An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors

Kerntechnik ◽  
2014 ◽  
Vol 79 (5) ◽  
pp. 451-459
Author(s):  
Do Quang Binh ◽  
Nguyen Hoang Hai
Keyword(s):  
Temperature Coefficient ◽  
Neutron Source ◽  
Coefficient Of Reactivity

scholarly journals CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

2012 ◽  
Vol 15 (3) ◽  
pp. 26-35
Author(s):  
Binh Quang Do ◽  
Hai Hoang Nguyen
Keyword(s):  
Temperature Coefficient ◽  
Neutron Source ◽  
Safety Assessment ◽  
Analytical Representation ◽  
Reactor Safety ◽  
Reactor Kinetics ◽  
Reactor Calculation ◽  
Coefficient Of Reactivity

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.

scholarly journals CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

2012 ◽  
Vol 15 (2) ◽  
pp. 5-14
Author(s):  
Binh Quang Do ◽  
Hai Hoang Nguyen
Keyword(s):  
Temperature Coefficient ◽  
Neutron Source ◽  
Safety Assessment ◽  
Analytical Representation ◽  
Reactor Safety ◽  
Reactor Kinetics ◽  
Average Value ◽  
Different Temperatures ◽  
Reactor Calculation ◽  
Coefficient Of Reactivity

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and it’s average value in a range of temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.

scholarly journals Experimental Estimation of Moderator Temperature Coefficient of the IPEN-MB-01 Reactor

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Rubens Cavalcante Da Silva
Keyword(s):  
Temperature Coefficient ◽  
Reactor Core ◽  
Mean Value ◽  
Total Power ◽  
Rectangular Array ◽  
Neutron Reflection ◽  
Experimental Estimation ◽  
The Mean ◽  
Using Data ◽  
Coefficient Of Reactivity

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28x26 rectangular array of UO2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4oC, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and  (Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them.

scholarly journals INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

2020 ◽  
Vol 22 (2) ◽  
pp. 54
Author(s):  
R. Andika Putra Dwijayanto ◽  
Dedy Prasetyo Hermawan
Keyword(s):  
Molten Salt ◽  
Temperature Coefficient ◽  
Fuel Cycle ◽  
Minor Actinides ◽  
Molten Salt Reactor ◽  
Inherent Safety ◽  
Safety Aspects ◽  
Safe Option ◽  
Enriched Uranium ◽  
Coefficient Of Reactivity

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

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