skip to main content

Title: The new MCNP6 depletion capability

The first MCNP based in-line Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerfulmore » capability as well as dictate a path forward for future development to improve the usefulness of the technology. (authors)« less
Authors:
; ;  [1] ;  [2]
  1. D-5, Los Alamos, NM 87545 (United States)
  2. XCP-3 MCNP Code Development Project, MS C921, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)
Publication Date:
OSTI Identifier:
22107764
Resource Type:
Conference
Resource Relation:
Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 36 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
Publisher:
American Nuclear Society - ANS; La Grange Park (United States)
Research Org:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 97 MATHEMATICAL METHODS AND COMPUTING; BURNUP; COMPUTERIZED SIMULATION; GEOMETRY; MEMORY MANAGEMENT; MONTE CARLO METHOD; NUCLEAR POWER PLANTS; RADIANT HEAT TRANSFER; RADIATION PROTECTION; RADIATION TRANSPORT; REACTOR CORES