E. S. Glushkov

Evaluated benchmark experiments at critical assemblies simulating features of an HTHR at the ASTRA facility

The design of the ASTRA facility and critical assemblies that simulate physics features of modular high-temperature reactors (HTHR-Ms) with a graphite moderator and reflectors loaded with fuel particles having multilayer ceramic coatings is described in detail. Geometrical dimensions of the main elements and regions of the critical assemblies, composition of the materials used, and experimental results for various configurations of the critical assemblies are presented. A detailed computational benchmark model allowing for the structural and compositional features of the critical assembly configurations in question is developed on the basis of all the above data. The results are to be used for verification of the neutronics codes used for calculations of high-temperature helium-cooled reactors.

Results of Neutronic Benchmark Analysis for a High Temperature Reactor of the GT-MHR Type

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark–results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (APOLLO2, TRIPOLI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section.Copyright

Development of Fast Helium Reactors in Russia

The need for fast reactors in nuclear power is discussed. The advantages of fast helium reactors over fast liquid-metal reactors are shown. Fast helium reactor concepts and the possibilities of improving breeding by using fuel of higher density than oxide fuel, specifically, carbides and nitrides, are examined. It is indicated that the development and use of high-temperature helium-cooled thermal- and fast-neutron reactors are promising as part of an international collaboration in nuclear power in the 21st century.

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

Experiments on small-size fast critical fuel assemblies at the AKSAMIT facility and their use for development of computational models

Small-size fast critical assemblies with highly enriched fuel at the AKSAMIT facility are described in detail. Computational models of the critical assemblies at room temperature are given. The calculation results for the critical parameters are compared with the experimental data. A good agreement between the calculations and the experimental data is shown. The physical models developed for the critical assemblies, as well as the experimental results, can be applied to verify various codes intended for calculation of the neutronic characteristics of small-size fast nuclear reactors. For these experiments, the results computed using the codes of the MCU family show a high quality of the neutron data and of the physical models used.

Determination of the Relative Power Density Distribution in a Heterogeneous Reactor from the Results of Measurements of the Reactivity Effects and the Neutron Importance Function

A method for experimental determination of the relative power density distribution in a heterogeneous reactor based on measurements of fuel reactivity effects and importance of neutrons from a californium source is proposed. The method was perfected on two critical assembly configurations at the NARCISS facility of the Kurchatov Institute, which simulated a small-size heterogeneous nuclear reactor. The neutron importance measurements were performed on subcritical and critical assemblies. It is shown that, along with traditionally used activation methods, the developed method can be applied to experimental studies of special features of the power density distribution in critical assemblies and reactors.

Profiling of energy deposition fields in a modular HTHR with annular core: Computational/experimental studies at the ASTRA critical facility

The paper presents the results obtained from the computational/experimental studies of the spatial distribution of the 235U fission reaction rate in a critical assembly with an annular core and poison profiling elements inserted into the inner graphite reflector. The computational analysis was carried out with the codes intended for design computation of an HTHR-type reactor.

Dry critical experiments and analyses performed in support of the TOPAZ-2 safety program

In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz‐2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water‐ or sand‐immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M‐II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz‐2 safety analyses. In this paper we describe the Narciss M‐II experimental configurations along with the associated calculational models and methods. We also present and compare t...

Test Facilities and Experience on Space Nuclear System Developments at the Kurchatov Institute

The complexity of space fission systems and rigidity of requirement on minimization of weight and dimension characteristics along with the wish to decrease expenditures on their development demand implementation of experimental works which results shall be used in designing, safety substantiation, and licensing procedures. Experimental facilities are intended to solve the following tasks: obtainment of benchmark data for computer code validations, substantiation of design solutions when computational efforts are too expensive, quality control in a production process, and “iron” substantiation of criticality safety design solutions for licensing and public relations. The NARCISS and ISKRA critical facilities and unique ORM facility on shielding investigations at the operating OR nuclear research reactor were created in the Kurchatov Institute to solve the mentioned tasks. The range of activities performed at these facilities within the implementation of the previous Russian nuclear power system programs is...

Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ‐II safety program

Presented is a brief description of the Narciss‐M2 critical assemblies, which simulate accidental water/wet‐sand immersion of the TOPAZ‐II reactor as well as water‐flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel‐out anticriticality‐device modification to the TOPAZ‐II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron‐transport method is outlined. Nuclear criticality safety of the TOPAZ‐II reactor with an incorporated anticriticality unit is demonstrated.