scholarly journals Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium-steady-state thermal-hydraulic analysis

2013 ◽  
Author(s):  
E. E. Feldman ◽  
B. Dionne ◽  
L. P. Foyto ◽  
M. Kalimullah ◽  
K. Kutikkad ◽  
...  
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Hydraulic Analysis ◽  
University Of Missouri ◽  
Enriched Uranium ◽  
The University ◽  
Technical Basis ◽  
Thermal Hydraulic Analysis

Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor under Steady-State and Transient Conditions

2001 ◽  
Vol 135 (1) ◽  
pp. 51-66 ◽  
Author(s):  
M. Q. Huda ◽  
S. I. Bhuiyan ◽  
T. K. Chakrobortty ◽  
M. M. Sarker ◽  
M. A. W. Mondal
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Hydraulic Analysis ◽  
Transient Conditions ◽  
Triga Mark Ii ◽  
Thermal Hydraulic Analysis

scholarly journals Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

2017 ◽  
Author(s):  
J. A. Stillman ◽  
E. E. Feldman ◽  
D. Jaluvka ◽  
E. H. Wilson ◽  
L. P. Foyto ◽  
...  
Keyword(s):  
Research Reactor ◽  
University Of Missouri ◽  
Enriched Uranium ◽  
The University

scholarly journals Startup, transition core, and molybdenum-99 production upgrade analyses for low-enriched uranium fuel conversion at the University of Missouri research reactor

2020 ◽  
Author(s):  
◽  
Wilson Cowherd
Keyword(s):  
Low Power ◽  
Research Reactor ◽  
The United States ◽  
Fuel Conversion ◽  
Reactor Physics ◽  
University Of Missouri ◽  
Safety Margins ◽  
Enriched Uranium ◽  
The University ◽  
The Impact

Under the direction of the United States Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program, the University of Missouri Research Reactor (MURR®) plans to convert from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Low power physics startup test predictions, transition core planning, and analysis for a proposed fission-based molybdenum-99 production upgrade were done in support of LEU fuel conversion. As a first step to LEU fuel conversion, low-power physics tests will be performed to calculate reactor physics parameters. These parameters include flux distributions, coefficients of reactivity, and critical assembly measurements. To facilitate this test, reactor physics calculations were performed using MCNP5 to predict the values of these parameters. Implications of these predictions and areas of uncertainty in the prediction analysis are also discussed. Once MURR completes the testing of the initial LEU core, MURR will enter into a series of transition cycles until steady-state mixed-burnup operation is reached. A Python program was developed that incorporated the constraints of MURR operation while minimizing the time MURR will have to operate atypically during the transition cycles. The impacts of the transition cycles on experiment performance are reported, as well as the number of fuel elements needed. Finally, preliminary analysis on a proposed molybdenum-99 production device at MURR was performed. This analysis shows the impact on the reactor power distribution with implications to predicted safety margins as a part of the larger scope of the experiment analysis.

scholarly journals Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

2011 ◽  
Vol 26 (1) ◽  
pp. 45-49 ◽  
Author(s):  
Atta Muhammad ◽  
Masood Iqbal ◽  
Tayyab Mahmood
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
The Core ◽  
Computer Codes ◽  
Neutronic Parameters ◽  
Thermal Hydraulic Analysis

The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

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