MCNP User Manual, Version 5

1 2/1/2008iTitle:MCNP A General monte CarloN-Particle Transport Code, Version 5 Volume I: Overview and TheoryAuthors:X-5 monte Carlo TeamLA-UR-03-1987 Approved for public release;distribution is unlimitedLos Alamos National Laboratory, an affirmative action/equal opportunity employer, is operated by the Los Alamos National Security, LLCfor the National Nuclear Security Administration of the Department of Energy under contract DE-AC52-06NA25396. By acceptance ofthis article, the publisher recognizes that the Government retains a nonexclusive, royalty-free license to publish or reproduce thepublished form of this contribution, or to allow others to do so, for Government purposes.

2 Los Alamos National Laboratory requeststhat the publisher identify this article as work performed under the auspices of the Department of Energy. Los Alamos NationalLaboratory strongly supports academic freedom and a researcher s right to publish; as an institution, however, the Laboratory does notendorse the viewpoint of a publication or guarantee its technical 836 (7/06)X-5 monte Carlo TeamApril 24, 2003(Revised 2/1/2008)ii2/1/2008 MCNP,MCNP5, and MCNP Version 5 are trademarks of Los Alamos National NOTICE & DISCLAIMER Copyright Los Alamos National Security, Software was produced under a Government contract DE-AC52-06NA25396) by Los Alamos NationalLaboratory, which is operated by Los Alamos National Security, LLC (LANS) for the Department of Energy,National Nuclear Security Administration.

3 Los Alamos National Security, LLC (LANS) has certain rights in theprogram pursuant to the contract and the program should not be copied or distributed outside your organization. Allrights in the program are reserved by the DoE and Los Alamos National Security, LLC (LANS). Neither theGovernment nor LANS makes any warranty, express or implied, or assumes any liability or responsibility for the userof this manual is a practical guide for the use of the general-purpose monte Carlo code MCNP. The previous Version ofthe manual (LA-13709-M, March 2000) has been corrected and updated to include the new features found in MCNPV ersion 5 (MCNP5).

4 The manual has also been split into 3 volumes:Volume I:MCNP Overview and TheoryChapters 1, 2 and Appendices G, HVolume II:MCNP User s GuideChapters 1, 3, 4, 5 and Appendices A, B, I, J, KVolume III:MCNP Developer s GuideAppendices C, D, E, FVolume I (LA-UR-03-1987) provides an overview of the capabilities of MCNP5 and a detailed discussion of thetheoretical basis for the code. The first chapter provides introductory information about MCNP5. The second chapterdescribes the mathematics, data, physics, and monte Carlo simulation techniques which form the basis for discussion is not meant to be exhaustive details of some techniques and of the monte Carlo method itself arecovered by references to the II (LA-CP-03-0245) provides detailed specifications for MCNP5 input and options, numerous exampleproblems, and a discussion of the output generated by MCNP5.

5 The first chapter is a primer on basic MCNP5 third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, andthe fifth chapter explains the output. The appendices provide information on the available data libraries for MCNP, theformat for several input/output files, and plotting the geometry, tallies, and III (LA-CP-03-0284) provides details on how to install MCNP on various computer systems, how to modifythe code, the meaning of some of the code variables, and data layouts for certain monte Carlo method for solving transport problems emerged from work done at Los Alamos during WorldWar II.

6 The method is generally attributed to Fermi, von Neumann, Ulam, Metropolis, and Richtmyer. MCNP, firstreleased in 1977, is the successor to their work and has been under continuous development for the past 30 the code nor the manual is static. The code is changed as needs arise, and the manual is changed to reflect thelatest Version of the code. This particular manual refers to Version and this manual are the product of the combined effort of many people in the monte Carlo Codes (X-3-MCC,formerly part of the X-5 group) section in the Applied Physics Division (X Division) at the Los Alamos NationalLaboratory:X-3 monte Carlo CodesThomas E.

7 BoothJohn T. GoorleyAvneet SoodForrest B. BrownH. Grady HughesJeremy E. SweezyJeffrey S. BullRoger MartzAnthony ZukaitisR. Arthur ForsterRichard E. PraelX-1 Data TeamRobert C. LittleMorgan C. WhiteMary Beth LeeHolley Trellue (D-5)Technical EditorSheila M. GirardThe code and manual can be obtained from the Radiation Safety Information Computational Center (RSICC),P. O. Box 2008, Oak Ridge, TN, E. SweezyX-3-MCC Deputy Group A General monte Carlo N-Particle Transport CodeVersion 5X-3 monte Carlo CodesApplied Physics DivisionLos Alamos National LaboratoryABSTRACTMCNP is a general-purposeMonteCarloN Particle code that can be used for neutron, photon,electron, or coupled neutron/photon/electron transport, including the capability to calculateeigenvalues for critical systems.

8 The code treats an arbitrary three-dimensional configuration ofmaterials in geometric cells bounded by first- and second-degree surfaces and fourth-degreeelliptical cross-section data are used. For neutrons, all reactions given in a particular cross-sectionevaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both thefree gas and S( , ) models. For photons, the code accounts for incoherent and coherent scattering,the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production. Electron/positron transport processes account for angular deflectionthrough multiple Coulomb scattering, collisional energy loss with optional straggling, and theproduction of secondary particles including K x-rays, knock-on and Auger electrons,bremsstrahlung, and annihilation gamma rays from positron annihilation at rest.

9 Electron transportdoes not include the effects of external or self-induced electromagnetic fields. Photonuclearphysics is available for a limited number of standard features that make MCNP very versatile and easy to use include a powerfulgeneral source, criticality source, and surface source; both geometry and output tally plotters; a richcollection of variance reduction techniques; a flexible tally structure; and an extensive collectionof cross-section of ContentsVolume I: Overview and TheoryCHAPTER 1 - MCNP OVERVIEW ..1 MCNP AND THE monte CARLO METHOD ..1 monte Carlo Method vs. Deterministic Method ..2 The monte Carlo Method ..2 INTRODUCTION TO MCNP FEATURES.

10 4 Nuclear Data and Reactions ..4 Source Specification ..5 Tallies and Output ..5 Estimation of monte Carlo Errors ..6 Variance Reduction ..8 MCNP GEOMETRY ..12 Cells ..13 Surface Type Specification ..17 Surface Parameter Specification ..17 REFERENCES ..19 CHAPTER 2 - GEOMETRY, DATA, PHYSICS, AND MATHEMATICS ..1 INTRODUCTION ..1 History ..1 MCNP Structure ..4 History Flow ..5 GEOMETRY ..7 Complement Operator ..8 Repeated Structure Geometry ..9 Surfaces ..9 CROSS SECTIONS ..14 Neutron Interaction Data: Continuous-Energy and Discrete-Reaction ..16 Photon Interaction Data ..20 Electron Interaction Data ..23 Neutron Dosimetry Cross Sections.