[en] Highlights: • An embedded measuring system with enhanced operational capabilities is introduced to the scientists. • The design is low cost and reprogrammable. • The system design is dedicated to multi-detector experiments with huge data collection. • Non count loss effect Feynman-α experiment is performed in EHWZPR. • The results is compared with endogenous/inherent pulsed neutron source experiment. - Abstract: In this work, an embedded multi-input multi-million-channel MCS in a newly design is constructed for multi-detector experimental research applications. Important characteristics of the system are possible to be tuned based on experimental case studies utilizing the reprogrammable nature of the silicon. By means of differentiation of the integrated counts registered in memory, this system is featured as a zero channel advance time measuring tool ideal for experiments on time correlated random processes. Using this equipment, Feynman-α experiment is performed in Esfahan Heavy Water Zero Power Reactor (EHWZPR) utilizing three different in-core neutron detectors. One million channel data is collected by the system in 5 ms gate time from each neutron detector simultaneously. As heavy water moderated reactors are significantly slow systems, a huge number of data channels is required to be collected. Then, by making in use of bunching method, the data is analyzed and prompt neutron decay constant of the system is estimated for each neutron detector positioned in the core. The results are compared with the information provided by endogenous pulsed neutron source experiment and a good agreement is seen within the statistical uncertainties of the results. This equipment makes further research in depth possible in a range of stochastic experiments in nuclear physics such as cross correlation analysis of multi-detector experiments.

[en] Highlights: • Kinetic parameters of a typical MTR research reactor are calculated based on MCNIC method. • MCNIC method is applied to such a nuclear experimental facility of an existing MTR research reactor. • The results are compared with $1/\mathrm{\upsilon }$ poisoning, k-ratio, perturbation, and ZPRN methods as standard techniques. • The impact of cross section libraries on calculation of kinetic parameters in thermal systems is investigated. - Abstract: The accurate estimation of effective delayed neutron fraction and prompt neutron lifetime are of major concerns in safety analysis of nuclear reactors. In this paper, Tehran Research Reactor (TRR) is taken as a case study of MTR research reactors and kinetic parameters of the facility are estimated utilizing Monte Carlo Neutron Importance Calculation (MCNIC) method. The results are compared with $1/\mathrm{\upsilon }$ poisoning, k-ratio, perturbation, and zero power reactor noise (ZPRN) methods. Taking perturbation results as reference information, the percent error in effective delayed neutron fraction calculated by MCNIC, k-ratio, and ZPRN methods are −3.7, −6.8, and 1.7 respectively. These results for ℓ_p parameter calculated by MCNIC, 1/υ poisoning, and ZPRN methods are 2.9, 5.7, and 6.4 respectively. A good agreement is seen among different methods of estimation. Note that error originated from cross section library is a different source of error which is also discussed in the text. As MCNIC method considers the problem in a more realistic manner, considering neutron importance in calculations, this method is more reliable than the other methods. A quantitative and qualitative discussion is given in the paper to indicate the advantages and disadvantages of each method for the case study in TRR.

[en] Non-random event losses due to dead time effect in nuclear radiation detection systems distort the original Poisson process into a new type of distribution. As the characteristics of the distribution depend on physical properties of the detection system, it is possible to estimate the dead time parameters based on time interval analysis, this is the problem investigated in this work. A BF₃ ionization chamber is taken as a case study to check the validity of the method in experiment. The results are compared with the data estimated by power rising experiment performed in Esfahan Heavy Water Zero Power Reactor (EHWZPR). Using Monte Carlo simulation, the problem is elaborately studied and useful range for counting rates of the detector is determined. The proposed method is accurate and applicable for all kinds of radiation detectors with no potential difficulty and no need for any especial nuclear facility. This is not a time consuming method and advanced capability of online examination during normal operation of the detection system is possible