Poster 1. Underwater fuel pin crud sampling equipment

1988 ◽  
pp. 181-183
Author(s):  
D. A. Bridle ◽  
P. Cake
Keyword(s):  
Fuel Pin

Investigation of fuel microstructure at the top of a metallic fuel pin after a reactor overpower transient

2021 ◽  
Vol 544 ◽  
pp. 152711
Author(s):  
Fidelma G. Di Lemma ◽  
Karen E. Wright ◽  
Luca Capriotti ◽  
Adam X. Zabriskie ◽  
Alexander J. Winston ◽  
...  
Keyword(s):  
Metallic Fuel ◽  
Fuel Pin

Analysis of Advanced PWR Loading Schemes for Transuranic Incineration in Thorium

2013 ◽  
Author(s):  
Benjamin A. Lindley ◽  
N. Zara Zainuddin ◽  
Fausto Franceschini ◽  
Geoffrey T. Parks
Keyword(s):  
Temperature Coefficient ◽  
Preliminary Analysis ◽  
Detrimental Effect ◽  
Spatial Separation ◽  
Trade Off ◽  
The Core ◽  
Positive Reactivity ◽  
Fuel Pin ◽  
Control Rods

It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favorable impact on the MTC and void coefficient fostered by use of thorium (Th), the possibility of performing Th-TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. The simplest way to reduce the moderation in a PWR is to increase the fuel pin diameter. This configuration improves the trade-off between achievable burn-up and MTC, but is ultimately limited by thermal-hydraulic constraints. Heterogeneous recycle with the bred uranium (U3) and the TRU are arranged in separate pins was found to be neutronically preferable to a homogeneous configuration. Spatial separation also enables the U3 and TRU to be refueled on different batch schemes. These techniques allow satisfactory discharge burn-up while ensuring negative MTC and fully voided reactivity, with the pin diameter of a standard PWR increased from 9.5 mm to 11 mm. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron. It seems necessary to control the core using control rods to keep the fully voided reactivity negative. A preliminary analysis indicates that this is feasible.

Optimizing fuel-pin assembly location in a research reactor

Atomic Energy ◽  
1992 ◽  
Vol 72 (6) ◽  
pp. 545-547
Author(s):  
Yu. P. Malers
Keyword(s):  
Research Reactor ◽  
Fuel Pin

Simulation of delayed neutron precursor release due to wet rupture in a fuel pin and its detection methodology in PFBR

2021 ◽  
Vol 382 ◽  
pp. 111372
Author(s):  
Abhitab Bachchan ◽  
K. Devan ◽  
K. Yernamma ◽  
M. Alagan ◽  
K. Natesan ◽  
...  
Keyword(s):  
Delayed Neutron ◽  
Fuel Pin

A comparison of four tomographic methods for nuclear fuel pin analysis

1995 ◽  
Vol 224 (3) ◽  
pp. 199-206 ◽  
Author(s):  
Teddy Craciunescu ◽  
Carmen Niculae ◽  
Gheorghe Mateescu ◽  
Catrinel Turcanu
Keyword(s):  
Nuclear Fuel ◽  
Fuel Pin

scholarly journals Analysis of the cladding melt relocation along the surface of the fuel pin with help of the SAFR module

2021 ◽  
Vol 2057 (1) ◽  
pp. 012126
Author(s):  
E V Usov ◽  
P D Lobanov ◽  
I A Klimonov ◽  
T V Sycheva ◽  
V I Chuhno ◽  
...  
Keyword(s):  
Flow Regime ◽  
Good Accuracy ◽  
Film Flow ◽  
Main Flow ◽  
Heated Part ◽  
Laminar Film ◽  
Fuel Pin

Abstract The presented work is dedicated to the development of approaches to simulate cladding melt relocation along the surface of the fuel pin. Development of the approaches is based on the results of the experiments carried out at the NSI RAS and IT SB RAS. Features of the melt relocation are studied in the experiments. It is demonstrated that the laminar film flow regime in the heated part of the fuel simulator is the main flow regime. Model of the melt relocation is constructed. This model is the part of the SAFR module of the EUCLID/V2 coupled code. It is shown that the proposed approaches allow simulating the melt relocation with good accuracy.

scholarly journals Analisis Fraksi Volume Bahan Bakar Uranium Karbida Pada Reaktor Cepat Berpendingin Gas Menggunakan SRAC Code

2021 ◽  
Vol 3 (1) ◽  
pp. 13-18
Author(s):  
Ratna Dewi Syarifah ◽  
Nabil Nabhan MH ◽  
Zein Hanifah ◽  
Iklimatul Karomah ◽  
Ahmad Muzaki Mabruri
Keyword(s):  
Power Density ◽  
Volume Fraction ◽  
Japan Atomic Energy Agency ◽  
Energy Agency ◽  
Core Diameter ◽  
Homogeneous Core ◽  
Fuel Pin ◽  
Core Height ◽  
Data Libraries ◽  
Code Standard

Analysis of fuel volume fraction with uranium caride fuel in Gas Cooled Fast Reactor (GFR) with SRAC Code is has been done. The calculation used SRAC Code (Standard Reactor Analysis Code) which is developed by JAEA (Japan Atomic Energy Agency), and the data libraries nuclear used JENDL 4.0. There are two calculation has been used, fuel pin cell calculation (PIJ Calculation) and core calculation (CITATION Calculation). In core calculation, the leakage is calculated so the calculation more precise. The CITATION calculation use two type of core configuration, i.e. homogeneous core configuration and heterogeneous core configuration. The power density value of two type core configuration is quite difference. It is better use heterogeneous core configuration than homogeneous core configuration, because the power density of heterogeneous core configuration is flatter than the other. From the analysis of fuel volume fraction, when the volume fraction is increase, the k-eff value is increase. And the optimum design after has been analysis for fuel volume fraction, that is the fuel volume fraction is 49% with a heterogeneous core configuration of three types of fuel percentages, for Fuel1 9%, Fuel2 12% and Fuel3 15%. This reactor is cylindrical, has a core diameter of 240 cm and a core height of 100 cm.

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