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Title: Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

Abstract

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combatmore » this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

Authors:
 [1];  [1];  [1]
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  1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1221218
Report Number(s):
LA-UR-15-20309
Journal ID: ISSN 0149-1970; PII: S0149197015000827
Grant/Contract Number:  
AC52-06NA25396
Resource Type:
Accepted Manuscript
Journal Name:
Progress in Nuclear Energy
Additional Journal Information:
Journal Volume: 83; Journal Issue: C; Journal ID: ISSN 0149-1970
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
42 ENGINEERING; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; MCNP6, Burnup, Monte Carlo Linked Burnup

Citation Formats

Fensin, M. L., Galloway, J. D., and James, M. R. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations. United States: N. p., 2015. Web. doi:10.1016/j.pnucene.2015.03.017.
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Fensin, M. L., Galloway, J. D., & James, M. R. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations. United States. https://doi.org/10.1016/j.pnucene.2015.03.017
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Fensin, M. L., Galloway, J. D., and James, M. R. Sat . "Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations". United States. https://doi.org/10.1016/j.pnucene.2015.03.017. https://www.osti.gov/servlets/purl/1221218.
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@article{osti_1221218,
title = {Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations},
author = {Fensin, M. L. and Galloway, J. D. and James, M. R.},
abstractNote = {The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.},
doi = {10.1016/j.pnucene.2015.03.017},
journal = {Progress in Nuclear Energy},
number = C,
volume = 83,
place = {United States},
year = {Sat Apr 11 00:00:00 EDT 2015},
month = {Sat Apr 11 00:00:00 EDT 2015}
}
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https://doi.org/10.1016/j.pnucene.2015.03.017
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Works referenced in this record:

Improved Reaction Rate Tracking and Fission Product Yield Determinations for the Monte Carlo-Linked Depletion Capability in MCNPX
journal, October 2008

  • Fensin, Michael L.; Hendricks, John S.; Anghaie, Samim
  • Nuclear Technology, Vol. 164, Issue 1
  • DOI: 10.13182/NT08-A4003

The Enhancements and Testing for the MCNPX 2.6.0 Depletion Capability
journal, April 2010

  • Fensin, Michael L.; Hendricks, John S.; Anghaie, Samim
  • Nuclear Technology, Vol. 170, Issue 1
  • DOI: 10.13182/NT10-2

Efficient Generation of One-Group Cross Sections for Coupled Monte Carlo Depletion Calculations
journal, May 2008

  • Fridman, E.; Shwageraus, E.; Galperin, A.
  • Nuclear Science and Engineering, Vol. 159, Issue 1
  • DOI: 10.13182/NSE07-34

An Optimum Approach to Monte Carlo Burnup
journal, June 2007

  • Haeck, W.; Verboomen, B.
  • Nuclear Science and Engineering, Vol. 156, Issue 2
  • DOI: 10.13182/NSE07-A2695

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system
journal, September 2011

Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment
journal, February 2000

  • Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
  • Journal of Nuclear Science and Technology, Vol. 37, Issue 2
  • DOI: 10.1080/18811248.2000.9714876

Development of the point-depletion code DEPTH
journal, May 2013

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    Works referencing / citing this record:

    Computation of the neutron multiplicity moments for research reactor fuels using MCNP6 and SOURCES4c
    journal, August 2019

    • Dim, Odera U.; Aghara, Sukesh K.
    • International Journal of Energy Research
    • DOI: 10.1002/er.4814
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