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Criticality calculations with MCNP(trademark): A primer.

DE94016571

Publication Date	1994
Personal Author	Harmon, C. D.; Busch, R. D.; Briesmeister, J. F.; Forster, R. A.
Page Count	174
Abstract	With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.
Keywords	Criticality Alpha Particles Beta Particles Configuration Geometry M Codes Moderators Radiation Transport EDB/420203 EDB/990200
Source Agency	Technical Information Center Oak Ridge Tennessee
NTIS Subject Category	77 - Nuclear Science & Technology 77F - Radiation Shielding, Protection, & Safety
Corporate Authors	Los Alamos National Lab., NM.; Department of Energy, Washington, DC.; New Mexico Univ., Albuquerque, NM (United States).
Supplemental Notes	Sponsored by Department of Energy, Washington, DC.
Document Type	Technical Report
NTIS Issue Number	199503
Contract Number	W-7405-ENG-36

Criticality calculations with MCNP(trademark): A primer.

Criticality calculations with MCNP(trademark): A primer.
DE94016571

Publication Date	1994
Personal Author	Harmon, C. D.; Busch, R. D.; Briesmeister, J. F.; Forster, R. A.
Page Count	174
Abstract	With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.
Keywords	Criticality Alpha Particles Beta Particles Configuration Geometry M Codes Moderators Radiation Transport EDB/420203 EDB/990200
Source Agency	Technical Information Center Oak Ridge Tennessee
NTIS Subject Category	77 - Nuclear Science & Technology 77F - Radiation Shielding, Protection, & Safety
Corporate Authors	Los Alamos National Lab., NM.; Department of Energy, Washington, DC.; New Mexico Univ., Albuquerque, NM (United States).
Supplemental Notes	Sponsored by Department of Energy, Washington, DC.
Document Type	Technical Report
NTIS Issue Number	199503
Contract Number	W-7405-ENG-36