scholarly journals DEVELOPMENT OF METHODS FOR THE REFABRICATION OF EBR-II FUEL ELEMENTS. PART I. ENGINEERING CONSIDERATIONS FOR EBR-II FUEL REFABRICATION. PART II. DEVELOPMENT OF INJECTION CASTING METHODS AND EQUIPMENT. PART III. DEVELOPMENT OF FUEL PIN PROCESSING METHODS AND EQUIPMENT. PART IV. ASSEMBLY, WELDING, AND LEAK TESTING EBR-II FUEL RODS. PART V. DODIUM BONDING AND BOND TESTING EBR-II FU

1961 ◽  
Author(s):  
A.B. Shuck ◽  
J.E. Ayer ◽  
H.F. Jelinek ◽  
G.M. Iverson ◽  
N.J. Jr. Carson ◽  
...  
Keyword(s):  
Fuel Rods ◽  
Processing Methods ◽  
Equipment Part ◽  
Leak Testing ◽  
Injection Casting ◽  
Fuel Pin ◽  
Fuel Elements

scholarly journals Analysis of different factors on the features of thermal destruction of a fuel pin

2021 ◽  
Vol 2119 (1) ◽  
pp. 012099
Author(s):  
E V Usov ◽  
T A Saikina ◽  
V I Chuhno
Keyword(s):  
Numerical Simulation ◽  
Weight Loss ◽  
Thermal Destruction ◽  
Theoretical Calculations ◽  
Time Step ◽  
Fuel Rods ◽  
Program Module ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Fuel Elements

Abstract The presented work studies the influence of various factors that affect the specific features of fuel pins melting. For this purpose, fuel pins with different geometries and energy release are considered. Numerical simulation of melting is carried out using a program module for calculating the destruction of fuel rods. Comparison with theoretical calculations is made. The analysis of the convergence of calculations with respect to the time step value and the number of calculated cells along the radius and height is carried out. As a result of work with the use of numerical methods, the characteristic times of destruction of fuel elements during an accident with a loss of coolant flow rate (an accident of the ULOF type) and the dependence of weight loss on time are obtained under various conditions.

scholarly journals Preparation of U–Zr–Mn, a Surrogate Alloy for Recycling Fast Reactor Fuel

2015 ◽  
Vol 2015 ◽  
pp. 1-8 ◽  
Author(s):  
Jong-Hwan Kim ◽  
Ki-Hwan Kim ◽  
Chan-Bock Lee
Keyword(s):  
Fast Reactor ◽  
Reactor Fuel ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Thin Sections ◽  
Dynamic Vacuum ◽  
Target Composition ◽  
Injection Casting ◽  
Fast Reactor Fuel ◽  
Mn Content

Metallic fuel slugs of U–10Zr–5Mn (wt%), a surrogate alloy for the U–TRU–Zr (TRU: a transuranic element) alloys proposed for sodium-cooled fast reactors, were prepared by injection casting in a laboratory-scale furnace, and their characteristics were evaluated. As-cast U–Zr–Mn fuel rods were generally sound, without cracks or thin sections. Approximately 68% of the original Mn content was lost under dynamic vacuum and the resulting slug was denser than those prepared under Ar pressure. The concentration of volatile Mn was as per the target composition along the entire length of the rods prepared under 400 and 600 Torr. Impurities, namely, oxygen, carbon, silicon, and nitrogen, totaled less than 2,000 ppm, satisfying fuel criteria.

Design of High-Performance Fuel Pin Simulators for Thermodynamic Experiments with Nuclear Fuel Elements

1980 ◽  
Vol 47 (1) ◽  
pp. 153-162 ◽  
Author(s):  
Valentin Casal
Keyword(s):  
Nuclear Fuel ◽  
High Performance ◽  
Fuel Pin ◽  
Fuel Elements

Introduction of leak-testing method and system for cans of rod-type fuel elements of reactor of first block at the Beloyarsk Atomic Power Plant

1978 ◽  
Vol 45 (3) ◽  
pp. 891-891
Author(s):  
S. V. Shirokov ◽  
�. I. Snitko ◽  
V. N. Shiryaev
Keyword(s):  
Power Plant ◽  
Atomic Power Plant ◽  
Testing Method ◽  
Leak Testing ◽  
Fuel Elements ◽  
Type Fuel

scholarly journals A Procedure to Address the Fuel Rod Failures During LB-LOCA Transient in Atucha-2 NPP

2010 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria ◽  
Oscar Mazzantini
Keyword(s):  
Hydraulic System ◽  
Initial Conditions ◽  
Probabilistic Method ◽  
Three Dimensional ◽  
Computer Code ◽  
Calculation Methods ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pin ◽  
Computer Codes

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes. This may imply the use of a dedicated fuel rod thermo-mechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermo-mechanical code. The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes (RELAP5-3D©) calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, with the aim of a better understanding of the uncertainties involved and their technological consequences on the behavior of the fuel rods, not addressed in the current paper.

Leak Testing of EBR-II Fuel Rods

1962 ◽  
Vol 12 (3) ◽  
pp. 419-423 ◽  
Author(s):  
A. P. Grunwald
Keyword(s):  
Fuel Rods ◽  
Leak Testing

A Study of Superhomogenization Applied to PHWR Lattices

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Thomas A. Ferguson ◽  
Eleodor M. Nichita
Keyword(s):  
Cross Sections ◽  
Homogenization Method ◽  
Single Node ◽  
Rectangular Array ◽  
Water Reactor ◽  
Production Type ◽  
Fuel Pin ◽  
Heavy Water Reactor ◽  
Fuel Elements ◽  
Full Core

Abstract To reduce computational expenses, full-core production-type neutronics calculations are customarily performed using a simplified core-model whereby large regions of the core, called nodes, are assumed to be homogeneous. The process of generating the few-group homogenized-node macroscopic cross sections is called lattice homogenization. The simplest homogenization method is standard homogenization (SH) and full-core models based on it do not usually reproduce heterogeneous-core calculations too closely. To improve agreement between node-homogenized core results and heterogeneous-core results, advanced homogenization techniques are used. Such techniques tend to use additional parameters besides homogenized macroscopic cross sections. Superhomogenization (SPH) is an advanced lattice homogenization method, which has been developed initially for light-water-reactor (LWR) lattices whereby fuel elements are arranged in a rectangular array. It has the advantage of not requiring any modification to the full-core diffusion code for its implementation. For LWRs, SPH establishes neutronic equivalence between detailed-geometry heterogeneous fuel-pin cells and homogenized fuel-pin cells by adjusting homogenized multigroup macroscopic cross sections and diffusion coefficients. This work investigates the possible use of the SPH methodology for pressurized heavy-water reactor (PHWR) lattices whose fuel pins are arranged in concentric rings rather than in a rectangular array. Results for single-node (SN) as well as multinode (MN) lattice-calculation models are presented. Results show that, with proper region definition, the SPH methodology can be used for PHWR lattices but that improvement in homogenization accuracy is only marginal compared with SH when comparing results for the same type of lattice model (SN or MN).

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