WWR-K Reactor Research Articles

SPECIFIC FEATURES OF APPLICATION OF CADMIUM AS A NEUTRON-ABSORBING SCREEN WHILE DOPING SILICON IN A NUCLEAR REACTOR

The gradient of the neutron field in a nuclear reactor and the requirements for the permissible spread of the specific electrical resistance over the volume of the silicon ingot makes it necessary to develop an irradiation device. This is especially true for large silicon ingots. One of the options for reducing the gradient of the neutron flux along the height of the ingot is the use of neutron-absorbing screens in the design of the irradiation device. At the WWR-K reactor, cadmium with a natural isotopic composition is used as a neutron-absorbing screen material. The paper presents the results of a study of an irradiation device with a cadmium screen. The effect of a cadmium screen on the neutron-physical characteristics of an irradiation device for silicon doping is shown.

Analysis of the possibility of irradiation-induced coloring of topaz in the WWR-K reactor

Nowadays, radiation technologies are widely used in industry, medicine and other fields. One of the particular cases of application of such technologies is the irradiation-induced coloring of topaz. As a result of the application of this technology, colorless topaz acquires a more noble, respectively, economically advantageous color. For this, different types of radiation can be used. In the case of irradiation-induced coloring of topaz using neutrons, it is necessary to formation optimal conditions for irradiation of topaz. First of all, in the topaz irradiation zone it is necessary to formation a "hard" neutron spectrum not only for the formation of color centers, but also to reduce the residual radioactivity in the topaz. In this work, the possibility of implementing the technology of irradiation-induced coloring of topaz in the WWR-K research reactor was evaluated. For this, the energy distribution of neutrons in the central and peripheral irradiation positions of the WWR-K reactor was determined. On the basis of the obtained values, the time to reach the target neutron fluence (1018 cm-2) was estimated. The influence of the ratio of the flux of fast to thermal neutrons on the activation of natural impurities in topaz is shown. The results of trial irradiation of topazes in the WWR-K reactor without formation special conditions are presented. Ways to optimize the conditions for irradiation of topazes in the WWR-K reactor in order to improve the economic efficiency of the work were proposed.

Time History of Performance Parameters of WWR-K Reactor during Gradual Replacement of the Water Reflector by a Beryllium One

The water-cooled WWR-K research reactor has been operating since 2016 using low-enrichment uranium fuel. In order to maintain high levels of generated power and reactivity margin sufficient for a 21-day operation cycle, the water-cooled core reflector of neutrons was gradually, depending on the fuel burnup, replaced by a beryllium one in order to prevent decreasing of the reactor performance. A series of calculation studies are performed with a simulation of the core region of the WWR-K reactor using an MCNP transport code including determination of neutron flux density, reactivity, efficiency of the control rods of the control and protection systems, as well as kinetic parameters critical for ensuring safe reactor operation. The calculated curves of the principal performance characteristics of the WWR-K reactor are presented and analyzed in terms of their use in scientific research and their effect on the reactor operation safety.

Methodical experiments on the study of gas release from materials of thermonuclear reactor blankets

The work is devoted to the problem related to the development of methods for conducting irradiation experiments with tritium-generating materials of fusion reactor blanks. The experiments were carried out on a facility designed to study the release of tritium from lithium ceramics, which is located at the WWR-K reactor (Almaty, Kazakhstan). A special irradiation ampoule device was developed, in which samples of lithium ceramics are placed in reactor experiments. Due to the fact that the total length of the irradiation device is about 5 m and during irradiation, samples are installed in the lower part of the device, where pressure control is technically not possible, out-of-pile methodological experiments were conducted, during which calibration coefficients were determined that link gas flows from the area of investigated samples to the installation with partial pressures recorded in the mass spectrometer area. The results of neutron-physical, thermophysical and vacuum calculations showed that with the developed ampoule design, irradiation can be carried out at temperatures of 773-1473 K and that the developed installation can register the gas flow in the system at the operating level of tritium and helium in the studied samples above 10-11 mol/s. Key words: blanket, fusion reactor, lithium ceramics, tritium, helium.

Analysis of lutetium-177 production at the WWR-K research reactor

Production of lutetium-177 using direct nuclear reaction 176Lu(n,γ)177Lu by WWR-K reactor neutrons on enriched LuCl3 (up to 82% of 176Lu) is described. Calculations were performed by MCNP6 transport code. Two different irradiation positions of the WWR-K research reactor were considered. Estimates of the maximum specific activity of the luthetium-177 are obtained for the reactor irradiation positions located: (a) in the reactor core centre, (b) in the core periphery. In these positions, thermal neutron flux is two times different.Experimental data was shown that k-factor is 1.5 for considered irradiation positions. The study shows that for the position located in the core center, the estimated maximum specific activity of lutetium-177 is 819 GBq/mg, is to be achieved after 15 days of irradiation. For the position located in the core periphery, specific activity of lutetium-177 is 561 GBq/mg, is to be achieved after 20 days of irradiation. Ratio of Lu-177m to Lu-177 specific activity is not more than 0.025 for both irradiation positions.

New neutron radiography and tomography facility TITAN at the WWR-K reactor

The TITAN is a new experimental facility for neutron radiography and tomography at the research nuclear reactor WWR-K at the Institute of Nuclear Physics (Almaty, Republic of Kazakhstan). The TITAN facility is located on the radial beamline No1 of the nuclear reactor WWR-K. The neutron beam with a maximal field of view of 200 × 200 mm is formed by a collimator system, and the characteristic parameter L/D ratio can be varied in the range 75–1400. The sapphire filter is used to suppress fast neutron radiation. The detector system is based on a scintillator screen with the high-resolution high-sensitive CCD-camera. The design, schematic layout and main technical parameters of the neutron imaging station are presented. The first obtained results of the neutron radiography and tomography experiments with industrial materials and several cultural heritage objects are presented.

EXPERIMENTAL BASE OF THE WWR-K REACTOR FOR IN-SITU INVESTIGATION OF TRITIUM RELEASE FROM MATERIALS

The WWR-K reactor is a unique 6 MW multi-purpose light-water research reactor. Desalinated water is the coolant and moderator, and beryllium is the reflector. Uranium dioxide enriched up to 19.7% on the uranium-235 isotope serves as fuel. Since the launch of the WWR-K reactor, studies of various materials of research reactors have been carried out on it. Beginning in 2000, work was carried out at the WWR-K reactor to study the materials of fusion reactor, namely, experiments to study the release of tritium from lithium ceramics. A facility was developed for in-pile reactor studies of tritium release from various candidate materials for a breeder blanket of a fusion reactor in an inert gas environment. A methodology was also developed for estimating the retention time of tritium in irradiated materials. In 2018, an installation for studying the materials of fusion reactors was created at the WWR-K reactor, at which it became possible to conduct experiments on irradiating samples in a vacuum. In this work, a description of the WWR-K reactor, experimental facilities, and their general technical capabilities as applied to the expected studies of tritium yield from the fusion reactor materials at the WWR-K reactor is presented.

The WWR-K Reactor Experimental Base for Studies of the Tritium Release from Materials Under Irradiation

The WWR-K is 6-MW(thermal) light-water, tank-type reactor with thermal neutron spectrum. It is the exclusive multipurpose research reactor in the Republic of Kazakhstan. The WWR-K is owned by the Institute of Nuclear Physics of the Ministry of Energy of the Republic of Kazakhstan. The coolant is desalted water. The moderator and the reflector are desalted water and beryllium. The reactor operates on uranium dioxide that is enriched to 19.7% by 235U. The reactor is equipped with irradiation channels with the following characteristics: the thermal neutron flux density in the core center comprises 2 · 1014 cm−2s−1, whereas the fast neutron flux density (En > 0.1 MeV) comprises ~8 · 1013 cm−2s−1; in the core periphery, fluxes of the thermal and fast neutrons comprise, respectively, ~8 · 1013 and ~6 · 1012 cm−2s−1. The regular irradiation cycle length is 21 days. The annual number of cycles is ten.Since WWR-K reactor startup, the studies of various prospective reactor materials and fuels have been carried out in its core. Since 2000, activities on in-reactor tests of fusion reactor materials have been performed at the WWR-K reactor, such as experiments on tritium release out of lithium ceramics. Tests forced development and fabrication of an installation for in-reactor studies of tritium release from various candidate materials of fusion reactor blankets in the inert gas environment of an ampoule with specimens under study. Also, a technique has been developed for assessment of the time of tritium retention in materials under irradiation.In 2018, the WWR-K reactor facility was upgraded for studies of fusion reactor materials under irradiation, which makes it possible to carry out experiments on irradiation of specimens at vacuum conditions.This work presents the experimental facility description and block circuit along with its general technical capacities as applied to the expected studies of tritium release of fusion reactor materials at the WWR-K reactor. The developed irradiation ampoule device is presented schematically as well. Also, the obtained results of the neutron-physical, thermophysical, and vacuum calculations for the in-reactor experiment on irradiation of fusion reactor blanket materials are given.

Mapping S-Wave Attenuation Field of the North Tien Shan Region Using Seismogram Coda for Local Earthquakes and Quarry Blasts

We have mapped the short-period shear-wave attenuation field in the lithosphere of the North Tien Shan and analyzed seismograms of local earthquakes and quarry blasts, recorded by five stations installed in the area of Almaty and its nearest vicinities (including the site of the WWR-K nuclear reactor). Characteristics of common S coda envelopes for various sectors relative to seismic stations are described. Cross-sections of the attenuation field for the stations considered are constructed. Characteristics of the attenuation field to the north of the Zaili deep fault zone (southern margin of the Kazakh platform) and to the south of it (in mountain areas) are compared. This has allowed us to conclude about the different contents of deep-seated fluids in these areas. The maps of the attenuation field in various depth ranges and at different azimuths for all stations are made. The zones of the highest attenuation, where large earthquake can nucleate, are found. Recommendations are given regarding installation of additional stations in order to provide more operative and accurate determination of epicenter location of a possible earthquake and its magnitude for the safe operation of the WWR-K nuclear reactor.

Neutron-physical parameters at the outlet of the WWR-K reactor beam tube

Neutron-physical parameters at the outlet of the WWR-K reactor beam tube

DETERMINATION OF TOXIC ELEMENTS’ CONTENT IN THE HAIR OF THE FIRST RUSSIAN TSARINA ANASTASIA ROMANOVNA

An elemental analysis of the Tsarina Anastasia Romanovna hair fragments from the Moscow Kremlin necropolis has been carried out. The mass fractions of elements were determined by several methods: neutron activation analysis (using three facilities – the IBR-2 reactor, the IREN research facility, Joint Institute for Nuclear Research, Russia; and the WWR-K reactor, Institute of Nuclear Physics, Republic of Kazakhstan), inductively coupled plasma mass spectrometry, and atomic emission spectroscopy (Institute of Nuclear Physics, Republic of Kazakhstan). The results confirmed the hypothesis of mercury poisoning of the first Russian Tsarina Anastasia Romanovna.

THERMOPHYSICAL CALCULATIONS OF REACTOR AMPOULE FOR STUDYING OF GAS RELEASE FROM LITHIUM-CONTAINING MATERIALS WHEN CONDUCTING EXPERIMENTS ON WWR-K REACTOR

This article presents one of the stages to create a facility for long-term testing of fusion materials at the WWR-K reactor. Particularly, data are presented on thermophysical calculations of an irradiating ampoule made of stainless steelcontaining various materials of fusion breeders: lithium ceramics, lead-lithium eutectic.The 3-dimensional axisymmetric model was used for calculations, according to which all elements are on the same axis. To simulate heat and mass transfer, we used the multiphysical module Nonisothermal Flow, which integrates the Heat Transfer in Fluids and Laminar Flow submodules for calculating the heat transfer in the ampoule metal and in the air flow in contact with the ampoule surface, and for calculating the velocity and pressure distribution in the flow, respectively. The dynamics of heating and the distribution of temperature fields in the ampoule were evaluated on the basis of which conclusions were made about parameters of the irradiation experiment.

Динамика выгорания бора-10 в стержнях компенсации реактивности реактора ВВР-К

In 2015, as part of the conversion of the WWR-K reactor to low enrichment fuel, the main reactor systems were modernized, including the complete replacement of the reactor control and protection system (CPS). The energy intensity in the new, more compact core has become noticeably higher, the number and geometric parameters of the working bodies (WB) of the CPS have changed. Boron carbide with a natural isotopic composition is used as the material of the absorber in the CPS WB. CPS WB are responsible for managing the nuclear fission chain reaction of uranium. The WWR-K reactor is a stationary research reactor, so for its operation it is necessary to create an initial reactivity margin. At the beginning of the reactor operating cycle, most of the CPS rod is immersed in the core. During the operation of the reactor, the rod is gradually removed from the core due to the "poisoning" of the core and the burning of uranium. A nuclear reaction proceeds on the boron-10 isotope with absorption of a thermal neutron and the formation of an alpha particle and a lithium-7 isotope, which leads to a decrease in the nuclei of the initial isotope, i.e. burnout of boron-10. A decrease in the mass of the boron-10 isotope in the CPS rod will lead to a decrease in its physical efficiency, which directly affects the nuclear safety of the reactor. An important criterion for the operation of a nuclear reactor is the creation of subcriticality of the core for the safe shutdown of a nuclear chain reaction. The subcriticality of the core depends on the total efficiency of the reactivity compensation rods. In this work, a quantitative assessment of the burnup of boron-10 in the WB of the WWR-K reactor is presented. The effect of burnout of the boron-10 isotope on the physical efficiency of CPS rods during their long-term operation in the core of the WWR-K reactor is shown. The time intervals of 1, 3 and 10 years are considered. The obtained results can be used to justify the resource and safety of operation of the control and safety rods of the WWR-K reactor. The calculations were performed using the MNCP6 software with the library of nuclear constants ENDF/B-VII.1.

CALCULATION OF THE VACUUM DIAGRAM OF THE EXPERIMENT ON THE DEGASATION OF FUSION MATERIALS IN CONDITIONS OF NEUTRON IRRADIATION AT WWR-K REACTOR

This article presents the first stage of work on the implementation of the project for the creation of an installation for testing fusion materials at WWR-K reactor using the vacuum extraction method. In particular, the data of calculations of the vacuum system of the installation for various geometries of the vacuum path are presented. The developed model made it possible to estimate the range and distribution of pressures along the length of the irradiation device and the vacuum system. The obtained values of the pressure distribution showed that, in general, the physical ability to register the flow of tritium and helium into the system will be sufficient for the levels of production of tritium and helium in the samples under study above 10-11 mol/s.

COMPARISON OF NEUTRON SPECTRUM IN THE WWR-K REACTOR WITH LEU FUEL AGAINST HEU ONE

WWR-K is the research tank-type light-water heterogeneous reactor. Reactor operation started in 1967 with enrichment36 % in uranium-235. In 2016 reactor conversion to low-enriched uranium fuel (19.7 % in uranium-235) was implemented with the VVR-KN-type fuel assemblies (FA). In view of reactor operation, compact configuration of the core was chosen, where, following fuel burning up, side water reflector is gradually changed by beryllium one. Besides, an amount of work elements of the reactor control and protection system is increased in the new reactor core. Following results of measurement of the thermal neutron flux density, carried out in the core center in course of reactor physical startup, it has increased up to 2·1014 cm-2s-1. Thus, in terms of experimental capacities, reactor has gained after conversion in attractiveness. As is known, uranium fission is accompanied by generation of neutrons, which energies obey the Watt’s spectrum, that is, the spectrum peak is associated with the neutrons having the energy 0.7 MeV, whereas the neutron average energy comprises 2 MeV. Due to an enhanced amount of uranium-238 in the new low-enriched uranium fuel FA, neutron spectrum is making «harder». In the presented paper, neutron spectra in the core irradiation channels of re calculated by means of the computer code MCNP, and comparison of neutron spectra in the WWR-K reactor core with high-enriched uranium fuel against lowenriched fuel is conducted.

The 50th Anniversary of the WWR-K Research Reactor

The WWR-K is a multipurpose research reactor in the Republic of Kazakhstan. The WWR-K reactor was started up for the first time on October 30, 1967, under the direction of B.T. Dubovsky, V.N. Okolovicha, G.A. Batyrbekova, L.A. Yurovsky, and A.I. Maslova. The WWR-K research reactor is a water-water reactor of a heterogeneous type with a thermal neutron spectrum, a rated capacity of 10 MW, and a 36% enrichment of uranium-235. The reactor operated without any accidents until 1988. After the Chernobyl accident, the USSR Gospromatomnadzor decided to suspend the operation of the WWR-K reactor in October 1988 until the requirements for ensuring safe operation of the reactor under high seismicity conditions were met. Specialists at the Institute of Nuclear Physics endeavored to increase the seismic safety of the reactor and to provide for its safe operation in conditions of high seismicity. In 1998, the operation of the WWRK reactor was resumed with the permitted power of 6 MW. Prior to the collapse of the USSR, the main purposes of the WWR-K reactor were testing the thermionic reactor-converter elements, neutron activation analysis, studies of ultracold neutrons, studies of inert gas plasma parameters, and studies of the electroionization of CO2 and CO lasers. In 2003, a feasibility study was begun for the conversion of the reactor to lowenriched fuel with preservation of operational and experimental capabilities. As a result, a new fuel assembly, WWR-KN, was designed on its basis—a compact core, which allows improving the characteristics of the reactor. During the period from 2011 to 2013, the life-cycle tests of three test assemblies were conducted; on the basis of the test results, the WWR-KN FA was recommended for the conversion of the WWR-K reactor. In the first half of 2016, successful physical and power startups of the WWR-K reactor with low-enriched fuel were carried out. As a result of the conversion, the thermal neutron flux in the center of the core was doubled. The main current purposes of the WWR-K reactor are the following • testing fuel and structural materials of reactors of the fourth generation; • testing fusion reactor materials; • producing radioisotopes for medicine and industry; • neutron activation analysis.

Calculated and experimental studies at critical facility in view of development of a technology for neutron transmutation doping of a large size silicon specimen in WWR-K reactor

The Institute of Nuclear Physics using WWR-K research reactor were carried out studies to develop the technology of neutron-transmutation doping of silicon ingot with a diameter of 150 mm and a height of 200 mm. As a result, was obtained silicon ingot with an uneven electrical resistivity at the level of ∼6%. In 2017, together with the Chiyoda Technol Corporation, (Japan) were started the activities to develop a technology of neutron-transmutation doping of large sized silicon with a diameter of more than 150 mm and a height of up to 500 mm for the WWR-K reactor. The main goal is to achieve an irregularity electrical resistivity in terms of the volume of the silicon ingot at the level of 3-4 %. At the first stage, were performed the works on the formation of the thermal neutron field in irradiation device. The object of investigation is a silicon ingot with a diameter of 150 mm and a height of 500 mm. For reduction of the height irregularity are used neutron absorbers arranged along the height of the ingot. The development of irradiation device with a neutron absorber and the corresponding neutron measurements were performed at the critical facility. The silicon ingot is modeled by a block of the aluminium. At the second stage it is planned to perform the works on the WWR-K reactor to test the developed technology with the further irradiation of a silicon ingot.

Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions.

Microstructural defect evolution in neutron – Irradiated 12Cr18Ni9Ti stainless steel during subsequent isochronous annealing

Transmission electron microscopy and microhardness measurements were used to examine changes in microstructure and associated strengthening induced in austenitic stainless steel 12Cr18Ni9Ti irradiated to ∼0.001 and ∼5dpa in the WWR-K reactor before and after being subjected to post-irradiation isochronal annealing. The relatively low values of irradiation temperature and dpa rate (∼80°C and ∼1.2×10−8dpa/s) experienced by this steel allowed characterization of defect microstructures over a wide range of defect ensembles, all at constant composition, produced first by irradiation and then by annealing at temperatures between 450 and 1050°C. It was shown that the dispersed barrier hardening model with commonly accepted physical properties successfully predicted the observed hardening. It was also observed that when TiC precipitates form at higher annealing temperatures, the alloy does not change in hardness, reflecting a balance between precipitate-hardening and matrix-softening due to removal of solute-strengthening elements titanium and carbon. Such matrix-softening is not often considered in other studies, especially where the contribution of precipitates to hardening is a second-order effect.

Tritium migration in the materials proposed for fusion reactors: Li2TiO3 and beryllium

The results of tritium and helium gas release from lithium ceramics samples Li2TiO3 irradiated at the WWR-K reactor (Almaty, Kazakhstan) and from beryllium samples irradiated at the BN-350 reactor (Aktau, Kazakhstan) and the IVG.1M reactor (Kurchatov, Kazakhstan) are presented. Experimentally obtained thermal desorption (TDS) spectra have shown that the dependence of tritium release from lithium ceramics has a complicated behavior and to a large extent depends on lithium ceramics type. Nevertheless, it was found that the total amount of tritium released from all types of lithium ceramics has the same order of magnitude, equal to about 1011Bq/kg. It was found that in the temperature range from 523K to 1373K the process of tritium release from lithium ceramics involves volume diffusion and thermoactivated tritium release from the accumulation centers generated under irradiation. TDS of beryllium samples enables us to obtain characteristics of tritium and helium release during linear heating, to determine integrated quantities of generated helium and tritium, and to determine parameters of release processes.