Possibility of simulating natural circulation in fast neutron reactors using a light water test facility

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN-1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the “black box” method were used and their principles and applicability were analyzed. Using the “black box” method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.

Physical protection system effectiveness calculation in nuclear reactor facility using EASI code: case study sabotage scenario

Every nuclear facility must pay attention to the 3S aspect (Safety-Security-Safeguard) to prevent nuclear accidents. One element in the nuclear security aspects includes a reliable Physical Protection System (PPS), which aims to ward off security disturbances and other illegal acts, i.e., sabotage, theft, Etc. This study evaluates the PPS performance by adversary-path analysis approach using the EASI code for hypothetical nuclear reactor facility to anticipate sabotage attacks as the highest consequences scenario. We perform the probability of interruption (PI) calculation as represented by the effectiveness of the PPS. The study results show that in the PPS design, calculating the PI value using the EASI code confirms the need to pay attention in determining the MVP. The results provide feedback for the PPS designer to accept the current design or strengthen it to obtain a reliable PPS.

Assessment of operation of safety channel signal cables at nuclear power plants under fire conditions

Introduction. Signal cables of safety systems, installed at nuclear power plants (NPPs), retain the ability to conduct modulated signals during the time period needed to switch the reactor facility to a safe mode. However, the ability of signal cables to transmit signals correctly in the high temperature gas medium, which is typical for the early stage of a room fi re, has not been exposed to research.Aims and objectives. The co-authors offer a theoretical assessment of the ability of NPP safety system cables to correctly transmit modulated electric signals if exposed to fi re and current loads. The theoretical research into the temperature of the conductor of a signal cable at the initial stage of fi re has been performed towards this end.Theoretical background. The steady state heat conduction equation, describing heat transmission from the cable core to the environment through the cylinder-shaped insulation layer, is used to measure the temperature of the cable strand.Results and discussion. Temperature dependences describing the relation between the temperature of the conductor of a single - strand and single-wire cable KNEPng(А)-HF on the gas medium temperature are obtained. Relations between the temperature of the gas medium in the room on fi re and the current intensity in the electric cable (if the cable is laid vertically) are presented with account taken of the dependence between the specifi c resistance of the wire and the temperature if the maximal permissible operating temperature of cable strands is 70 °С, the maximal permissible operating temperature of cable strands in the overload operation mode is 80 °С, and the maximal cable strand heating temperature is equal to 160 °С when the short-circuit failure occurs. Maximal current intensity values are obtained for various operating modes in the condition of temperatures typical for the initial stage of an indoor fire, they allow to correctly conduct modulated signals within the time period needed to switch the reactor facility to a safe mode.Conclusions. The developed mathematical model and results of numerical experiments allow to assess the infl uence of the temperature in the room of a nuclear power plant in case of fi re on the ability of a signal cable of the safety system to transfer undistorted modulated signals depending on current loads and signal cable laying patterns (whether it is laid vertically or horizontally), and also to expand the range of the room temperature dependence on the current load provided in Electrical Installations Code (EIC).

Building and Use of Infrastructure for Detector Development and Testing at VR-1 Reactor Facility

Activities related to detector development, testing, characterisation and applications belong to key research objectives of the VR-1 reactor facility. The contribution gives a review of related improvements, achievements, used approaches, methods, and trends.