Burn-Up Dependent Modelling of Fuel-To-Clad Gap Conductance and Temperature Predictions for Mixed-Oxide Fuel in the ESFR-SMART Core

2021 ◽  
Author(s):  
Jean H J Lavarenne ◽  
Evaldas Bubelis ◽  
Una Davies ◽  
Simone Gianfelici ◽  
Solène Gicquel ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Mixed Oxide ◽  
Oxide Fuel ◽  
Hydraulic Analysis ◽  
Fuel Temperature ◽  
Accurate Calculation ◽  
General Increase ◽  
Mixed Oxide Fuel ◽  
Thermal Hydraulic Analysis ◽  
Good Agreement

Abstract Accurate coupled neutronic-thermal-hydraulic analysis of SFRs requires an accurate calculation of the fuel-to-clad gap conductance. In this paper, the gap conductance of the ESFR-SMART MOX pins is investigated through modelling in seven independent fuel performance codes, to provide confidence in results and understand the uncertainties associated with the predictions. This paper presents a comparison of the conductance and predicted fuel temperature distribution between codes. The values produced from the codes are then combined to produce a best-estimate prediction of the gap conductance expressed as a function of nodal fuel rating and burn-up for all seven codes. A fit was applied to the data thus obtained. The spread between results is such that, to 95% confidence, conductance predictions may vary from the correlation by up to a factor of ~4. The gap conductance results show a general increase of conductance with fuel rating and burn-up, from 0.22 at 0 burn-up and 10 kW.m^(-1) to 0.45 at 0 burn-up and 50 kW.m^(-1) and to 1.00 W.?cm?^(-2).K^(-1) at 150 GWd.t^(-1) and 50 kW.m^(-1). Some spread between codes has been noted and appears to be consistent with the spread previously published. There is good agreement between codes at low burn-up for fuel temperature predictions. The spread between codes increases with burn-up due to multiple phenomena including JOG formation and clad swelling.

Weapons-Derived Mixed Oxide Fuel Test Irradiation Summary

2005 ◽  
Author(s):  
S. A. Hodge ◽  
R. N. Morris ◽  
L. J. Ott
Keyword(s):  
Mixed Oxide ◽  
Oxide Fuel ◽  
Mixed Oxide Fuel

scholarly journals A Concept Study on Advanced Radioisotope Solid Solutions and Mixed-Oxide Fuel Forms for Future Space Nuclear Power Systems

2021 ◽  
pp. 1-9
Author(s):  
Richard M. Ambrosi ◽  
Daniel P. Kramer ◽  
Emily Jane Watkinson ◽  
Ramy Mesalam ◽  
Alessandra Barco
Keyword(s):  
Power Systems ◽  
Solid Solutions ◽  
Mixed Oxide ◽  
Nuclear Power ◽  
Oxide Fuel ◽  
Concept Study ◽  
Mixed Oxide Fuel ◽  
Future Space ◽  
Space Nuclear Power

Evaluation of thermal properties of mixed oxide fuel by molecular dynamics

2000 ◽  
Vol 307 (1-2) ◽  
pp. 1-9 ◽  
Author(s):  
Kazuhiro Yamada ◽  
Ken Kurosaki ◽  
Msayoshi Uno ◽  
Shinsuke Yamanaka
Keyword(s):  
Molecular Dynamics ◽  
Thermal Properties ◽  
Mixed Oxide ◽  
Oxide Fuel ◽  
Mixed Oxide Fuel

Failure behavior of plutonium-uranium mixed oxide fuel under reactivity-initiated accident condition

1992 ◽  
Vol 188 ◽  
pp. 154-161 ◽  
Author(s):  
T. Abe ◽  
N. Nakae ◽  
K. Kodato ◽  
M. Matsumoto ◽  
T. Inabe
Keyword(s):  
Mixed Oxide ◽  
Failure Behavior ◽  
Oxide Fuel ◽  
Mixed Oxide Fuel ◽  
Accident Condition

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

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