scholarly journals NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

2016 ◽  
Vol 18 (1) ◽  
pp. 29 ◽  
Author(s):  
Tukiran Surbakti ◽  
Tagor Malem Sembiring
Keyword(s):  
Key Words ◽  
Power Density ◽  
Research Reactor ◽  
Safety Margin ◽  
Reactor Core ◽  
Safety Analysis ◽  
High Density ◽  
Point Of View ◽  
The Core ◽  
Test Fuel

Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini akan dibahas analisis keselamatan uji bahan bakar mini U7Mo-Al dan U6Zr-Al ditinjau dari segi neutronik. Perhitungan dilakukan dengan menggunakan program komputer WIMSD-5B dan Batan-3DIFF. Hasil analisis menunjukkan bahwa kedua bahan bakar uji dapat diiradiasi dengan derajat bakar < 70 % selama 12 siklus operasi tanpa melampaui batas keselamatan neutronik. Kerapatan panas bahan bakar uji U7Mo-Al lebih besar dari bahan bakar U6Zr-Al.  Kata kunci: Bahan bakar mini, analisis neutronik, teras reaktor,  analisis keselamatan

Structural Confinement and Safety Design of a Fast Reactor Core

1981 ◽  
Vol 103 (2) ◽  
pp. 289-294
Author(s):  
F. D. Ju ◽  
J. G. Bennett
Keyword(s):  
Fast Reactor ◽  
Lateral Pressure ◽  
Numerical Solutions ◽  
Homogeneous Medium ◽  
Safety Margin ◽  
Reactor Core ◽  
Gas Pressure ◽  
Equivalent Stiffness ◽  
Friction Forces ◽  
The Core

In certain fast-reactor designs, the core is an assemblage of a large number of containers of long, hexagonal, hollow cylinders mounted vertically. These so-called “hex-cans” sit individually on coolant nozzles held down by their own weight, and are held as a group laterally at two levels by two constraint rings. At operating temperature, the rings bear on the hex-can assembly because of differences in thermal expansion. The compression of the rings on the hex-can assembly serves to prevent lifting of the can individually or in groups because of any accidental buildup of gas pressure. In the analysis, it is observed that the large number of hexcans and the distribution of the temperature field are such that the cross section of the reactor core can be treated as in a locally uniform dilatational field. An approximate equation was developed relating the plane deformation of a hollow hex cylinder to the global lateral pressure. The parameters are the material constitution and the thickness index (the ratio of the interior and the exterior cross-flat dimensions). The effective range of the equation covers the thickness ratio from zero to the stability limit when the wall becomes too thin resulting in buckling under the lateral pressure. The design equation is exact for zero thickness index. For hollow hex cylinders, numerical solutions were also obtained by the finite element method as a comparison. For a thickness index of 0.9 to 0.95, the difference is less than 0.1 percent. The cylinder constitutive equation is then used to determine an equivalent stiffness for a solid hex cylinder that is to have the same deformation as the given hex-can. The entire planar core region is then analyzed as a homogeneous medium of the equivalent stiffness. The method was applied to the core confinement design for a fast reactor. The thermoelastic solution was then applied to a relatively simpler configuration than the actual case to give a measure of the lateral pressure. The available friction forces for various lift configurations were then obtained. The gas pressure necessary to overcome the minimum friction force thus resulted. In addition, using the lateral pressure, the safety margin of the wall thickness of the hex-can for stability failures was determined.

LBLOCA Analysis of CPR1000 NPP With Advanced Accumulator

2013 ◽  
Author(s):  
Hongwei Hu ◽  
Jianqiang Shan ◽  
Junli Gou ◽  
Bo Zhang ◽  
Haitao Wang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Low Pressure ◽  
Point Of View ◽  
Primary Coolant ◽  
Pressure Drops ◽  
The Core ◽  
Injection Pump ◽  
Model Combining ◽  
Large Flow

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Technique

2009 ◽  
Author(s):  
Tewfik Hamidouche ◽  
El Khider Si-Ahmed ◽  
Anis Bousbia-Salah ◽  
Jack Legrand
Keyword(s):  
Nuclear Power ◽  
Research Reactor ◽  
Safety Margin ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Standard Material ◽  
Computer Tools ◽  
Operation Conditions ◽  
Coupled Technique ◽  
Test Reactor

This paper investigates the possibility to extend standard computer tools and methods, commonly used in the safety technology of nuclear power reactors, to research reactor safety analysis. A 3-D Neutron Kinetics Thermal-Hydraulic code (3D-NKTH), based on coupling PARCS and RELAP5/3.3 codes, was developed for a standard Material Test Reactor (MTR). The assessment of the model has been performed by comparison of steady state calculations against conventional diffusion codes and Monte Carlo code results. The model is applied for the analysis of a rod ejection accident. The comparison of the 3D-NKTH code, with conventional conservative research reactor tools showed that 3D-NKTH provided a more realistic course of the accident and did not require to define hot channel parameters. This approach could also open new frontiers in the safety analysis of research reactor such as setting realistic safety margin and adequate limits and operation conditions for optimal utilization of research reactors.

scholarly journals Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core

2018 ◽  
Vol 4 (1) ◽  
pp. 7-11
Author(s):  
Pavel A. Alekseev ◽  
Aleksei D. Krotov ◽  
Mikhail K. Ovcharenko ◽  
Vladimir A. Linnik
Keyword(s):  
Power Density ◽  
Hexagonal Lattice ◽  
Reactor Core ◽  
The Core ◽  
Intermediate Neutron ◽  
Density Factor ◽  
Fission Power ◽  
Concentric Circles ◽  
Thermionic Conversion ◽  
Fuel Elements

The paper investigates the possibility for reducing the radial power peaking factor kr inside the core of a water-cooled water-moderated thermionic converter reactor (TCR). Due to a highly nonuniform power density, the TCR generates less electric power and the temperature increases in components of the thermionic fuel elements, leading so to a shorter reactor life. A TCR with an intermediate neutron spectrum has its thermionic fuel elements (TFE) arranged inside the core in concentric circles, this providing for a nonuniform TFE spacing and reduces kr. The water-cooled water-moderated TCR under consideration has a much larger number of TFEs arranged in a hexagonal lattice with a uniform pitch. Power density flattening in a core with a uniform-pitch lattice can be achieved, e.g., through using different fuel enrichment in core or using additional in-core structures. The former requires different TFE types to be taken into account and developed while the latter may cause degradation of the reactor neutronic parameters; all this will affect the design’s economic efficiency. It is proposed that the core should be split into sections with each section having its own uniform lattice pitch which increases in the direction from the center to the periphery leading so to the radial power density factor decreasing to 1.06. The number of the sections the core is split into depends on the lattice pitch, the TFE type and size, the reflector thickness, and the reactor design constraints. The best lattice spacing options for each section can be selected using the procedure based on a genetic algorithm technology which allows finding solutions that satisfy to a number of conditions. This approach does not require the reactor dimensions to be increased, different TFE types to be taken into account and developed, or extra structures to be installed at the core center.

scholarly journals Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

2015 ◽  
Vol 2015 ◽  
pp. 1-10 ◽  
Author(s):  
Patrícia A. L. Reis ◽  
Antonella L. Costa ◽  
Claubia Pereira ◽  
Maria Auxiliadora F. Veloso ◽  
Amir Z. Mesquita
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Reactor Core ◽  
Hydraulic Parameters ◽  
The Core ◽  
Steady State Operation ◽  
Transient Simulations ◽  
Abnormal Operation ◽  
Flow Blockage ◽  
Relap5 Code

Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

scholarly journals Prediction of Peak Temperatures of Nigeria Research Reactor Core Components under Several Reactivity Accident Tests

2017 ◽  
Vol 2 (1) ◽  
Author(s):  
Abdulhameed Salawu ◽  
Ganiyu I Balogun
Keyword(s):  
Surface Temperature ◽  
Research Reactor ◽  
Reactor Core ◽  
Small Dimension ◽  
Accident Analysis ◽  
Small Water ◽  
The Core ◽  
Positive Reactivity ◽  
Core Components ◽  
Reactivity Accident

The Nigeria Research Reactor-1 (NIRR-1) consists of small water cooled square cylindrical core of 23cm in diameter and 23cm high. The small dimension of the core of this reactor facilitated our choice of PARET to perform reactivity accident analysis for NIRR-1 system. Our goal in this work is to predict the peak temperature of some important Nigeria Research Reactor (NIRR-1) core components under several reactivity accident tests. At power levels below 80kW, there were no significant differences between the peak fuel centerline temperatures, the peak fuel surface temperature and the peak clad surface temperature in the hot channel as well as in the average channel. The result from the reactivity accident test shows that power can never rise to an uncontrollable level in the core of NIRR-1 under ramp or step insertion of up to 4mk of reactivity. The calculated temperature of the important core components (e.g. fuel and clad) in the two channels (during this reactivity accident test) were far below their melting point temperatures. Boiling of any kind was not observed during this reactivity accident test. Therefore, NIRR-1 can be operated safely even if there is an inadvertent addition of up to 4mk of positive reactivity

scholarly journals Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

2011 ◽  
Vol 26 (1) ◽  
pp. 45-49 ◽  
Author(s):  
Atta Muhammad ◽  
Masood Iqbal ◽  
Tayyab Mahmood
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
The Core ◽  
Computer Codes ◽  
Neutronic Parameters ◽  
Thermal Hydraulic Analysis

The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

scholarly journals Core parametric study for enhancing the radioisotope production in the IEA-R1 research reactor

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Giovanni Laranjo Stefani ◽  
Frederico Antônio Genezini ◽  
Thiago Augusto Santos ◽  
João Manoel de Losada Moreira
Keyword(s):  
Neutron Flux ◽  
Parametric Study ◽  
Research Reactor ◽  
Reactor Core ◽  
Radioisotope Production ◽  
Spent Fuel ◽  
The Core ◽  
Safety Parameters ◽  
Fuel Elements ◽  
Average Neutron

In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices.  The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install  irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed  core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.   

Generation of Pressure Distribution Inside a Decay Tank in a Research Reactor Using CFD

2013 ◽  
Author(s):  
Seong Hoon Kim ◽  
Kyoungwoo Seo ◽  
Dae-young Chi ◽  
Juhyeon Yoon
Keyword(s):  
Research Reactor ◽  
Gamma Ray ◽  
Cooling System ◽  
Perforated Plate ◽  
Reactor Core ◽  
Perforated Plates ◽  
Pump Operation ◽  
The Core ◽  
Cooling Loop ◽  
Irradiation Device

The Primary Cooling System (PCS) of a research reactor circulates coolant to remove the heat produced in the fuel or irradiation device. The core outlet coolant contains many kinds of radionuclides because it passes the reactor core [1]. As N-16 among them emits a very hard gamma ray, it not only causes radiation damage to some components but also requires very heavy shielding of the primary cooling loop. Since its half-life is 7.13s, its level can be effectively lowered by installing a decay tank including an internal structure to provide a transit time [2]. To ensure that the N-16 activity decreases enough before the coolant leaves the heavily shielded decay tank room, perforated plates are installed inside the decay tank. The perforated plates are designed to disturb and delay the PCS flow. Normally, when a flow from a relative narrow inlet nozzle goes out to an enlarged tank, it becomes a complex turbulent flow inside the tank. In addition, the PCS flow is frequently changed from zero to a normal flow rate owing to the research reactor characteristics. Thus, the integrity of the perforated plate shall be verified with the pump operation and shutdown condition.

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