steam condensation process in the containment could be slowed down by controlling the action of PCCS, and the steam concentration in the containment could be in the range of high level, while the oxygen concentration could be in the range of low level. If the PARs were added, the PARs could consume the hydrogen and oxygen in the containment sustainedly. The containment atmosphere could be in an inerted condition during the accident process, even though the hydrogen concentration in the containment is high. The gas distribution non-uniformity analysis result shows that oxygen concentration was low in the extent of high hydrogen concentration and high steam concentration, the steam, oxygen and hydrogen distribution non-uniformity would not affect the inerted atmosphere of containment.

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.

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