Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors

2021 ◽  
Author(s):  
Mohammad Alrwashdeh ◽  
Saeed Alameri
Keyword(s):  
High Temperature ◽  
Core Analysis ◽  
Two Dimensional ◽  
Triso Fuel ◽  
High Temperature Reactors ◽  
Fuel Particles ◽  
Very High ◽  
Full Core

Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors

2020 ◽  
Author(s):  
Mohammad Alrwashdeh ◽  
Saeed A. Alameri
Keyword(s):  
Monte Carlo ◽  
High Temperature ◽  
Coating Layer ◽  
Reactor Core ◽  
Two Dimensional ◽  
Triso Fuel ◽  
High Temperature Reactor ◽  
Reactor Type ◽  
Very High ◽  
Full Core

Abstract The Prismatic-core Advanced High Temperature Reactor (PAHTR) is a very high temperature reactor type which is one of promising reactor type technologies classified as Generation IV by the International Forum. The new technology designs are identified as being proliferation resistant, safe, economical, efficient, and long fuel cycle. In this paper, the continuous-energy Monte Carlo method is capable of capturing all of the necessary reactor physics parameters using high fidelity two dimensional model with Serpent Monte Carlo code, and applied for a large scale reactor core loaded with TRi-structural ISOtropic (TRISO) particle by taking into account the double heterogeneity effect. These analyses were performed for PAHTR reactor core that utilizes TRISO particles fuel embedded in graphite matrix by applying a new innovative idea of adding Integral Fuel Burnable Absorber (IFBA) as an additional coating layer with a designated thickness. Adding IFBA coating could lead to compressed excess reactivity at the Beginning of Cycle (BOC), and extended burnup cycle. The additional IFBA coating layer is placed in the outer surface of the fuel kernel and covered by the buffer layers that compose the TRISO fuel particle. Neutronic calculations were performed for both TRISO particle unit cell and for full core with homogenous distribution of IFBA coating.

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

2008 ◽  
Author(s):  
William J. O’Donnell ◽  
Amy B. Hull ◽  
Shah Malik
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Structural Integrity ◽  
Cyclic Creep ◽  
Creep Rupture ◽  
Creep Fatigue ◽  
Asme Code ◽  
High Temperature Reactors ◽  
Gen Iv ◽  
Very High

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

Tritium Permeation Characterization of Materials for Fusion and Generation IV Very High Temperature Reactors

2015 ◽  
Vol 67 (3) ◽  
pp. 475-478
Author(s):  
S. Thomson ◽  
K. Pilatzke ◽  
K. McCrimmon ◽  
I. Castillo ◽  
S. Suppiah
Keyword(s):  
High Temperature ◽  
High Temperature Reactors ◽  
Tritium Permeation ◽  
Characterization Of Materials ◽  
Very High

Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230, and Model Alloys

2009 ◽  
Vol 131 (6) ◽  
Author(s):  
Stéphane Gossé ◽  
Thierry Alpettaz ◽  
Sylvie Chatain ◽  
Christine Guéneau
Keyword(s):  
High Temperature ◽  
Heat Exchangers ◽  
Mass Spectrometric ◽  
Operating Conditions ◽  
Pure Substance ◽  
Haynes 230 ◽  
Inconel 617 ◽  
High Temperature Reactor ◽  
High Temperature Reactors ◽  
Very High

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHXs) of (very) high temperature reactors ((V)-HTRs). The behavior under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra- and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. The alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow (Rouillard, F., 2007, “Mécanismes de formation et de destruction de la couche d’oxyde sur un alliage chrominoformeur en milieu HTR,” Ph.D. thesis, Ecole des Mines de Saint-Etienne, France). To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T≈1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617, and model alloys 1178, 1181, and 1201. This coupling makes it possible for the thermodynamic equilibrium to be obtained between the vapor phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared with that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behavior in these alloys. These activity results call into question those previously measured by Hilpert and Ali-Khan (1978, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: I. Alloy IN-643,” J. Nucl. Mater., 78, pp. 265–271; 1979, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: II. Inconel Alloy 617 and Nimomic Alloy PE 13,” J. Nucl. Mater., 80, pp. 126–131), largely used in the literature.

Dodecahedron- and Bowl-Shaped Structures C20

1992 ◽  
Vol 270 ◽  
Author(s):  
Zdenék Slanina ◽  
Ludwik Adamowicz
Keyword(s):  
High Temperature ◽  
Temperature Region ◽  
Relative Stability ◽  
Three Dimensional ◽  
Two Dimensional ◽  
High Temperature Region ◽  
Relative Stabilities ◽  
One Dimensional ◽  
Very High ◽  
Quantumchemical Method

ABSTRACTPurely carbonaceous aggregates C20 have been studied by the AM1 quantumchemical method. In addition to one dodecahedron-shaped structure possessing C1 symmetry another three-dimensional species is revealed, viz. a bowl-shaped structureof C5v symmetry (and also one two-dimensional and two one-dimensional species). Temperature dependence of the relative stabilities of both three-dimensional structures is evaluated, showing that in the relevant temperature region the fullerenic species is prevailing. However, in a very high temperature region a relative-stability interchange has been predicted.

Development of High Temperature Gas-Cooled Reactor Fuel for Extended Burnup

2010 ◽  
Author(s):  
Shohei Ueta ◽  
Jun Aihara ◽  
Masaki Honda ◽  
Noboru Furihata ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Coating Layer ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Principal Function ◽  
Energy Agency ◽  
Triso Fuel ◽  
Coating Layers ◽  
Test Reactor ◽  
Fuel Particles

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].

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.

Development and Implementation of a 3D Heat Conduction Model for (Very) High Temperature Reactors Into the Reactor Dynamics Code DYN3D

2012 ◽  
Author(s):  
Silvio Baier ◽  
Ulrich Rohde ◽  
Soeren Kliem ◽  
Emil Fridman
Keyword(s):  
Heat Conduction ◽  
High Temperature ◽  
Block Type ◽  
Conduction Model ◽  
Heat Conduction Model ◽  
Reactor Dynamics ◽  
3D Effects ◽  
High Temperature Reactors ◽  
Short Time ◽  
Very High

The reactor dynamics code DYN3D was extended to treat phenomena in Block-type High Temperature Reactors (HTR). Therefor, a new heat conduction model was implemented into the code to tackle 3D effects of heat conduction and heat transfer. The first part of the paper describes the details of the heat conduction model. In the second part results of coupled neutron-kinetics/thermal-hydraulics calculations of steady state and short-time transients in block-type HTRs are discussed.

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