A comparative analysis of the neutronic performance of thorium mixed with uranium or plutonium in a high‐temperature pebble‐bed reactor

2021 ◽  
Author(s):  
Ouadie Kabach ◽  
Abdelouahed Chetaine ◽  
Abdelfettah Benchrif ◽  
Hamid Amsil ◽  
Fadi El Banni
Keyword(s):  
Comparative Analysis ◽  
High Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Neutronic Performance

Fuel elements of the high temperature pebble bed reactor

1975 ◽  
Vol 34 (1) ◽  
pp. 93-108 ◽  
Author(s):  
L. Wolf ◽  
G. Ballensiefen ◽  
W. Fröhling
Keyword(s):  
High Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Fuel Elements

scholarly journals Computational Model for the Neutronic Simulation of Pebble Bed Reactor’s Core Using MCNPX

2014 ◽  
Vol 2014 ◽  
pp. 1-12 ◽  
Author(s):  
J. Rosales ◽  
A. Muñoz ◽  
C. García ◽  
L. García ◽  
C. Brayner ◽  
...  
Keyword(s):  
High Temperature ◽  
Computational Model ◽  
Inherent Safety ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Triso Fuel ◽  
Very High Temperature Reactor ◽  
Fuel Design ◽  
High Temperature Reactor ◽  
Very High

Very high temperature reactor (VHTR) designs offer promising performance characteristics; they can provide sustainable energy, improved proliferation resistance, inherent safety, and high temperature heat supply. These designs also promise operation to high burnup and large margins to fuel failure with excellent fission product retention via the TRISO fuel design. The pebble bed reactor (PBR) is a design of gas cooled high temperature reactor, candidate for Generation IV of Nuclear Energy Systems. This paper describes the features of a detailed geometric computational model for PBR whole core analysis using the MCNPX code. The validation of the model was carried out using the HTR-10 benchmark. Results were compared with experimental data and calculations of other authors. In addition, sensitivity analysis of several parameters that could have influenced the results and the accuracy of model was made.

scholarly journals Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

2014 ◽  
Vol 2014 ◽  
pp. 1-16 ◽  
Author(s):  
Shixiong Song ◽  
Xiangzhou Cai ◽  
Yafen Liu ◽  
Quan Wei ◽  
Wei Guo
Keyword(s):  
Fluid Dynamics ◽  
High Temperature ◽  
Pressure Drop ◽  
Surface Temperature ◽  
Molten Salt ◽  
Thermal Hydraulics ◽  
Pore Scale ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Body Center Cubic

The present paper systematically investigated pore scale thermal hydraulics characteristics of molten salt cooled high temperature pebble bed reactor. By using computational fluid dynamics (CFD) methods and employing simplified body center cubic (BCC) and face center cubic (FCC) model, pressure drop and local mean Nusselt number are calculated. The simulation result shows that the high Prandtl number molten salt in packed bed has unique fluid-dynamics and thermodynamic properties. There are divergences between CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

Numerical Simulation of Accident Scenario in HTGR (Pebble Bed Reactor) Using COMSOL® Code

2013 ◽  
Author(s):  
Geoffrey J. Peter
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Computer Code ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Accident Scenario ◽  
Flow And Heat Transfer ◽  
Custom Made ◽  
Accident Scenarios ◽  
High Temperature Gas

High Temperature Gas Cooled Reactor (HTGR) development and operation is expanding in the United Kingdom, Russia, USA (Generation IV Reactors), and France (Pebble Bed Modular Reactor, PBMR). A prototype pebble bed reactor producing 10 MW thermal, High Temperature Reactor (HTR-10) is in operation in China. However, the general public remains skeptical of the safety and the perceived dangers of possible accidents. Of particular concern are blockages caused by local variations in flow and heat transfer that lead to hot spots within the bed. This paper models the accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors using the commercially available computer code COMSOL. Numerical modeling of flow and heat transfer in a packed bed produces an Elliptical Non-Linear Partial Differential equation that requires custom made computer codes. Previously published results obtained from the use of a custom-made verified computer code are limited to one accident scenario and involve considerable modification to study different accident scenarios. Thus the use of a commercially available computer code that can simulate many different accident scenarios is of considerable advantage. Further, this paper compares numerical solutions obtained from custom-made computer code with COMSOL simulation and discusses the advantages and limitations of both codes.

scholarly journals KARAKTERISASI RADIONUKLIDA PADA BAHAN BAKAR BEKAS DARI EXPERIMENTAL PEBBLE BED REACTOR

2019 ◽  
Vol 25 (1) ◽  
Author(s):  
Aisyah Aisyah ◽  
Mirawaty Mirawaty ◽  
Dwi Luhur Ibnu Saputra ◽  
Risdiyana Setiawan
Keyword(s):  
High Temperature ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
High Temperature Gas

KARAKTERISASI RADIONUKLIDA PADA BAHAN BAKAR NUKLIR BEKAS DARI EXPERIMENTAL PEBBLE BED REACTOR. Arbeitsgemeinschaft Versuchsreaktor (AVR) merupakan reaktor nuklir jenis High Temperature Gas Cooled Reactor (HTGR) yang menggunakan bahan bakar berbentuk pebble berlapis TRISO dengan tipe yang sama  dengan Reaktor Daya Eksperimental (RDE) yang direncanakan akan dibangun di Indonesia. Oleh karena itu karakteristik radionuklida dalam bahan bakar bekas (BBNB) reaktor AVR dapat digunakan untuk mempelajari karakteristik BBNB reaktor RDE. Salah satu hal penting dalam operasional reaktor nuklir adalah pengelolaan BBNB yang ditimbulkannya. Pengelolaan BBNB reaktor AVR dilakukan dengan penyimpanan dalam dry cask untuk jangka waktu yang lama. Upaya untuk mendisain keselamatan dalam sistem penyimpanan BBNB salah satu kajian penting yang diperlukan adalah karakterisasi radionuklida yang terkandung dalam BBNB. Pada penelitian ini dilakukan karakterisasi radionuklida yang terkandung dalam BBNB dengan menggunakan software ORIGEN 2.1 yang didasarkan pada operasional reaktor AVR. Penelitian ini bertujuan untuk analisis keselamatan penyimpanan BBNB pebble pada dry cask dalam jangka panjang. Hasil penelitian menunjukkan bahwa sampai dengan waktu penyimpanan selama 100 tahun, BBNB sebuah pebble memiliki karakteristik radionuklida hasil aktivasi, aktinida dan anak luruhnya, serta radionuklida hasil fisi dengan total konsentrasi aktivitas sebesar 4,03x1010 Bq/g. Sampai dengan waktu penyimpanan 100 tahun konsentrasi aktivitas radionuklida total dalam dry cask sebesar 7,66x1013 Bq/g untuk kapasitas dry cask yang berisi BBNB pebble berjumlah 1900 buah. Terdapat BBNB pebble dalam dry cask yang mengalami kerusakan pada lapisan TRISO, sehingga dalam  dry cask kemungkinan terdapat beberapa radionuklida hasil fisi yang dapat lepas dari BBNB  seperti 85Kr, 135Xe, dan 131I yang berupa gas, serta  137Cs,106Ru, 110mAg dan 107Pd yang bersifat logam.Kata kunci: Karakterisasi radionuklida, AVR, bahan bakar nuklir bekas, pebble berlapis TRISO

Gas-Cooled Modular Reactors With Spherical Fuel Elements — Their Inherent Safety and Applications Not Just for Power Generation

2014 ◽  
Author(s):  
Walter Jaeger ◽  
H. J. Hamel ◽  
Heinz Termuehlen
Keyword(s):  
Power Generation ◽  
High Temperature ◽  
Reactor Design ◽  
Inherent Safety ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
Fuel Elements ◽  
High Temperature Gas

The gas-cooled reactor design with spherical fuel elements, referred to as high-temperature gas-cooled reactors (HTGR or HTR reactors) or pebble bed reactors has been already suggested by Farrington Daniels in the late 1940s; also referred to as Daniels’ pile reactor design. Under Rudolf Schulten the first pebble bed reactor, the 46MWth AVR Juelich reactor (Atom Versuchs-Reactor Jülich) was built in the late 1960s. It was in operation for 22 years and extensive testing confirmed its inherent safety.

Effect of Helium Injection on Natural Circulation Onset Time in a Simulated Pebble Bed Reactor

2008 ◽  
Author(s):  
Joseph P. Yurko ◽  
Katrina M. Sorensen ◽  
Andrew Kadak ◽  
Xing L. Yan
Keyword(s):  
High Temperature ◽  
Natural Circulation ◽  
Onset Time ◽  
Cfd Model ◽  
Pebble Bed ◽  
Analytical Estimate ◽  
Pebble Bed Reactor ◽  
Experimental Apparatus ◽  
Air Ingress ◽  
High Temperature Gas

This paper describes the experimental validation of a proposed method that uses a small amount of helium injection to prevent the onset of natural circulation in high temperature gas reactors (HTGR) following a depressurized loss of coolant accident. If this technique can be shown to work, air ingress accidents can be mitigated. A study by Dr. Xing L. Yan et al. (2008) developed an analytical estimate for the minimum injection rate (MIR) of helium required to prevent natural circulation. Yan’s study used a benchmarked CFD model of a prismatic core reactor to show that this method of helium injection would impede natural circulation. The current study involved the design and construction of an experimental apparatus in conjunction with a CFD model to validate Yan’s method. Based on the computational model, a physical experimental model was built and tested to simulate the main coolant pipe rupture of a Pebble Bed Reactor (PBR), a specific type of HTGR. The experimental apparatus consisted of a five foot tall, 2 inch diameter, copper U-tube placed atop a 55-gallon barrel to reduce sensor noise from outside air movement. Hot and cold legs were simulated to reflect the typical natural circulation conditions expected in reactor systems. FLUENT was used to predict the diffusion and circulation phases. Several experimental trials were run with and without helium injection. Results showed that with minimal helium injection, the onset of natural circulation was prevented which suggests that such a method may be useful in the design of high temperature gas reactors to mitigate air ingress accidents.

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