scholarly journals RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

2018 ◽  
Vol 20 (1) ◽  
pp. 23 ◽  
Author(s):  
Andi Sofrany Ekariansyah ◽  
Endiah Puji Hastuti ◽  
Sudarmono Sudarmono
Keyword(s):  
Fuel Element ◽  
Secondary Flow ◽  
Research Reactor ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Reactor Power ◽  
Severe Accident ◽  
Accident Analysis ◽  
Plate Temperature ◽  
The One

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational chararacteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element.Keywords: loss of flow, blockage, fuel plate, RSG-GAS, RELAP5 SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK) dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar.Kata kunci: kehilangan aliran, penyumbatan, pelat bahan bakar, RSG-GAS, RELAP5

Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

2010 ◽  
Author(s):  
Aimin Zhang ◽  
Yalun Kang
Keyword(s):  
Fuel Element ◽  
Neutron Flux ◽  
Research Reactor ◽  
High Heat ◽  
Reactor Power ◽  
High Heat Flux ◽  
Water Tank ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Plate Type

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Analysis of a Total Flow Blockage of a Fuel Assembly in a Typical MTR Research Reactor by RELAP5/MOD3.3

2004 ◽  
Author(s):  
Franco Pierro ◽  
Beniamino Di Maro ◽  
Martina Adorni ◽  
Anis Bousbia Salah ◽  
Francesco D’Auria
Keyword(s):  
Fuel Assembly ◽  
Research Reactor ◽  
Reactor Power ◽  
Severe Accident ◽  
Total Flow ◽  
Flow Area ◽  
Forced Circulation ◽  
Flow Blockage ◽  
Pool Type ◽  
Extreme Scenario

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool Type Research Reactor [1]. The reactor is core cooled and moderated by downward forced circulation of light water. The transient herein considered is the related to partial and total obstruction of a single Fuel Assembly (FA) cooling channel. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FA integrity. Two cases are analysed to emphasize the severity of the accident. The first one is a partial blockage of a single FA considering four different obstruction levels: 50%, 75%, 85% and 95% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FA. This study constitutes the first step of a larger work which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation.

MELCOR severe accident analysis for a natural circulation small modular reactor

2017 ◽  
Vol 100 ◽  
pp. 197-208 ◽  
Author(s):  
Longze Li ◽  
Tae Woon Kim ◽  
Yapei Zhang ◽  
Shripad T. Revankar ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Severe Accident ◽  
Accident Analysis ◽  
Small Modular Reactor

TRICO II Core Inventory Calculation and its Radiological Consequence Analyses

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
J. L. Muswema ◽  
G. B. Ekoko ◽  
J. K.-K. Lobo ◽  
V. M. Lukanda ◽  
E. K. Boafo
Keyword(s):  
Fuel Element ◽  
Research Reactor ◽  
Source Term ◽  
Democratic Republic Of Congo ◽  
Reactor Core ◽  
Lessons Learned ◽  
Severe Accident ◽  
Plane Crash ◽  
Reactor Building ◽  
Accident Scenarios

Two severe accident scenarios are investigated in this paper as they have never been considered previously in the safety analysis report (SAR) of the Congo TRIGA Mark II research reactor (TRICO II) in Kinshasa, the Democratic Republic of Congo. The source term is derived from the reactor core after two postulated accidents: (1) a large plane crash with total destruction of the reactor building and (2) full damage of one fuel element while the reactor building remains intact. Total effective dose (TED), after core inventory, and dose profiles to human organs are derived to assess the operational safety of the reactor. Results from the study will be used to upgrade the current SAR of the reactor as the reactor safety and licensing concepts are changing over the years; the knowledge and lessons learned from the past experience are being updated accordingly with the available data. TEDs to workers of the facility show that higher values are obtained at areas near the source term at the time of the plane crash accident, which dies out as quickly as the plume is carried away following predominant meteorological conditions at the site. The situation with one fuel element totally damaged poses no threat as far as radiation protection is concerned and reveals a maximum TED of 1.30×10−7  mSv at 100 m from the reactor core. This shows that the operation of this type of research reactor is reliable and safe.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

2015 ◽  
Vol 80 ◽  
pp. 409-415 ◽  
Author(s):  
F. Mohamed ◽  
A. Hassan ◽  
R. Yahaya ◽  
I. Rahman ◽  
M. Maskin ◽  
...  
Keyword(s):  
Research Reactor ◽  
Safety Analysis ◽  
Reliability Study ◽  
Probabilistic Safety

Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

scholarly journals Severe Accident Analysis Code SAMPSON Improvement for IMPACT Project.

2002 ◽  
Vol 45 (3) ◽  
pp. 607-614 ◽  
Author(s):  
Hiroshi UJITA ◽  
Takashi IKEDA ◽  
Masanori NAITOH
Keyword(s):  
Severe Accident ◽  
Accident Analysis

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

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