Band 85, Heft 5

Turbulent heat transfer through gas-cooled systems is a key factor for the improvements of the gas-cooled fast reactors. Within the FP7 European projects GoFastR (European Gas Cooled Fast Reactor) [1] and THINS (Thermal-hydraulics of Innovative Nuclear Systems) [2], numerical tools for the simulation of the thermal-hydraulics of next generation reactor systems were developed, applied and validated for innovative coolants. One of their objects was the L-STAR (Luft - STab, Abstandshalter, Rauheiten) facility, which has been designed and erected at the Karlsruhe Institute of Technology to study turbulent flow behaviour and its heat transfer enhancement characteristics in gas-cooled annular channels under a wide range of conditions. The test section consists of a hexagonal cross-section channel with an inner electrical rod heater placed coaxially. This design represents the flow domain around a single fuel rod in a future GFR. In this paper experimental results of the fluid flow with uniform heat release conditions for the smooth heater rod are considered. Design of the facility allows using various values of heating power and fluid mass flow rates depending on the chosen scenario. The purpose of this study is to create CFD models for the L-STAR facility, which are capable of reproducing the results of reference measurements. The other aim is to find an applicable turbulence model for the GFR relevant cases. Steady-state simulations are performed for selected cases. We also present mesh, structural element and turbulence model sensitivity studies during this work. With these model results it becomes possible to validate our CFD models to be applicable for the ALLEGRO gas-cooled reactor fuel assemblies.

This work deals with the updated reference solution of the Full-Core VVER-440 PK-3+ benchmark which is based on the extended calculation benchmark from 2018. [1] The modification consists of a new type of fuel assembly with PK-3 design in the loading pattern. The main aim of this calculation is to create a reference solution to test the power distribution predicted by macro-code. The solution has been calculated by the transport code Serpent, code using Monte Carlo method and including all possible levels of power distribution in a 2-D problem. The results consist of the effective multiplication factor (K eff ) and the power distribution on fuel assembly (Kq) and the pin-by-pin (Kq) level. The paper also includes a detailed evaluation of the precision of Monte Carlo solution and a study of the influence of nuclear data library suitable for comparison and validation.

This paper presents a validation of multiphase computational fluid dynamics (CFD) code Ansys CFX 18 on selected experiments with the subcooled boiling flow. Ansys CFX code can simulate multiphase flow by solving three balance equations for each phase (Euler-Euler approach). Boiling on the heated wall can be simulated with the model proposed by Kurul and Podowski. A set of mathematical models of physical phenomena in boiling bubbly flow was selected and tested on the following experiments with subcooled boiling: DEBORA experiments, ASU experiments, FRIGG FT-6a experiment, and Bartolomej et al. experiment. It was found out that CFX code can reasonably reproduce measured data in all these experiments. At the end of this paper, we present an application of this modeling approach to the real case with a complicated geometry: simulation of subcooled boiling in the VVER-1000 fuel assembly with bent rods

One of the most important issues in nuclear reactor operation and its designing is considering the interaction between thermal hydraulics and neutronics physics because there is an important relationship between the states of fluid and neutron spectrum and power distribution. In this research, the MCNP4C and COBRA-EN nuclear codes were coupled with each other to precisely analyze the fuel assembly of the light water reactor core. This coupling was carried out using iterative processes between the linked neutronic and thermal-hydraulic codes applying successive procedures while the desired convergence was made in both. The newly designed code was checked for three test problems, and the obtained results showed the improvement of the computations procedures by the developed code.

A dynamic radioecological model was developed to estimate radiation doses and stochastic health risks due to atmospheric releases in case of a nuclear reactor accident. Activities in some foodstuffs, feedstuffs and grass, and doses from different pathways for age groups and stochastic risks can be calculated by the model. Dynamic features include food harvesting, sowing times, feeding regimes, and the growing up of a person. The model can be coupled to any atmospheric dispersion model which can calculate radionuclide concentrations or to any measurement data at predetermined time intervals. Model features are based on ECOSYS-87 model with some modifications. Since the Chernobyl accident, there had been an increase in real world data to assess capabilities of software developed to calculate radionuclide concentrations in environment. In this study, the developed software was validated against measured concentrations and estimated individual doses for 137 Cs following the accident at Chernobyl in 1986, presented in International Atomic Energy Agency Scenario S. Results obtained using the model compare well with dose estimates and activities in the environment.

Uncertainty and sensitivity analysis were conducted for a radioecological model which was developed and applied for severe accident scenario for Akkuyu NPP in Turkey. Uncertainty was investigated in the doses and activity concentrations. Sensitivity analysis was conducted to find out most influencing parameters on the doses. 96 parameters, related with foodstuffs, feedstuffs and grass, inhalation rate and reduction factor for external irradiation, and transfer of 137 Cs, 134 Cs and 131 I isotopes, were studied by probabilistic methods. Latin Hypercube Sampling method was used to generate random parameters. In sensitivity analysis Spearman rank coefficients between the parameters and doses were calculated. The uncertainty analysis proves that the lifetime doses are in the range of 4.8 - 32 mSv and short-term doses are in the range of 1.1 - 2.8 mSv when all parameters vary simultaneously. Sensitivity analysis proves that soil-plant and feed-cow milk transfer factors for cesium, beef and grain consumption, and reduction factor for external irradiation have more influence on the lifetime doses, and transfer factor for cow milk for iodine and interception factor for grass also affect the short-term doses.

The effect for different types of scattering on the critical half thickness in slab geometry for one speed neutron transport equation is studied for isotropic, linearly anisotropic and quadratic anisotropic scatterings. An extensive numerical survey is carried out for the critical thickness in order to provide the effect of the different scattering types. The numerical results are obtained by P N , T N and U N methods. The P N method is the Legendre polynomial solution that is accepted as the exact result for the neutron transport theory calculations and the U N and T N methods are the types of Chebyshev polynomials. Critical thickness values are calculated by using Mark boundary condition. Results are compared with the literature

The paper presents results of certification of the ATHLET/ BIPR-VVER software system (version 1.0) in the Russian Federation. The following stages of the software system certification are described: preparation of a validation report; work with appointed experts; performance of additional studies to increase the level of the software system validation; and preparation of a software system certificate. Examples of validation computations simulating experiments in operating power units and thermal-hydraulic experimental facilities are given. The ATHLET/BIPR-VVER software system (version 1.0) was certified by the Expert Council on software certification under Rostechnadzor on 24.10.2018 (Certificate No. 455).

This paper discusses potential safety challenges to continued operation of nuclear installations during pandemic lockdown. While operating organizations are still learning the lessons, experience with the Corona virus outbreak has shown that continuity of operation requires an effective business continuity plan ensuring that adequate safety measures, with due account for all potential challenge, are maintained as the highest priority. These challenges may affect various organizational and technical areas related to operational safety, such as leadership and management for safety, staffing arrangements, training and qualification of personnel, maintenance, testing and inspection, supply chain, and operational radiation protection and radioactive waste management. The pandemic could also affect the effectiveness of regulatory oversight and emergency preparedness. Analysis of these challenges and suggested measures to address them, along with elements to be covered by the business continuity plan are presented and discussed in this paper.

In a medical linear accelerator, the primary and secondary collimators are generally made of high atomic weight metals. The energy of the x-rays generated by accelerated electrons exceeds the bonding energies of high atomic number materials such as tungsten, and lead in the structure of the components in the accelerator head, it causes the production of neutron from the nucleus. Consequently, the resulting photo-neutrons can easily be scattered and spread into the treatment room. Thus, an additional neutron dose cannot be ignored in the tumor volume of the patient and in areas outside this volume. The contribution of the primary and secondary collimators to the neutron contamination is found to be approximately 52% and 30%, respectively. In this paper, the effects of using different composite materials such as Boron, Cadmium, and Polyethylene in secondary collimators on neutron contamination have been investigated by using Fluka Monte Carlo and TALYS 1.9 codes