Submission Title
Estimation of minimum critical dimensions of process vessels handling Pu Nitrate Solution Using OpenMC
Presentation Type
Contributed
Start Date
20-12-2018 9:00 AM
Abstract
The exact values of minimum critical values (MCV) for Pu nitrate and Mixed U-Pu nitrate solutions are needed during the design and safe operations of any Fast Reactor Fuel Cycle Facilities. Although in open literatures and many reactor physics handbook MCV values are reported, but parameters pertaining to exact composition of fuel used is scarce. In this study, efforts are made to calculate and validate the scheme for evaluating the MCV values for Pu nitrate solutions. The criticality calculations were carried out using the OpenMC code with the continuous-energy cross sections ENDF/B-VII.1 obtained from NNDC. The Pu nitrate solutions is assumed to be pure aqueous, i.e., without acidity [H+=0] and is fully surrounded by 30 cm of water reflection.
Existing criticality analysis codes iterate over the desired parameters until a desired neutron multiplication factor keff is reached and one has to manually sweep the parameters to arrive at the MCV, which is very tedious and also prone to errors. In order to circumvent these deficiencies, an automated process for the evaluation of MCV is developed using OpenMC and Python scripts. In this paper we’ll discuss about the use of OpenMC for criticality searches in evaluating the MCV for a critical system and validate our finding with benchmark values.
Recommended Citation
Rawat, Ajay; Chandrasekaran, S.; Baskaran, R.; and Venkatraman, B. (2018). "Estimation of minimum critical dimensions of process vessels handling Pu Nitrate Solution Using OpenMC," Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety.
Estimation of minimum critical dimensions of process vessels handling Pu Nitrate Solution Using OpenMC
The exact values of minimum critical values (MCV) for Pu nitrate and Mixed U-Pu nitrate solutions are needed during the design and safe operations of any Fast Reactor Fuel Cycle Facilities. Although in open literatures and many reactor physics handbook MCV values are reported, but parameters pertaining to exact composition of fuel used is scarce. In this study, efforts are made to calculate and validate the scheme for evaluating the MCV values for Pu nitrate solutions. The criticality calculations were carried out using the OpenMC code with the continuous-energy cross sections ENDF/B-VII.1 obtained from NNDC. The Pu nitrate solutions is assumed to be pure aqueous, i.e., without acidity [H+=0] and is fully surrounded by 30 cm of water reflection.
Existing criticality analysis codes iterate over the desired parameters until a desired neutron multiplication factor keff is reached and one has to manually sweep the parameters to arrive at the MCV, which is very tedious and also prone to errors. In order to circumvent these deficiencies, an automated process for the evaluation of MCV is developed using OpenMC and Python scripts. In this paper we’ll discuss about the use of OpenMC for criticality searches in evaluating the MCV for a critical system and validate our finding with benchmark values.
