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Title: Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.
Authors:
;  [1]
  1. Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)
Publication Date:
OSTI Identifier:
22307857
Resource Type:
Journal Article
Resource Relation:
Journal Name: AIP Conference Proceedings; Journal Volume: 1615; Journal Issue: 1; Conference: ICANSE 2013: 4. international conference on advances in nuclear science and engineering, Denpasar, Bali (Indonesia), 16-19 Sep 2013; Other Information: (c) 2014 AIP Publishing LLC; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COOLANTS; CRITICALITY; CROSS SECTIONS; INTERFACES; IRRADIATION; M CODES; MONTE CARLO METHOD; NUCLEAR DATA COLLECTIONS; NUCLEAR FUELS; PROBABILITY DENSITY FUNCTIONS; PWR TYPE REACTORS; URANIUM DIOXIDE