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.   

Neutronic Analysis for Tehran Research Reactor Upgrading

2002 ◽  
Author(s):  
Ebrahim Afshar ◽  
Alireza Shahidi
Keyword(s):  
Research Reactor ◽  
Three Dimensional ◽  
Operating Conditions ◽  
Monte Carlo Code ◽  
Radioisotope Production ◽  
The Core ◽  
Transport Code ◽  
Tehran Research Reactor ◽  
Fuel Elements ◽  
Good Agreement

This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of Tehran Research Reactor (TRR) from 5 to 10 MW (th), keeping the same core configuration and with minimum changes in the primary cooling circuit. The main aim of TRR upgrade is to increase the volume of radioisotope production. The neutronic analysis was carried out for a fresh core with 22 Standard Fuel Elements (SFE) under normal operating conditions. Two different calculational lines were used to simulate the neutronic behavior in the core and perform the necessary neutronic calculations. First, combination of cell calculation transport code WIMS-D/4 [1] and three dimensional core calculation diffusion code CITATION [2] were used to and next a Monte Carlo code MCNP-4B [3] together with point depletion code ORIGEN-2 [4] were used. The results obtained show good agreement between these two different schemes.

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.

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 Effects of various types of reflectors on the 99Mo production in the VVER-Ts reactor targets

2020 ◽  
Vol 6 (2) ◽  
pp. 89-92
Author(s):  
Oleg Yu. Kochnov ◽  
Pavel A. Danilov
Keyword(s):  
Neutron Flux ◽  
Reactor Core ◽  
Raw Material ◽  
The Core ◽  
Assessment Tasks ◽  
Materials Used ◽  
Core Characteristics ◽  
Water Space ◽  
Reactivity Margin ◽  
Effective Multiplication

The effects from introducing various types of reflectors in the VVR-Ts reactor core on the 99Мо production were analyzed. Earlier the effects of only the beryllium reflector on the VVR-Ts reactor core characteristics, such as reactivity margin, neutron flux in experimental channels, and activity of the accumulated 99Мо, were calculated. The calculations are based on a generated precision model of the core which comprises one experimental channel where targets are irradiated for the 99Мо production. The model was built using the SCALE code. The code allows a fairly broad range of calculations to be performed, from criticality estimation to radiological assessment tasks. As the result of the computational analysis of the model, such characteristics were obtained as the effective multiplication factor, the power density in the 99Мо production targets, the neutron flux in the target raw material, and the quantity of the produced 99Мо after 120 hours of irradiation. The data was compared with the results of similar calculations of the VVR-Ts reactor core parameters. Further, the list of the materials used extensively as the reactor core reflector or moderator was formed based on reference literature. A number of models were obtained and analyzed on its basis, in which the water space on the core periphery was substituted for the investigated materials.

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Yuiko Motome ◽  
Yoshiya Akiyama ◽  
Hiroyuki Murao
Keyword(s):  
Radiation Effects ◽  
Research Reactor ◽  
Reactor Core ◽  
Fission Products ◽  
Internal Exposure ◽  
Spent Fuel ◽  
Loss Of Function ◽  
Safety Research ◽  
Spent Fuel Storage ◽  
Fuel Storage

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

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 A proposal for a new U-D2O criticality benchmark: RB reactor core 39/1978

2012 ◽  
Vol 27 (1) ◽  
pp. 75-83
Author(s):  
Milan Pesic
Keyword(s):  
Heavy Water ◽  
Uranium Dioxide ◽  
Reactor Core ◽  
Nuclear Fission ◽  
Natural Uranium ◽  
Single Type ◽  
Uranium Metal ◽  
The Core ◽  
Enriched Uranium ◽  
Fuel Elements

In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidric (now Vinca) Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum) fuel elements (known as slugs) were obtained from the USSR in 1960 and 1976, respectively. The so-called ?clean? cores of the RB reactor were assembled from a single type of fuel elements. The ?mixed? cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations) have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8?2 cm and 78

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

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