Effect of the burnable absorber arrangement on the VVER-1200 fuel assembly neutronic performance

Optimizing the use of fuel in a power reactor is a task of current concern. However, little attention has been given to investigating the dependences among the enrichment used, the content of gadolinium oxide in fuel elements, and the life time in combination with assessing the efficiency of using Gd fuel elements with different Gd2O3 contents. The paper considers fuel assembly (FA) versions for VVER-1200 reactors having different enrichments for fuel elements, including those with Gd, and different contents of gadolinium oxide in fuel. A comparative analysis is presented for assemblies with homogeneous Gd2O3 arrangements in each fuel element and with profiled Gd2O3 arrangements. In the latter case, profiling depends on the neutron flux density in the layer which includes Gd fuel elements. This suggests that the arrangement of gadolinium oxide proportionally to the neutron flux density will improve the FA neutronic performance. The results were obtained using SERPENT (a continuous-energy multi-purpose three-dimensional Monte Carlo particle transport code). The assemblies with the used parameters for a 12-month fuel cycle have shown the method under consideration to be inefficient for a period of over 300 eff. days. With increased enrichment and content of gadolinium oxide, the use of profiled versions has turned out to be more rational for longer periods (up to 900 eff. days). Therefore, this phenomenon is relevant for the reactor life, whereas it proves to be insignificant for the fuel life. A complex relationship is noted between the gadolinium and uranium content in an assembly and the effective multiplication factor for the profiled and standard assemblies. This relationship requires further detailed consideration.

Structural Changes in Concrete under the Influence of Reactor Spectrum Neutrons

The paper considers structural changes in the concrete composition that occur under the influence of neutrons of the reactor spectrum, using the example of the IR-100 research nuclear reactor, taking into account its real time and operating conditions. Thus, taking into account the energy output, power operation modes, and neutron flux density in the core, over time, nuclides that are not characteristic of the original composition of the concrete component are formed in the nodes of the crystal lattice. However, these changes do not lead to significant structural changes.

Production and Monitoring of Neutron Flux by Activation Detectors

The neutron generation technique was tested on the microtron M-10 with an output electron beam of 8.7 MeV. Given the low energy that the microtron can provide to electrons, the bremsstrahlung induced photonuclear reaction 9Be (γ, n), which has a low threshold, was chosen for neutron generation. Cobalt and indium targets were tested as activation detectors to estimate the neutron flux density. In the cobalt target, the isomeric state of 60mCo with an energy of 58.6 keV and a half-life of 10.5 minutes is well activated. Two well-known additional gamma lines of standard cobalt source permit to clarify the absolute value of the neutron flux. The activated indium target has four gamma lines bound to the 116mIn isomer β- decaying with the half-life of 54.4 minutes, what is convenient for measurement of gamma spectrum. Despite the low energy of the output electron beam, at a beam intensity of 5 μA it is possible to obtain an almost isotropic neutron flux of 107 n/(s∙cm2).

Modeling of neutron flux density anomalies in the automated system of «Shelter» Object’s nu-clear safety control.

Analysis of the dynamics of neutron flux density (NFD) from fuel-containing masses (FCM) in the «Shelter» Object shows the presence of values that exceed the average values for different observation periods. Identification of such values by the criterion of «anomaly / non-anomaly» will allow excluding uninformative events from the array of observations. Or, in the case of the anomaly confirmation, it will allow forming effective actions for decision-making in order to eliminate the consequences of such events. To solve the problem of detecting anomalous measurements, now there is utilized the theory of statistical solutions which is based on the use of parametric methods. The utilization of these methods requires a priori information about the nature of the distribution of the measured process and its parameters. In order to find an effec-tive solution to the problem of detecting and eliminating anomalous measurements, it is necessary to know the statistical characteristics of normal and anomalous components. This paper proposes statistical criteria for estimating anomalies in time series of NFD which have different approaches to the formation of observation intervals, power and reliability of anomaly detection. Depending on the type of distribution of the array of observations, a set of criteria is proposed. These criteria are most expedient to use when checking the anomaly of series levels for the exponential distribution, the Poisson and Weibull distribution. The capacity of the criteria has been evaluated depending on the sample size. The article defines the accuracy of determination of anomalies by criteria with the known values of the mean and dispersion in the studied sample. As the result, it is recommended to use the Grubbs test to study the anomaly of the sample levels with n>700, and the Dixon and Smolyak-Titarenko criteria for the samples with n<50. The utilization of optimal criteria depending on the characteristics of the studied samples will increase the mathematical significance of the obtained results and, as a result, will improve the quality of management decisions and nuclear safety on the «Shelter» Object as a whole.

Uncertainty Analysis of WWER-1000 Core Macroscopic Cross Sections due to Spectral Effects

The paper presents the results of uncertainty analysis of WWER‑1000 core macroscopic cross sections due to spectral effects during WWER‑1000 fuel burnup and the analysis of cross section sensitivity from thermophysical parameters of the calculated cell, which affect energy spectrum of neutron flux density. The calculation of changes in the isotopic composition during burnup and the preparation of macroscopic cross sections used the developed HELIOS computer model [1] for TVSA, which is currently operated at most Ukrainian WWER‑1000 units. The GRS approach applying Software for Uncertainty and Sensitivity Analyses (SUSA) [2] was chosen to assess the uncertainty of the macroscopic cross sections due to spectral effects and analysis of cross section sensitivity from thermophysical parameters. The spectral effect on macroscopic cross sections was taken into account by calculating the fuel burnup for variational sets of thermophysical parameters (fuel temperature, coolant temperature and density, boric acid concentration) prepared in advance by the SUSA program, as a result of which fuel isotopic composition vectors were obtained. After that, neutronic constants for the reference state were developed for each of the sets of isotopic composition, which corresponded to a certain set of thermophysical parameters. At the next stage, the uncertainty of macroscopic cross sections of the interaction due to the spectral effects on the isotopic composition of the fuel was analyzed using SUSA 4, followed by the analysis of cross section sensitivity from thermophysical parameters of the calculated cell affecting energy spectrum of neutron flux density. In the future, the uncertainty of two-group macroscopic diffusion constants can be used to estimate the overall uncertainty of neutronic characteristics in large-grid core calculations, in particular, in the safety analysis.

RESEARCH OF MODELS OF NEW PLATE HEAT-RELEASING ELEMENT

The main parameter of research reactors is the neutron flux density. To obtain high neutron fluxes, the research reactor must be compact and the reactor power must be maximized. Nuclear fuel plays the main role in high-flow research reactors. Nuclear fuel using UO2is limited by the density of uranium in fuel elements (FUEL ELEMENTS) 3 g / sm3

MULTI-GROUP MODELING OF PROTECTION AGAINST NEUTRON AND GAMMA RADIATION BY MATERIALS BASED ON TITANIUM HYDRIDE

The paper provides a comparative calculation of the radiation protective efficiency of various composite materials based on titanium hydride using multi-group modeling methods using the ANISN program. The calculations showed the high efficiency of titanium hydride composites with respect to neutron and gamma radiation. The relaxation length of the fast neutron flux density in titanium hydride materials is 5.1…7.0 cm. The spatial-energy distribution of neutron radiation in materials is formed by fast neutrons. The dose rate of gamma rays behind the material is determined mainly by capturing gamma rays arising in the initial layer of protection. Introduction to the composition of the protection of boron atoms reduces the level of capture gamma radiation, but does not affect the attenuation of fast neutrons.

EFFECT OF NACL ON NEUTRON FLUX DENSITY IN GRAPHITE BLOCK

Aim of this work is to study the effect of NaCl on neutron flux through graphite block. Graphite block had 39 channels drilled through it and filled with NaCl. Au and In foils and In wire was used for neutron activation measurement of the neutron flux. AmBe neutron source with nominal activity of 92.5 GBq was used. The results were compared with MCNP simulations with channels filled with salt and air. Twelve different locations for Au and In foils were used and In wires were place in three locations to measure vertical profile of neutron flux.