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.

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

2005 ◽  
Vol 32 (15) ◽  
pp. 1679-1692 ◽  
Author(s):  
Martina Adorni ◽  
Anis Bousbia-Salah ◽  
Tewfik Hamidouche ◽  
Beniamino Di Maro ◽  
Franco Pierro ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Research Reactor ◽  
Reactor Core ◽  
Total Flow ◽  
Flow Blockage

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.

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

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

scholarly journals Design and installation of a hot water layer system at the Tehran research reactor

2013 ◽  
Vol 28 (1) ◽  
pp. 18-24
Author(s):  
Sayedeh Mirmohammadi ◽  
Morteza Gharib ◽  
Parnian Ebrahimzadeh ◽  
Reza Amrollahi
Keyword(s):  
Research Reactor ◽  
Cooling Water ◽  
Fission Products ◽  
Water Layer ◽  
Partial Solution ◽  
Time Frame ◽  
Reactor Power ◽  
Hot Water ◽  
Layer System ◽  
Tehran Research Reactor

A hot water layer system (HWLS) is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

scholarly journals Evaluation of technological radiation sources in the pumping station of the primary circulation loop and on the top of the pool type research reactor

2018 ◽  
Vol 33 (1) ◽  
pp. 31-46
Author(s):  
Stoyan Kadalev
Keyword(s):  
Heat Carrier ◽  
Decay Time ◽  
Research Reactor ◽  
Irradiation Time ◽  
Nuclear Research ◽  
Circulation Loop ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
Radiation Sources ◽  
Pool Type

The present paper considers the approach to an assessment of technological radiation sources in the primary water-water reactor circulation loop. In principle, such an evaluation is a multidisciplinary task that covers not only the irradiation of the nuclei, the formation of new isotopes and their decay when they are unstable, but also calculations in the field of hydraulics in order to perform an assessment of the irradiation time and the decay time. A general and a more detailed review of the radiation sources formation in the nuclear facilities and the pool type research reactors with demineralized water as a heat carrier are prepared. The initial isotopic composition of the heat carrier has been adopted according to the Vienna Standard Mean Ocean Water recommended by the International Atomic Energy Agency. The general mathematical model of the processes of nuclei irradiation, the formation of new isotopes and their decay, the assessment of the irradiation time and the decay time is described in details, enabling the repetition of this evaluation to a particular facility. The presented approach is applied in the reconstruction design of the nuclear research reactor IRT-2000, Sofia, Bulgaria.

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