The Effects of the Number of Coated Fuel Particles on the Neutronic Aspects of 25 MWt Pebble Bed Reactor with Thorium Fuel

High-Temperature Gas Reactor (HTGR) is a type of reactor that continues to be developed because of its advantages in terms of economic aspects, proliferation resistance, and safety aspects. One of the safety aspect improvements is due to the use of the Coated Fuel Particle (CFP). A coated fuel particle is a fuel with a diameter smaller than 1 mm and is protected by several carbon layers. In the Pebble Bed Reactor (PBR) type of HTGR design, the CFP is placed in a 6 cm fuel ball. How much CFP is put into the fuel ball will determine the neutronic characteristics of the reactor. In this study, the effect of the amount of CFP in the fuel ball on the 25 MWt PBR design using Thorium fuel and its impact on several important neutronic aspects, such as the effective multiplication factor, the amount of fuel enrichment, the utilization of fissile material, and the density of the fissile material formed. The calculation was performed by the Monte Carlo MVP / MVP-BURN code. This study found that the coated fuel particle fraction of 15% was the optimum value for the studied neutronic parameters.

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Studi Pengaruh Fraksi Coated Fuel Particle pada Desain Pebble Bed Reactor 40 MWt Berbahan Bakar Uranium

Jurnal Pengembangan Energi Nuklir ◽

10.17146/jpen.2021.23.1.6096 ◽

2021 ◽

Vol 23 (1) ◽

pp. 1

Author(s):

Dwi Irwanto ◽

Nining Yuningsih

Keyword(s):

Monte Carlo ◽

Fuel Particle ◽

Pebble Bed ◽

Pebble Bed Reactor ◽

Coated Fuel Particle

Coated Fuel Particle (CFP) adalah tipe elemen bakar mikro berdiameter lebih kecil dari 1 mm, yang di dalamnya terdapat material fisil yang dilapisi oleh beberapa lapisan karbon. Pebble Bed Reactor (PBR) menggunakan konsep CFP untuk elemen bakarnya. CFP dimasukan ke dalam bola elemen bakar berukuran 6 cm dan disebar di dalam zona elemen bakar. Tujuan penelitian ini adalah untuk mempelajari pengaruh dari fraksi CFP terhadap beberapa parameter neutronik penting seperti faktor multiplikasi efektif, spektrum energi neutron, perubahan densitas material fisil dan fertil, serta tingkat utilisasi material fisil. Analisa dilakukan untuk pada sistem PBR berdaya 40 MWt dengan menggunakan kode Monte Carlo MVP/MVP-BURN, dengan fraksi CFP yang dianalisa berkisar antara 5-60%. Dari penelitian ini didapatkan bahwa fraksi CFP sebesar 10% memberikan nilai optimal untuk beberapa parameter neutronik terkait dan dapat dijadikan acuan untuk desain Pebble Bed Reactor berdaya 40 MWt dengan elemen bakar uranium.

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Effect of PyC Thermal Behavior on the Stress of TRISO-Coated Fuel Particles

Key Engineering Materials ◽

10.4028/www.scientific.net/kem.697.852 ◽

2016 ◽

Vol 697 ◽

pp. 852-857

Author(s):

Rong Li ◽

Bing Liu ◽

Chun He Tang

Keyword(s):

Pressure Vessel ◽

Elastic Strain ◽

Fission Products ◽

Fuel Particle ◽

Coated Particles ◽

Inner Pressure ◽

Higher Temperature ◽

Fuel Particles ◽

Coated Fuel Particle ◽

Increasing Temperature

TRISO coated fuel particle is the most important component in HTR fuel, the silicon carbide (SiC) coating layer is regarded as the pressure vessel to contain the fission products. During reactor operation, the inner pressure resulting from fission products and pyrocarbon (PyC) thermal effect will contribute to the failure of TRISO-coated particles. The higher temperature will result in the increasing of inner pressure and PyC thermal expansion, which will then change the stress of SiC layer. Considering the effects of temperature on inner-pressure expansion and elastic strain into the pressure vessel failure model, thermal effects on the stress of TRISO-coated particles were studied with analytical solution. The results indicated that the effects of inner pressure on the particle stresses were increasingly highlighted at the late stage of irradiation. And the increasing temperature caused a slight effect on PyC elastic modulus while elastic strain is unaffected greatly, either. Therefore, CFP stresses remain unchanged basically.

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Fission Products Distribution in TRISO Coated Fuel Particles Irradiated to 3.22 X 1021 n/cm2 Fast Fluence at 1092°C

ASME 2015 Nuclear Forum ◽

10.1115/nuclrf2015-49695 ◽

2015 ◽

Author(s):

Haiming Wen ◽

Isabella J. Van Rooyen ◽

Connie M. Hill ◽

Tammy L. Trowbridge ◽

Ben D. Coryell

Keyword(s):

Electron Microscopy ◽

Transmission Electron Microscopy ◽

Ion Beam ◽

Fission Products ◽

Fuel Particle ◽

Next Generation ◽

Research Activities ◽

Transmission Electron ◽

Fuel Particles ◽

Coated Fuel Particle

Mechanisms by which fission products (especially silver [Ag]) migrate across the coating layers of tristructural isotropic (TRISO) coated fuel particles designed for next generation nuclear reactors have been the subject of a variety of research activities due to the complex nature of the migration mechanisms. This paper presents results obtained from the electron microscopic examination of selected irradiated TRISO coated particles from fuel compact 1-3-1 irradiated in the first Advanced Gas Reactor experiment (AGR-1) that was performed as part of the Next Generation Nuclear Plant (NGNP) project. It is of specific interest to study particles of this compact as they were fabricated using a different carrier gas composition ratio for the SiC layer deposition compared with the baseline coated fuel particles reported on previously. Basic scanning electron microscopy (SEM) and SEM montage investigations of the particles indicate a correlation between the distribution of fission product precipitates and the proximity of the inner pyrolytic carbon (IPyC)-silicon carbide (SiC) interface to the fuel kernel. Transmission electron microscopy (TEM) samples were sectioned by focused ion beam (FIB) technique from the IPyC layer, the SiC layer and the IPyC-SiC interlayer of the coated fuel particle. Detailed TEM and scanning transmission electron microscopy (STEM) coupled with energy dispersive X-ray spectroscopy (EDS) were performed to identify fission products and characterize their distribution across the IPyC and SiC layers in the areas examined. Results indicate the presence of palladium-silicon-uranium (Pd-Si-U), Pd-Si, Pd-U, Pd, U, U-Si precipitates in the SiC layer and the presence of Pd-Si-U, Pd-Si, U-Si, U precipitates in the IPyC layer. No Ag-containing precipitates are evident in the IPyC or SiC layers. With increased distance from the IPyC-SiC interface, there are less U-containing precipitates, however, such precipitates are present across nearly the entire SiC layer.

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Optimization of a Radially Cooled Pebble Bed Reactor

Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 1 ◽

10.1115/htr2008-58117 ◽

2008 ◽

Author(s):

B. Boer ◽

J. L. Kloosterman ◽

D. Lathouwers ◽

T. H. J. J. van der Hagen ◽

H. van Dam

Keyword(s):

Pressure Drop ◽

Temperature Profile ◽

Flow Direction ◽

Outer Region ◽

Fissile Material ◽

Fuel Temperature ◽

Pebble Bed ◽

Pebble Bed Reactor ◽

The Core ◽

Power Profile

By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as 1 cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = −2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = −50 °C) can be achieved compared to present axially cooled designs.

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Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

Nuclear Science and Technology ◽

10.53747/jnst.v6i2.143 ◽

2016 ◽

Vol 6 (2) ◽

pp. 21-30

Author(s):

Huu Tiep Nguyen ◽

Viet Phu Tran ◽

Tuan Khai Nguyen ◽

Vinh Thanh Tran ◽

Minh Tuan Nguyen

Keyword(s):

Monte Carlo ◽

Power Distribution ◽

Reactor Core ◽

Multiplication Factor ◽

Benchmark Problem ◽

Fuel Assemblies ◽

Effective Multiplication Factor ◽

Neutronic Characteristics ◽

Effective Multiplication ◽

Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

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Design Study of 300 MWt PWR Fueled With UO2 Coated Fuel Particle

Volume 4: Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition ◽

10.1115/icone14-89287 ◽

2006 ◽

Author(s):

Abu Khalid Rivai ◽

Ferhat Aziz ◽

Minoru Takahashi

Keyword(s):

Optimization Design ◽

Computer Code ◽

Cycle Period ◽

Fuel Particle ◽

Pressurized Water ◽

Type Reactor ◽

Ordinary Type ◽

Fuel Particles ◽

Safety Characteristic ◽

Coated Fuel Particle

A neutronic design was performed for 300 MWt Pressurized Water Reactor (PWR) with UO2 compacts made of coated fuel particles (CFP) comparing that with sintered pellets made of UO2 powder as ordinary fuel type. UO2 CFP type was enriched 4.8% of 235U and UO2 ordinary type was enriched 5% of 235U. Both reactors were operated with single batch refueling system with a cycle period of 3 years. The purpose of the design was to investigate the applicability of UO2 CFP type to PWR comparing with UO2 ordinary type that commonly used for PWR. The calculation was done with SRAC (Standard Reactor Analysis Code) computer code and nuclear library of JENDL-33. The results of calculation showed that k-effective for both type of fuel could be maintained at critical condition for 3 years operation without refueling. The k-effective and the Doppler coefficients have been calculated for all types of fuel at 600 K and 900 K degrees. The results of calculation showed that for all types of fuel Doppler coefficient was negative, which was good for inherent safety characteristic. The size optimization design showed that the active core dimensions of UO2 CFP type reactor was about 2 times larger than the UO2 ordinary type reactor.

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Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel

Computational And Experimental Research In Materials And Renewable Energy ◽

10.19184/cerimre.v2i2.27368 ◽

2019 ◽

Vol 2 (2) ◽

pp. 64

Author(s):

Ratna Dewi Syarifah ◽

Alvi Nur Sabrina

Keyword(s):

Spent Nuclear Fuel ◽

Reactor Core ◽

Fissile Material ◽

Uranium Carbide ◽

Burnable Poison ◽

The World ◽

Homogeneous Core ◽

Effective Multiplication Factor ◽

Carbide Fuel ◽

Effective Multiplication

A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and Americium in the world and also to reduce the effective multiplication factor (k-eff) value from the reactor. The results obtained that the addition of Neptunium and Americium causes the k-eff value was decreased at the beginning of burn-up time, but increase at the end of burn-up time. It was because Neptunium and Americium absorb neutrons at the beginning of burn-up time and turns into fissile material at the end of burn-up time. The addition of protactinium in the reactor causes the k-eff value to be decreased both at the beginning of the burn-up time and at the end of the burn-up time. It happens because Protactinium absorbs neutrons both at the beginning of the burn-up time and at the end of the burn-up time. Therefore protactinium is often called a burnable poison.

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Hot Case

Mechanical Engineering ◽

10.1115/1.2004-jan-1 ◽

2004 ◽

Vol 126 (01) ◽

pp. 27-29

Author(s):

Jeffrey Winters

Keyword(s):

National Laboratory ◽

Department Of Energy ◽

Zirconium Nitride ◽

Fast Neutrons ◽

Pebble Bed ◽

Pebble Bed Reactor ◽

Oak Ridge ◽

Ball Size ◽

Breeder Reactors ◽

Fuel Particles

The US Department of Energy has commissioned research into advanced gas-cooled designs that would employ extreme heat to generate hydrogen for use in fuel cell-powered vehicles, and such designs are being considered as part of the Generation IV Nuclear Energy Systems Initiative. Another design that has received much attention is known as the pebble bed reactor. In this scheme, coated fuel particles are formed into billiard ball-size spheres, which are stacked in the containment vessel. Gas flows through the gaps between the stacked balls to convey the heat to a heat exchanger. The pebble-bed design has been tested in Germany, and South Africa has an active program. The Oak Ridge National Laboratory team has been investigating using non-carbon coatings that would wave by fast neutrons. Materials such as zirconium nitride have been looked at with the goal of creating an easily dissolved ceramic that could be used in fast-breeder reactors.

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