[en] The precision of Monte Carlo calculations of quantities of neutron dosimetry strongly depends on precision of reaction rates prediction. Research reactor represents a very useful tool for validation of the ability of a code to calculate such quantities as it can provide environments with various types of neutron energy spectra. Especially, a zero power research reactor with well-defined core geometry and neutronic properties enables precise comparison between experimental and calculated data. Thus, at the VR-1 zero power research reactor, a set of benchmark experiments were proposed and carried out to verify the MCNP Monte Carlo code ability to predict correctly the reaction rates. For that purpose two frequently used reactions were chosen: He-3(n,p)H-3 and Au-197(n,γ)Au-198. The benchmark consists of response measurement of small He-3 gas filled detector in various positions of reactor core and of activated gold wires placed inside the core or to its vicinity. The reaction rates were calculated in MCNP5 code utilizing a detailed model of VR-1 reactor which was validated for neutronic calculations at the reactor. The paper describes in detail the experimental set-up of the benchmark, the MCNP model of the VR-1 reactor and provides a comparison between experimental and calculated data. - Highlights: • Use of zero power reactor for validation of reaction rates calculations. • Reaction rates measurement in reactor core by He-3 detector and Au wires. • Validation of reaction rates calculation by MCNP5. • Comparison of measured and calculated RR for different positions in core

[en] At the zero power reactor VR-1 a new experimental instrumentation for temperature reactivity effects demonstration and measurement is being designed. To create a credible design and to estimate relatively small reactivity changes with temperature in a precise and accurate way, the calculation scheme for such effects determination was verified through available benchmark at similarly designed reactor. Thus an older well described set of experiments performed at KUCA reactor and dedicated to isothermal temperature coefficient determination in the range from ambient temperatures to some 80^oC were recalculated using MCNP5. The effect of employment various data libraries and of the two codes and means of temperature-dependent nuclear data processing was evaluated and compared. Thus, calculations were performed with ENDF/B-VII, JEFF 3.1 a JENDL 3.3 libraries using the same data processing scheme; in the latter case, employment of NJOY and makxsf codes for processing the data to the temperature of interest was compared. Also, attention was paid to thermal-scattering-data temperature dependence and the importance of their proper handling to obtain accurate results was pointed out. Finally it was demonstrated the current status of achievable precision and accuracy of temperature reactivity effects calculations is sufficient for performing reliable design of the experimental instrumentation for zero power reactors. (author)

[en] VR-1 is a light water zero power pool type reactor. The core is assembled on a grid of 8 x 8 positions (each position possessing 71.5 mm x 71.5 mm) and is composed of 3 various tubular fuel assembly types: four-, six- and eight-tube fuel assemblies of IRT-4M type. This fuel consists of UO₂ dispersed in Al with an enrichment on U-235 of 19.7 %. This paper deals with the design of a new instrumentation for the measurement of thermal effect. As at the nominal power of ca. 1 kW the reactor operates at an ambient temperature and there is almost no heat generated, the instrumentation has to be designed as a closed loop possessing some external system for water heating. There are two types of nuclear fuels under consideration to be potentially used in the instrumentation that are: IRT-4M fuel assemblies which are used for standard operation of the reactor and and EK-10 pins. EK-10 pins use Mg-UO₂ enriched on 10 % of U-235 as a fuel meat. With both fuel types, an optimization of the instrumentation was carried out. From the results, several modules covering 1 to 4 positions of reactor grid were proposed and the corresponding achievable changes of reactivity for typical core of the reactor containing these module variations were calculated. Finally, the results were inter-compared and the optimal design was found and discussed. The calculation analysis was carried out with MCNP5 code and with the utilization of detailed MCNP model of the VR-1 reactor

[en] The possibilities of reactor core analysis by precise Monte Carlo codes are gradually increasing along with the accessibility of computing power. In the case of zero power research reactors, where temperature and burn-up effects remain negligible, model can approximate the reality to a very high degree. In such a case, most of calculation uncertainty can be caused by uncertainties in technical specifications of fuel and reactor internals. Thus performance of the modelling and its predictive power can be significantly improved via comparison with a large set of experimental data that can be acquired during reactor operation and via subtle tuning and improving the calculation model. The paper describes the case for neutronics calculations of VR-1 zero power reactor core. (author)

[en] Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. ⁷LiF:Mg,Ti and ⁷LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayed gamma rays from fission, the relative gamma dose rate distribution is within ± 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)

[en] Linearity of response belongs to fundamental characteristics of neutron detection systems. Research reactors are valuable tools for neutron detector non-linearity studies as they offer a wide range of neutron fluxes. For neutron detection systems working in pulse mode they enable to characterise detector response non-linearity from some hundreds of cps up to the maximum reachable count rates. The paper presents comparison of two methods for neutron pulse-mode detector non-linearity characterisation using VR-1 zero power reactor: (1) comparative method utilising the comparison of studied pulse-mode detection system with a response of gamma compensated ionisation chamber working in current mode, and (2) kinetics method utilising the asymptotic exponential power rise after positive reactivity insertion as a source of information on true signal. Further several approaches for dead time determination based on theoretical formulae describing paralysable and non-paralysable dead time behaviour of detectors were studied and their usability to characterise the count-rate dependent detector response was analysed. (authors)

[en] The description of human operator dynamic characteristics in mathematical terms compatible with control engineering practice is an essential prerequisite to the analytical treatment of manual reactor control systems. Safe reactor operation requires effective operator control through interaction with plant dynamics, manipulators and displays. Traditional static analysis methods consider only specific situations; they fail to adequately explain the mutual interactions between the operator and the reactor plant characteristics. In this paper we investigate the theory for describing operator-reactor characteristics based on the methods of conventional control engineering techniques. The primary purpose of the experiments reported is the validation of the quasi-linear operator model. (authors)

[en] Hands-on training with critical or subcritical assemblies is recommended in standards as a part of training and qualification for criticality safety engineers. It is supported by argumentation that it can provide better understanding of the factors that contribute to criticality safety. To develop the idea the paper shows the example of the VR-1 training reactor operated by the Czech Technical University in Prague, and describes its education and training capabilities and activities. It relates them to the perspective of criticality safety. (author)

[en] The Czech Technical University in Prague (CTU), particularly the Faculty of Nuclear Sciences and Physical Engineering has long tradition in nuclear education in the Czech Republic. The faculty has been established 60 years ago as a scientific support and human resource base for the beginning of nuclear programme in Czechoslovakia. An important milestone in the nuclear education was the construction and commissioning of the training reactor VR-1. This experimental educational facility has become an invaluable tool in nuclear education not only for CTU, but also for other universities in the Czech Republic. Although the VR-1 reactor provides enough opportunities for practical education in the field of nuclear engineering, reactor management permanently makes efforts to extend and enhance the practical education. The latest improvement is based on integration of a small compact neutron generator into the educational experiments at the VR-1 reactor. The coupling of the reactor and the neutron generator significantly increase utilization of the VR-1 reactor and enhance educational experiments. Particularly, the pulsed source methods will be implemented at the VR-1 reactor for determining the reactivity and kinetic parameters of the reactor. The neutron generator allows also the study of the reactor response to the neutron pulses at various reactor states (i.e. subcritical, critical and supercritical). (author)

[en] Highlights: • A non-linear model structure is fit to manual-control of the VR-1 reactor. • The estimated model is a good fit to experimental data. • The frequency response indicates a good closed-loop control system performance. • The output non-linearity shows potential evidence of operator performance. - Abstract: The description of human operator dynamic characteristics in mathematical terms compatible with control engineering practice is an essential prerequisite to the analytical treatment of manual reactor control systems. Safe reactor operation requires effective operator control through interaction with plant dynamics, manipulators and displays. Traditional static analysis methods consider only specific situations; they fail to adequately explain the mutual interactions between the operator and the reactor plant characteristics. In this paper we investigate the cause-and-effect behaviours of three VR-1 research reactor operators during reactivity disturbance experiments, based on the methods of conventional control engineering techniques. The primary purpose of experiments carried out at the VR-1 training reactor and reported in this paper is the validation of a quasi-linear cause-and-effect operator model.