Nuclear reactor physics by Weston Stacey

[vanesch]

"nuclear reactor physics" by Weston Stacey

I just bought it and started reading it. Anybody know the book ? Opinions, remarks ? I might have a few questions when reading it, are some people prepared to discuss it ?

As a first question, actually, on the cross section plots, there's always something like MT = 18 (for instance, page 7 already) or MT = 27 or the like. What does it mean ? Some kind of temperature ?

cheers,
Patrick.

 

[theCandyman]

I am familiar with the book, only because Professor Stacey is the only professor who does research with fusion here. I asked him a question one time and after he answered me, he said I could find out more in his book.

I think MT refers to the macroscopic total cross section, as in totaling the fission, absorpion, and scattering cross sections.

 

[Astronuc]

The data for the cross sections in Stacey's textbook originate from the Cross Section Evaluation Working Group (CSEWG) at Brookhaven National Lab (BNL). The MF and MT designations may be designations of data types or fields. They seem to be used in ENDF cross-section files.

BTW the BNL site links in Stacey's book are no longer valid.

Use - http://www.nndc.bnl.gov/csewg/ for the CSEWG, where you will find some information on Cross Section libraries and contact information. On this site, look on right side at top for - ENDF-6 Formats and Procedures Manual (ENDF-102) (PDF). This file (18.5 MB) contains a description of the nomenclature:

File MF

subdivision of a material (MAT); each file contains data for a certain class of information (e.g., MF=3 contains reaction cross sections, MF=4 contains angular distributions). MF runs from 1 to 99. (See Table 0.2 for a complete list of assigned MF numbers).

Section MT

subdivision of a file (MF) ; each section describes a particular reaction or a particular type of auxiliary data (e.g., MT=102 contains capture data). MT runs from 1 to 999. (See Appendix B for a complete list of assigned MT numbers).

For other inquiries use - http://www.nndc.bnl.gov

Stacey's book is for more advanced classes in reactor theory, usually first year graduate class in reactor theory.

 

[theCandyman]

Nevermind me, I was guessing from context. That file says that MT = 27 is rarely used though. I still cannot see how they get a number from a reaction, what an odd system!

 

[Astronuc]

From Appendix A, Glossary, of the ENDF-102 Data Formats and Procedures (pdf) document, MT is the "Reaction type number, or, covariance file section identifier."

Looking in Appendix B:

MT=18 is the fission cross-section.

MT=27 is the Absorption cross-section; sum of MT=18 and MT=102 through MT=117. Comment: rarely used.

Basically ENDF is used to generate macroscopic cross-sections with the appropriate codes.

The pdf document is 345 pages.

 

[vanesch]

Astronuc, thanks a lot ! I would never have found that out on my own (indeed, the link in the book is defective, thanks for providing an update on that too). Ok, Stacey could have said in a footnote what it was, no ?
Because finding an appendix on a 300+ pdf file on a server which is specified with a defective link is maybe not directly the most obvious way to understand this!

I'll be back with questions on the book I guess...

cheers,
Patrick

 

[Grogs]

If you haven't seen it, this is a pretty useful site for looking up cross-section information:

http://atom.kaeri.re.kr/ton/

Go to the nuclide you want, click on cross-section (XS) graphs, select the ENDF, and you can plot any of the categories (MT=whatever) you want. You can also download the data points into a text file if you need to use it in a computer program.

 

[theCandyman]

I just found this errata for the book; it may be useful to know whether something is a mistake or not if you get caught up on it.

http://www.frc.gatech.edu/ErrataReactor.htm

 

[N.E.moh]

hi all
please if anyone have the manual for this book because i need it to solve the problems for this book

 

[russ_watters]

We don't provide such things here.