PBMR Fuel Kernel Model for the Prediction of Accurate Temperature Profiles

2008 ◽  
Author(s):  
Onno Ubbink ◽  
Pieter S. du Toit ◽  
Pierre Lourens ◽  
Wessel R. Joubert
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Fission ◽  
Calculation Model ◽  
Tennis Ball ◽  
Fuel Temperature ◽  
Fuel Kernel ◽  
Graphite Sphere ◽  
Enriched Uranium ◽  
Definition Of ◽  
Fuel Elements

PBMR (Pebble Bed Modular Reactor) is a High-Temperature Gas-cooled Reactor (HTGR) utilizing spherical pebble like fuel elements. A pebble is a moulded graphite sphere about the size of a tennis ball that contains approximately 15000 homogeneously distributed, triso-coated low-enriched uranium dioxide (UO2) particles, about 1mm in diameter. In the case of fast reactivity transients the accurate time-dependent calculation of the uranium temperature is essential as the neutron balances in the nuclear reactor are strongly influenced by the actual fuel temperature. This paper presents a calculation model that calculates the temperature profile through a representative fuel kernel, its coating layers and the associated graphite moderator. The local nuclear fission heat is deposited directly in the fuel itself. Great care is taken with the definition of the boundary conditions and implementation thereof to ensure that the kernel temperature calculation model describes the physics as accurately as possible. This paper reports on this in detail. A sample calculation is included to illustrate the effect of and need for a more accurate model.

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 Changes of the Neutron Flux of the Nuclear Reactor Triga Mark III Since the Conversion from High to Low 235U Enrichment

2021 ◽  
Vol 8 (2) ◽  
pp. 149-153
Author(s):  
C. Vázquez-López ◽  
O. Del Ángel-Gómez ◽  
R. Raya-Arredondo ◽  
S. S. Cruz-Galindo ◽  
J. I. Golzarri-Moreno ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Nuclear Research ◽  
Reactor Power ◽  
Uranium Fuel ◽  
Acrylic Sheet ◽  
Enriched Uranium ◽  
Reactor Operating ◽  
Fuel Elements

The neutron flux of the Triga Mark III research reactor was studied using nuclear track detectors. The facility of the National Institute for Nuclear Research (ININ), operates with a new core load of 85 LEU 30/20 (Low Enriched Uranium) fuel elements. The reactor provides a neutron flux around 2 × 1012 n cm-2s-1 at the irradiation channel. In this channel, CR-39 (allyl diglycol policarbonate) Landauer® detectors were exposed to neutrons; the detectors were covered with a 3 mm acrylic sheet for (n, p) reaction. Results show a linear response between the reactor power in the range 0.1 - 7 kW, and the average nuclear track density with data reproducibility and relatively low uncertainty (±5%). The method is a simple technique, fast and reliable procedure to monitor the research reactor operating power levels.

scholarly journals Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Fuel Temperature ◽  
Safety Level ◽  
The Core ◽  
Mox Fuel ◽  
Calculation Results ◽  
Reactivity Coefficient ◽  
Fuel Elements

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>

Proposed Solutions to Achieve Nuclear Fusion

Engevista ◽  
2017 ◽  
Vol 19 (5) ◽  
pp. 1496
Author(s):  
Relly Victoria Virgil Petrescu ◽  
Raffaella Aversa ◽  
Antonio Apicella ◽  
Florian Ion Petrescu
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Nuclear Fusion ◽  
Nuclear Fission ◽  
Light Nuclei ◽  
Fast Neutrons ◽  
Fusion Reactions ◽  
Nuclear Fusion Reactions ◽  
The Masses ◽  
Fusion Products

Despite research carried out around the world since the 1950s, no industrial application of fusion to energy production has yet succeeded, apart from nuclear weapons with the H-bomb, since this application does not aims at containing and controlling the reaction produced. There are, however, some other less mediated uses, such as neutron generators. The fusion of light nuclei releases enormous amounts of energy from the attraction between the nucleons due to the strong interaction (nuclear binding energy). Fusion it is with nuclear fission one of the two main types of nuclear reactions applied. The mass of the new atom obtained by the fusion is less than the sum of the masses of the two light atoms. In the process of fusion, part of the mass is transformed into energy in its simplest form: heat. This loss is explained by the Einstein known formula E=mc2. Unlike nuclear fission, the fusion products themselves (mainly helium 4) are not radioactive, but when the reaction is used to emit fast neutrons, they can transform the nuclei that capture them into isotopes that some of them can be radioactive. In order to be able to start and to be maintained with the success the nuclear fusion reactions, it is first necessary to know all this reactions very well. This means that it is necessary to know both the main reactions that may take place in a nuclear reactor and their sense and effects. The main aim is to choose and coupling the most convenient reactions, forcing by technical means for their production in the reactor. Taking into account that there are a multitude of possible variants, it is necessary to consider in advance the solutions that we consider them optimal. The paper takes into account both variants of nuclear fusion, and cold and hot. For each variant will be mentioned the minimum necessary specifications.

Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor

2018 ◽  
Author(s):  
Chi Wang ◽  
Xuebei Zhang ◽  
Jingchao Feng ◽  
Muhammad Shehzad Khan ◽  
Minyou Ye ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Coolant Temperature ◽  
Medium Size ◽  
High Tech ◽  
Fuel Temperature ◽  
Flow Phenomena ◽  
Nuclear Reactor Safety ◽  
Cfd Method ◽  
Point Kinetics

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

scholarly journals Quantifying and Qualifying Alloys Based on Level of Homogenization: A U-10Mo Alloy Case Study

2019 ◽  
Vol 142 (1) ◽  
Author(s):  
Chao Wang ◽  
Zhijie Xu ◽  
Deborah Fagan ◽  
David P. Field ◽  
Curt Lavender ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Thermomechanical Processing ◽  
Measurement Data ◽  
Statistical Strategy ◽  
Minimum Number ◽  
Statistical Tolerance ◽  
Consistent Performance ◽  
Enriched Uranium ◽  
Formed Product ◽  
Molybdenum Concentration

Abstract Homogenization heat treatment is performed to attain uniformity in microstructure which is helpful to achieve the desired workability and microstructure in final products and, eventually, to gain predictive and consistent performance. Fabrication of low-enriched uranium alloys with 10 wt% molybdenum (U-10Mo) fuel plates involves multiple thermomechanical processing steps. It is well known that the molybdenum homogeneity in the final formed product affects the performance in the nuclear reactor. To ensure uniform homogenization, a statistical method is proposed to quantify and characterize the molybdenum concentration variation in U-10Mo fuel plates by analyzing the molybdenum concentration measurement data from scanning electron microscopy energy dispersive spectroscopy line-scan. Statistical tolerance intervals (TI) are employed to determine the qualification of the U-10Mo fuel plate. We formulate an argument for the minimum number of independent samples to define fuel plate qualification if no molybdenum measurement data are available in advance and demonstrate that the given TI requirements can be equivalently reduced to a sample variance criterion in this application. The outcome of the statistical analysis can be used to optimize casting design and eventually increase productivity and reduce fabrication costs. The statistical strategy developed in this paper can be implemented for other applications especially in the field of material manufacturing to assess qualification requirements and monitor and improve the process design.

CAREM X INPRO Case Study

2004 ◽  
Author(s):  
Pablo C. Florido ◽  
Dari´o Delmastro ◽  
Daniel Brasnarof ◽  
Osvaldo E. Azpitarte
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Surface Intermediate ◽  
Geological Repository ◽  
Enriched Uranium ◽  
Intermediate Storage ◽  
High Level

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

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