Validation of the Serpent Monte Carlo burnup code for the Takahama-3 benchmark experiment
Kaznacheevskaia, Tatiana (2020)
Diplomityö
2020
School of Energy Systems, Energiatekniikka
Kaikki oikeudet pidätetään.
Julkaisun pysyvä osoite on
https://urn.fi/URN:NBN:fi-fe2020120499522
https://urn.fi/URN:NBN:fi-fe2020120499522
Tiivistelmä
A burnup calculation is a part of nuclear criticality safety analysis utilizing burnup credit. In many cases, it is not possible to realise a real experiment with fuel burnup. The solution to this problem is using reactor physics codes that have been validated for these purposes.
In this thesis validation of the Serpent Monte Carlo burnup code is presented. Serpent is a Monte Carlo physics code developed by VTT Technical Research Centre of Finland. The Japanese reactor Takahama-3 was chosen as a benchmark experiment. Such a benchmark can be used for validating and verification of different package of codes because it is one of the few samples that publicly available. Results are compared with experimentally measured nuclide data that are available in the SFCOMPO database.
In the current research results of 3D fuel assembly burnup calculation are presented. Each fuel assembly calculation was performed using two different burnup schemes: traditional predictor-corrector with linear extrapolation/linear interpolation and Stochastic Implicit Euler. Calculations of fuel assemblies have been done with approximations in boron concentration and axial dimensions of fuel assemblies from similar types of reactors. At this stage, the results of the calculations are satisfactory for some nuclides that already used in nuclear criticality safety analysis taking into account the approximations in calculations and simplifications of the models. For some nuclides that errors exceed the maximum permissible value, calculations should be clarified in further research.
In this thesis validation of the Serpent Monte Carlo burnup code is presented. Serpent is a Monte Carlo physics code developed by VTT Technical Research Centre of Finland. The Japanese reactor Takahama-3 was chosen as a benchmark experiment. Such a benchmark can be used for validating and verification of different package of codes because it is one of the few samples that publicly available. Results are compared with experimentally measured nuclide data that are available in the SFCOMPO database.
In the current research results of 3D fuel assembly burnup calculation are presented. Each fuel assembly calculation was performed using two different burnup schemes: traditional predictor-corrector with linear extrapolation/linear interpolation and Stochastic Implicit Euler. Calculations of fuel assemblies have been done with approximations in boron concentration and axial dimensions of fuel assemblies from similar types of reactors. At this stage, the results of the calculations are satisfactory for some nuclides that already used in nuclear criticality safety analysis taking into account the approximations in calculations and simplifications of the models. For some nuclides that errors exceed the maximum permissible value, calculations should be clarified in further research.