tailieunhanh - Sensitivity and uncertainty analysis of beff for MYRRHA using a Monte Carlo technique

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction beff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. | Sensitivity and uncertainty analysis of beff for MYRRHA using a Monte Carlo technique EPJ Nuclear Sci. Technol. 4 42 2018 Nuclear Sciences H. Iwamoto et al. published by EDP Sciences 2018 amp Technologies https epjn 2018023 Available online at https REGULAR ARTICLE Sensitivity and uncertainty analysis of beff for MYRRHA using a Monte Carlo technique Hiroki Iwamoto1 2 Alexey Stakovskiy1 Luca Fiorito1 3 and Gert Van den Eynde1 1 Institute for Advanced Nuclear Systems Belgian Nuclear Research Centre SCK CEN Boeretang 200 2400 Mol Belgium 2 J-PARC Center Japan Atomic Energy Agency JAEA 2-4 Shirakata Tokai-mura Naka-gun Ibaraki 319-1195 Japan 3 Data Bank OECD Nuclear Energy Agency NEA 46 Quai Alphonse Le Gallo 92100 Boulogne-Billancourt France Received 3 October 2017 Received in final form 26 January 2018 Accepted 14 May 2018 Abstract. This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction beff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The beff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the beff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a 20 is the most suitable for the uncertainty quantification of beff. Using the calculated beff sensitivities and the covariance data the beff uncertainties for the critical and subcritical cores are determined to be and respectively which are dominated by delayed neutron yield of 239Pu and 238U. 1 Introduction its nuclear data sensitivities and kinetic parameters including beff. Although these codes have no capability To promote research and development of nuclear technol- to directly calculate the beff sensitivities owing to technical ogy for various applications such as accelerator-driven cumbersomeness it is approximately expressed as

• Năng lượng
• Nuclear data sensitivity
• Delayed neutron fraction beff
• The MYRRHA reacto
• Continuous energy Monte Carlo N Particle
• Transport code MCNP
• Nuclear data uncertainty and sensitivity analysis with xsusa
• Fuel assembly depletion calculations
• UAM LWR Benchmark
• Discrete ordinates code NEWT
• The SCALE 6
• 1 code system
• Gadolinium cross sections evaluations
• Gadolinium fuel pins
• Sensitivity and uncertainty analysis
• LWR fuel assemblies
• Nuclear cross sections data on
• NEA Nuclear Data Sensitivity Tool
• Half Monte Carlo method
• Uncertainty propagation techniques
• Total Monte Carlo sampling process
• Monte Carlo simulation
• Nuclear energy
• Proper orthogonal decomposition
• Data driven
• Molten salt fast reactor
• Uncertainty and sensitivity analysis
• Transient analysis