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.

Heat Transfer Problems in Nuclear Reactor Safety

1976 ◽  
Vol 190 (1) ◽  
pp. 215-224
Author(s):  
W. B. Hall
Keyword(s):  
Heat Transfer ◽  
Safety Assessment ◽  
Nuclear Reactor ◽  
Reactor Safety ◽  
Operating Characteristics ◽  
Nuclear Reactor Safety ◽  
Thermal Explosions ◽  
Transient Boiling ◽  
Transfer Problems

Reactor safety assessment is a highly specialized topic which, in many of its aspects, depends heavily on a satisfactory understanding of a wide variety of heat transfer phenomena. It is the aim of the paper to air some of these problems outside the ranks of the reactor safety specialists. Typical liquid-cooled reactors, their operating characteristics, and some heat transfer aspects of their safety assessment are discussed: for example, transient boiling, quenching of hot surfaces and thermal explosions.

Analytical representation of the solution of the point reactor kinetics equations with adaptive time step

2014 ◽  
Vol 70 ◽  
pp. 112-118 ◽  
Author(s):  
Sérgio Q. Bogado Leite ◽  
Daniel A.P. Palma ◽  
Marco Tullio de Vilhena ◽  
Bardo E.J. Bodmann
Keyword(s):  
Analytical Representation ◽  
Time Step ◽  
Reactor Kinetics ◽  
Adaptive Time Step ◽  
Adaptive Time ◽  
Point Reactor Kinetics

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

NPP

Gamification ◽  
2015 ◽  
pp. 735-769
Author(s):  
Eugene Brezhnev ◽  
Vyacheslav Kharchenko
Keyword(s):  
Nuclear Power ◽  
Input Data ◽  
Safety Assessment ◽  
Hybrid Approach ◽  
Bayesian Belief Network ◽  
Complex Nature ◽  
Reactor Safety ◽  
Integration Framework ◽  
Related Information ◽  
Different Types

The problem of the safe interaction between a Nuclear Power Plant (NPP) and a Power Grid (PG), considering the Fukushima nuclear accident, is becoming topical. There are a lot different types of influences between NPPs and PG, which stipulate NPPs' safety levels. To evaluate the influences, two metrics are proposed: linguistic and numerical. The approach to the NPP-PG safety assessment is based on the application of Bayesian Belief Network (BBN), where nodes represent different PG systems and links are stipulated by different types of influences (physical, informational, geographic, etc). It is suggested to evaluate criticality of the PG system considering the change of criticalities of all connected systems. The total criticality of each node in BBN is assessed considering particular criticalities caused by different types of influence. The complex nature of NPP and PG mutual interaction calls for the need for integration of different methods that use input data of different qualimetric nature (deterministic, stochastic, linguistic). Application of one specified group of risk methods might lead to loss and/or disregard of a part of safety-related information. BBN and Fuzzy Logic (FL) represent a basis for development of the hybrid approach to capture all information required for safety assessment of NPP – PG under uncertainties. Integration of FL-based methods and BBNs allows decreasing the amount of input information (measurements) required for safety assessment, when these methods are used independently outside from the proposed integration framework. An illustrative example for the NPP reactor safety assessment is considered in this chapter.

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