Description
Chairs: Emil Fridman and Xingkai Huo
The application of Generation IV reactors offers improved sustainability of nuclear energy production and extends the current reserves while it helps to reduce the amount of nuclear waste. Therefore multiple demonstrator reactor designs are now under development. In fast reactors, due to its characteristic spectra and the reactor design, the effects of the leakage is much higher, which suggest...
The results of Phase-1 neutronics benchmark analysis of CEFR start-up tests was presented in the 2nd Research Coordination Meeting (RCM-2), held at Beijing, China during 28th October to 1st November 2019. Participation of this benchmark exercise has provided us a wide international forum to inter-compare the method, computer codes and cross section data employed at IGCAR for the physics design...
AZNHEX is a deterministic code that solves the neutron diffusion equation with Hexagonal-Z geometry. It is part of the AZTLAN Platform, an initiative gathering the leading Mexican universities and nuclear energy research centers aimed to position Mexico in the midterm as a nuclear code developer for reactor core analysis and design. To tackle the challenges of a small core, like CEFR, with...
The OECD/NEA Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) was launched in 2015 to study reactivity feedback coefficients and their uncertainties for a medium-sized 1,000 MWth metallic core and a large 3,600 MWth oxide core. In addition to investigations of the full core level, stand-alone multi-scale...
During the last decades, the development of new powerful computers and high performance analytical tools, along with the reduction of the approximations due to new methods implemented in the algorithms for the solution of the transport equation, pushed nuclear cross-sections data as the main source of uncertainty in neutronic calculations. This points out the importance in quantifying nuclear...
ESFR-SMART belongs to the family of Gen-IV sodium cooled reactors. For its safety performance demonstration the detailed knowledge of selected safety related parameters is required. It is very important to know not only their spatial distribution and amplitude but also their mutual interdependence. In this paper the sodium void effect, the Doppler effect and fuel and cladding density effects...
The knowledge about fuel nuclear concentration behavior, fission product etc. both in fast reactor core and after withdrawing fuel from reactor is necessary from the point of view of nuclear safety as well as in future for fuel cycle closing up. There are a lot of methods and calculation codes in practice of burnup equation solution. Three different approaches can be highlighted, among which:...
To date, French nuclear power plants mainly use uranium extracted from the mines and enriched. The used fuel is reprocessed to extract useful materials, such as plutonium, which is recycled once in dedicated PWRs of the park. Once used, the quality of the plutonium decreases and it becomes more difficult to recycle this plutonium in thermal spectrum reactors.
A molten salt reactor (MSR)...
