Response to Federico from Alejandro, 7/16/2008 (sonzogni at bnl.gov)

Hi Federico,

We use PREPRO for the doppler broadening and linearization. For most materials, it takes a few seconds to run; for some actinides, however, it may take up to several minutes. Our web servers are dual-core only, we will replace them soon with double-processor, quad-core servers and we'll test again if running fortran codes for 10 seconds or so can be done without compromising the server.
Thanks for communicating with us.

Alejandro Sonzogni

Response to Daniel from Alejandro, 5/30/2008 (sonzogni at bnl.gov)

We looked into the problem you presented yesterday, basically, why the cross sections from the "interpreted" table are different from those in the plot. Cross section data is kept in 2 sections of the ENDF file, one contains resonance parameters and the other non-resonant cross section at a temperature of 0 Kelvins. The plot shows reconstructed cross sections, that is combining the results of those 2 sections, and then doppler-broadened, at a temperature of 300 Kelvins. The rise of the cross section at the lower energy end is due to the doppler broadening. This rise disappears as one decreases the temperature.
Hope this helps, if not, let me know.

Alejandro Sonzogni

Response to Alejandro from Daniel, 5/30/2008 (36D at bechteljacobs.org)

Thank you so much for the explanation.

Daniel

Response to Gerard from Mike, 3/26/2008 (mwherman at bnl.gov)

Dear Sir,

the most reliable cross sections and uncertainties (actually full covariances) for the 238U(n,f) reaction can be found in the neutron cross section standards sublibrary of ENDF/B-VII.0. The file can be accessed through the NNDC Sigma interface located at http://www.nndc.bnl.gov/sigma/index.jsp
Please select Sublibrary: Neutron Cross Section Standards and proceed further to select 238U and fission. Note, that easy to read uncertainties are contained in the 'Introduction' to the evaluation while the full covariance matrix is given in File 33.
Best regards

MIke

Response to Peter from Mike, 1/14/2008 (mwherman at bnl.gov)

Dear Peter,

If I understand correctly the confusion arises from the fact that in the 'Basic Retrieval' panel, for the sake of simplicity, only major data library are explicitly listed while actual database contains more libraries. The full list is given only in the 'Advanced Retrieval' panel. If the libraries are selected the interface is searching for the evaluations only among those selected ones. However, if no selection is made all libraries are searched. In the particular case of carbon all major libraries contain only elemental evaluation (C-0) while C-12 evaluation is only included in the High-Energy data file ENDF/HE-VI, which normally does not show up in the 'Basic' retrieval as it is not recommended for typical low-energy applications.

As I mentioned above ENDF/HE is a particular, relatively old, library addressing the high-energy applications. The resonance part of the C-12 evaluations might not be up to date and one should rather use more recent elemental evaluations available in the major libraries. In any case there is NOTHING WRONG with C-12(N,XG) being larger than total. C-12(N,XG) (MT=202) is photon production cross section (not capture) and as such, due to the multiplicity of cascading gamma-rays, might be larger than the total cross section.
Best regards

MIke Herman

Response to David from Boris, 8/21/2007 (pritychenko at bnl.gov)

David,

17O(alpha,n) reaction data are available on NNDC website:

a) Experimental data from EXFOR/CSISRS database: http://www.nndc.bnl.gov/exfor

b) Evaluated data from JENDL-3.3 library, for example if you will use Sigma Web interface (http://www.nndc.bnl.gov/sigma):
1) Access “Periodic Table Browse” Tab and select “JENDL-3.3 (Japan, 2002)” and “alpha reactions” drop-downs; select 17O
2) Or Access “Directory Tree Browse” Tab and access “Charged Particle Reactions” folder and “Alpha Reactions” subfolder; select 17O
Thank you,

Boris.

Response to David from Boris, 8/16/2007 (pritychenko at bnl.gov)

David,

When thermal neutron reacts with proton: n + p -> 2H + (2.23 MeV) where 2.23 MeV is binding energy of deuteron. Deuteron has no excited states and energy can be released as gamma-ray and deuteron recoil energy. Obviously higher neutron energy may lead to higher gamma-ray energies.
Gamma-ray energy estimate should be based on energy and momentum conservation. You have to estimate energy that will be carried away by gamma-ray and (smaller) energy that will be transferred into deuteron recoil. You may look into " S.S.M. Wang, Introductory Nuclear Physics, page 65" for additional information on deuteron.
Additionally p(n,gamma)d reaction cross section is relatively small around 1 MeV, please see the attached picture.
Thank you,

Boris.

Response to Larry from Mike, 7/24/2007 (mwherman at bnl.gov)

Please consult 'ENDF-6 Formats Manual' available from the navigation bar on the left hand side of our ENDF web page (http://www.nndc.bnl.gov/exfor/endf00.htm).

The 'ENDF-6 Formats Tutorial' to be found right below the previous link might also be helpful for understanding general concept of the ENDF-6 format. This Tutorial much easier to follow but unfortunately it is somewhat incomplete and fission product yields are not covered.
Best regards

MIke Herman

Response to Arzu from Mike, 7/23/2007 (mwherman at bnl.gov)

They just do not exist. The ENDF/B-VII.0 is the first release of the ENDF library to contain photo-nuclear reactions. This sublibrary is to a large extent based on the results of the IAEA CRP but several evaluations were replaced with new files.
Best regards

MIke Herman

Response to William from Mike, 7/10/2007 (mwherman at bnl.gov)

Dear Dr. Kang,

The ENDF system was developed, about 40 years ago, for the storage and retrieval of evaluated nuclear data to be used for applications of nuclear technology. From the very beginning it was designed for being read by the processing codes while the 'human friendliness' was always a secondary issue. Although nowadays things could be done in a more transparent way it would cost millions of $$ and adequate manpower (both in high demand and low supply) to rewrite all codes that check, read and process the current ENDF-6 format.
To bring the cryptic ENDF-6 format closer to the ordinary users that would like to understand the content and read the numbers without running complicated processing codes such as NJOY or AMPEX we provide on our web site (www.nndc.bnl.gov/endf/) the so called 'interpreted' version of the original data. This form can be accessed by clicking on the 'Interpreted' button/link that shows up in the final stage of the retrieval, as shown on the two attached screen shots for the traditional and the new SIGMA interface to the ENDF libraries offered by the NNDC.





The 'interpreted' format is the result of preprocessing the original file with a code written by R. MacFarlane. This code translates most of the ENDF-6 formated file into a human (i.e., physicist) readable form, although some new extensions of the format or rarely used quantities might escape proper translation. In case, the interpreted version is not sufficient, consulting the ENDF-6 tutorial should help. Finally reading the ENDF-6 manual is the ultimate resource. Both are available from left side bar on www.nndc.bnl.gov/endf.
Best regards

MIke Herman

Response to Aaron from Mike, 6/27/2007 (mwherman at bnl.gov)

Yes, since the angular distribution of scattered neutrons is generally assumed to have azimuthal symmetry, the distributions are represented as a Legendre polynomial series in cosine of the scattering angle.

Yes, indeed. The L=0 coefficient is implicitly set to 1 and is not included in the file.

The Legendre expansion in the ENDF-6 format gives only the relative distribution normalized to 1. The absolute cross sections can be obtained by multiplying this distribution with the actual value of the angle integrated cross section given in File 3 (MF=3). The (2L+1)/2 factor is not included in the coefficients and must be taken into account explicitly when angular distribution calculations are being performed. All relevant equations and more detailed explanations can be found in the ENDF-6 format manual available from http://www.nndc.bnl.gov/csewg/.
Best regards

MIke Herman

Response to kevin from Alejandro, 6/21/2007 (sonzogni at bnl.gov)

Hi Kevin,

The data is entered in the simplest way, If you enter:
100, 10
200, 20
Sigma will think that you want to plot 2 points, one at 100 eV with a cross sectin of 10 barns, and the other at 200 eV and a cross section of 20 barns. If you input
100, 10, 1
It will understand a cross section of 10 +- 1 barns at an energy of 100 keV. If you input
100 10 10 1
It will understand a cross section of 10 +- 1 barns and an energy of 100 +- 10 eV. Hope this helps, if not, let me know.

Alejandro

Response to kevin from Mike, 6/14/2007 (mwherman at bnl.gov)

Yes, it is possible. You would have to modify cross sections in MF=3 MT=2 section of the file. Be careful to change only the values; don't change any positions, flags, number of points etc. - these would require deeper knowledge of the ENDF format. In order to find out cross sections in the MF=3 MT=2 you may use the 'interpreted' representation of the ENDF file available from the NNDC web site. All the changes, however, have to be done in the original ENDF-6 formatted file (not interpreted).
Best regards

MIke Herman

Response to Hans from Alejandro, 6/5/2007 (sonzogni at bnl.gov)

Hi Hans,

The cross sections are in the lab system. We'll modify the help file to make this clear.
Cheers

Alejandro

Response to Eric from Mike, 5/10/2007 (mwherman at bnl.gov)

Unfortunately there is no good reference to ENDF/B-VI. This is why we took care to fix it for VII.0. When referring to ENDF/B-VI the most important is to specify the release (the most recent was 8 and we usually write it as ENDF/B-VI.8 although some people prefer to use ENDF/B-VI release 8). Having said this, I believe that the most relevant reference to ENDF/B-VI.x would be the CSEWG web page: http://www.nndc.bnl.gov/csewg/
Best regards

MIke Herman

Response to Qi from Ramon, 1/24/2007 (arcilla at bnl.gov)

Dear Dr. Qi Ao,

As requested, please find attached the NJOY input decks for generating ACE files from the ENDF/B-VII.0 neutron and thermal scattering sublibraries. I hope these files would help you get started while waiting for your DVD from RSICC. For a description of these input decks (and the DVD itself), please click on the link below: http://www.nndc.bnl.gov/exfor/4web/acefiles.html. Thank you for giving NNDC the opportunity to serve your nuclear data processing needs.
With best regards,

Ramon E. Arcilla Jr.

Response to Qi from Ramon, 1/25/2007 (arcilla at bnl.gov)

Dear Dr. Ao,

I am sorry to inform you that NNDC does not have a RUNJOY script for MS Windows because we perform all our nuclear data processing on a Linux cluster. Thank you again for your interest in using the ENDF/B-VII.0 neutron and thermal scattering sublibraries. And all the best to your nuclear data processing activities!
Best regards,

Ramon

Response to Qi from Ramon, 1/31/2007 (arcilla at bnl.gov)

Dear Dr. Ao,

May I ask what version of NJOY you are using? For your information, my input decks have been tested using NJOY-99.161 only.
Thanks.
Best regards,

Ramon

Response to Qi from Ramon, 2/1/2007 (arcilla at bnl.gov)

Hi Qi,

I am trying to understand what may have gone wrong here. Did you say in one of your e-mails that you are running NJOY under MS Windows? My input deck was fully tested only on a Red Hat Linux system using Portland Group's Fortran compiler. Further, under Linux one has to use a patch to handle accuracy problems in the math SLATEC function which is being used to calculate the incomplete gamma function in the Madland-Nix fission spectrum. Thus, I cannot fully guarantee that it would work under MS Windows using a different compiler. If you are indeed running NJOY under MS Windows, you may have to contact Skip Kahler (LANL) who uses the same platform.
Best regards,

Ramon

Response to Qi from Ramon, 2/2/2007 (arcilla at bnl.gov)

Qi,

As requested, Skip's complete contact information is:
Dr. Albert C. KAHLER
Group T-16
P.O. Box 1663, MS B243
Los Alamos National Laboratory
Los Alamos, NM 87545
akahler at lanl.gov
Good luck! And have a great day.

Ramon

Response to Mohamed from Boris, 1/28/2007 (pritychenko at bnl.gov)

Dear Mohamed,

Sorry for this issue with our Web interface. The whole problem originates from the fact that original ENDF/B-VI.8 neutron evaluation for H3 lists it as "1-T-3", the same problem with JEF-2.2 library. As possible work around solution I will suggest to put two isotopes separated by semicolon in the target field as "T3;H3". This will allow you to extract tritium neutron data from all libraries.
Sincerely,

Boris.

Response to Royce from Mike, 1/25/2007 (mwherman at bnl.gov)

Royce,

the ENDF/B-VII.0 are stable, we'll not change them even if the most critical error is discovered (we would release VII.1 in such case but I hope there will be no need for such a desperate action). The difference you've noted is due to the fact that you had downloaded the basic version of the file previously and the reconstructed (300K) version today. The first one is actually nearly twice bigger than the second one! This is pretty unusual, I must admit, but happens because RECENT is not adding anything (no resonances in the Cl35 file) while LINEAR is removing redundant energy points throughout the file.
Generally, I would recommend to stick to the basic version of the file but being the author of the original evaluation you may also discuss it with your colleagues and decide to choose the processed version if you believe there is no need for a very dense grid in certain reactions (such as (n,p), ...)
Regarding wrong NC values - they are only wrong in the reconstructed (300K) file. The reason is that after reconstruction these files were not run through STANEF as the only purpose of having these files was to have cross sections in the resonance region included in the plots. They are not intended for use in any application. Actually, our new interface will only use them for plotting and the retrieval will always download the basic version of the file. Hope it helps to understand the current situation.
Cheers

MIke

Response to Iosif from Ramon, 1/16/2007 (arcilla at bnl.gov)

Dear Dr. Prodea,

Since you belong to the IAEA's area of responsibility, I would like to refer your CD request to the following contact person:

VIKTOR ZERKIN
Nuclear Data Section
International Atomic Energy Agency
P.O. Box 100, Wagramer Strasse 5
A-1400 Vienna, Austria
E-mail: V.Zerkin at iaea.org
zerkin at ndsalpha.iaea.org

If you prefer to download a particular ENDF/B-VII.0 sublibrary(ies) instead, please visit our Web site at 'www.nndc.bnl.gov'. Click on 'ENDF/B-VII.0 released' in the center, right-hand side of the home page. On the ENDF/B-VII.0 Web page, place your mouse over the sliding menu item 'Download Sublibrary' which is among the items on the left-hand side of the page. A pop-pup menu will allow you download any of the 14 sublibraries in 'ZIP' format. If you encounter further problems, kindly let me know. Thank you for your interest on using the new ENDF/B-VII.0 library.
With best regards,

Ramon E. Arcilla Jr.

Response to Gunter from Pavel, 12/27/2006 (oblozinsky at bnl.gov)

Gunter,

ENDF/B libraries do not contain activation cross sections in an explicit way. You should go to 'Advanced Retrieval' option of our ENDF retrieval system, and search for data in the European Activation File labeled as 'JEFF-3.1/A (Activation)'. This library contains huge amount of neutron activation cross sections.
I hope this helps,

Pavel

Response to Phil from Mike, 12/21/2006 (mwherman at bnl.gov)

Phil,

it might depend on how you get the ENDF evaluations. If you go through the basic request (the default option when you come to the ENDF retrieval site) than you get reconstructed file at 300K which implies it was run through LINEAR, RECENT and SIGMA1. We choose to do it this way since otherwise newcomers looking at the plots believed that the cross sections are 0 in the resonance region or at least there is nothing in the file in this energy range. With the current interface we were not able to distinguish between plotting and retrieving and that's why we offer 300K for both. If you want to download the basic version of the file you have to go to "Extended retrieval" and check 'Basic version' of the library.

In the new SIGMA interface (currently under development, preliminary version already shown on ENDF/B-VII.0 web) we will always offer basic version of the file for download while the reconstructed one will only be used for plotting. This should eliminate current confusion. I must admit you are not the only one who notice this.
-- Cheers

MIke

Response to Reuven from Pavel, 12/20/2006 (oblozinsky at bnl.gov)

Dear Reuven,

Thanks for congratulations. Release of ENDF/B-VII.0 was announcened to full CSEWG-USNDP distribution list of which the ENDF discussion-group list is a subset. We will check once more to make sure that nobody was omitted. The official ENDF/B-VII.0 library is not identical to beta3 release, though the changes between the two are relatively minor. Best wishes for the holiday season,

Pavel

Response to Arjan from Pavel, 12/19/2006 (oblozinsky at bnl.gov)

Dear Arjan,

NNDC as well as LANL are purchasing quite a large number of copies, but they were not delivered yet. You and Steven are on the NNDC priority list. In addition, either LANL or NNDC will send to NEA a bunch of copies for free distribution. Best wishes for Christmas and New Year,

Pavel

Response to Alan from Pavel, 12/19/2006 (oblozinsky at bnl.gov)

Thanks Alan,

This was possible also thanks to important contributions from the IAEA as explicitely stated in Acknowledgements of the Big Paper. Best wishes for Christmas and New Year,

Pavel

Response to Yaron from Mike, 12/19/2006 (mwherman at bnl.gov)

The new interface is currently under development and only ENDF/B-VII.0 library was loaded into the database. More libraries and more features will be added shortly so you might keep checking it on regular bases - we'll be grateful for any suggestions regarding the design and functionality as well as for eventual bug reports.
Best regards

MIke Herman

Response to Balraj from Pavel, 12/19/2006 (oblozinsky at bnl.gov)

Thanks Balraj,

Pavel

Response to Roger from Mike, 12/19/2006 (mwherman at bnl.gov)

Thank you for so immediate comments regarding the new ENDF web site. Your observations are perfectly correct, however, ENDF-6 tutorial is a LANL product, is being hosted at T-16 and is not maintained by NNDC. We just provide a link to it since we believe it is a good introduction to the ENDF-6 format for newcomers. Unfortunately, this tutorial is a bit out of date. I will see whether we can have it updated.
Considering nu-bar we shall include it in our help pages
--
Best regards

MIke Herman

Response to Ken from Arnold, 2/3/2007 (feroah at westinghouse.com)

Thank you for including me on the distribution of this landmark work. I really appreciate it.

If it isn't too late (and I fully understand that it might be), I would like to add two colleagues here at Westinghouse to the distribution list. They are:

Dr. F. Arzu Alpan
Westinghouse Electric Company
Engineering Services / Radiation Analysis
P. O. Box 158
Madison, PA 15663 USA

Dr. Gianluca Longoni
Westinghouse Electric Company
Engineering Services / Radiation Analysis
P. O. Box 158
Madison, PA 15663 USA
Thanks and best regards,

Arnie

Response to Emily from Donald, 8/30/2006 (Donald.L.Smith at anl.gov)

Pavel:

I could probably answer these questions, but they seem to be of an NNDC nature and it would seem appropropriate for you or one of your staff to respond to the request.

Don

Response to Emily from Pavel, 8/31/2006 (oblozinsky at bnl.gov)

Emily,

Your inquiry to Don Smith was forwarded to me. My answers:

- CSEWG is the organization responsible for ENDF library, all new evaluations should be submited to CSEWG through the NNDC for review. CSEWG is also the final judge regarding inclusion of these data to official ENDF/B library. Technically, any new file will be first reviewed by the ENDF database manager Mike Herman. The process itself may be quite long. The current ENDF/B-VI.8 was released in October 2001, while new ENDF/B-VII.0 should be released Nov 2006. This means that there was no release in the last 5 years, but it is not clear how updates will be handled in ENDF/B-VII. This should be decided by CSEWG in its annual meeting, Nov 2006.

- To the best of my knowledge, all actinide files with cross sections are freely available. The latest versions are under extensive validation, you can find them under ENDF/B-VII beta2 that is directly accessible from the top of the NNDC webpage,

Pavel

Response to Woo from Mike, 6/29/2006 (mwherman at bnl.gov)

Am-242 lacks data for the mt=18 of mf=4.  I was not able to run NJOY (version99.112) because of this. Please note that explicitly stated on the processing web page (http://www.nndc.bnl.gov/exfor2/4web/processing.html) is: " WARNING: Some materials will not process through NJOY versions older than 99.125". You have to install patch 125 in NJOY-99 in order to process the whole ENDF/B-VIIb2 library.
Best regards

MIke Herman

Response to Cornelius from Mike, 6/29/2006 (mwherman at bnl.gov)

Dear Dr. Broeders,

I'm sorry for the problems with the zip file. It has been prepared on Red Hat Enterprise 4 Linux and it seems to work fine on Fedora Core 3 & 5. We do not have SuSe so I was not able to test it on this version of Linux. In principle, I would prefer to use tar and gzip (instead of zip) but some Windows users were complaining, although recent WinZip should be able to open .tgz files. There is no way to make everybody happy (at least as long as Windows is around). Regarding NJOY, I know that MacFarlane and Kahler are tracing compiler incompatibilities so I copy them on this mail.
Best regards

MIke Herman

Response to Thomas from Boris, 3/9/2006 (pritychenko at bnl.gov)

Thomas,
I do not see such data in evaluated neutron libraries for H(n,p), charge-exchange reactions. However you probably dealing with neutron moderation in Hydrogen, i.e. elastic and inelastic neutron scattering on Hydrogen. In this case:
1) go to NNDC website at: http://www.nndc.bnl.gov/endf
2) target: H-1
3) reaction: n,el
4) quantity: sig
5) select all 300 K libraries (ENDF, JEFF and JENDL)

Retrieve the data in original ENDF or Interpreted format, you also can plot it and retrieve plotted data.
Thank you,

Boris.

Response to Daniel from Boris, 3/9/2006 (pritychenko at bnl.gov)

Daniel,
You probably working on the "Sudbury" class of experiments when solar neutrinos interact with D2O and produce neutrons. The main task is to detect them and Gd is a good choice because it will produce plenty of gamma-rays and you can dissolve it in a heavy water. Neutron cross sections for gadolinium isotopes are available at our website and I will suggest to use Evaluated Nuclear Data File library for this purpose:
1) http://www.nndc.bnl.gov/endf
2) Target: Gd*
3) Reaction: n,g
4) Quantity: sig
5) Library: select all three 300 K libraries (ENDF,JEFF,JENDL).
6) Click "Submit" button and access "ENDF Data Selection" page containing 21 evaluations for different Gd isotopes.
7) Use an "Interpreted" button to access data in the interpreted format.
8) You can plot the data by selecting evaluations and clicking on the "plot" button.
9) An "Output Data" page will display the plotting results and also contains "pltted data" link that will allow you to access the data in a commonly-understandable format.
Thank you,

Boris.

Response to Chris from Mike, 3/6/2006 (mwherman at bnl.gov)

Combination of output format check boxes ("ENDF", and "Plot") and the "Submit" button found in the previous interface have been replaced by the "Retrieve" and "Plot" buttons (see the attached screen-shot). This change was introduced to simplify the interface (one needs only one click instead of two to plot or retrieve data). It is enough to select data from the list and click on the "Plot" button in the top of the screen to get both a plot and a link to the ENDF-6 formated data ("Text" or "Zip") displayed on the screen (see second screen-shot).
Best regards

Mike Herman

Response to Meehan from Mike, 1/31/2006 (mwherman at bnl.gov)

The ENDF interface has been modified to make it simpler (more approachable) for the users that are not familiar with the ENDF-6 format and ENDF libraries in general. Hiding the above mentioned links was intentional and should provide for a cleaner look of the interface. As a matter of fact the links are still active but invisible and I suppose that this is the reason for the confusion. However, the links are still available in the 'Advanced Request' for those who would like to use them.
Best regards

Mike Herman

Response to Mikhail from Boris, 2/24/2006 (pritychenko at bnl.gov)

There is no problem in finding Thermal Neutron data that is currently available. Just go to NNDC website, select ENDF application/Advanced Request: http://www.nndc.bnl.gov/exfor3/endf11.htm and two input parameters:
Projectile= "tsl" and Library = "All". Click on the "submit" button and retireve 34 data sets in original and interpreted formats from ENDF/B-VI, JEFF-3.1, JEFF-3.0 and JEF-2.2.
Basically everything was is currently available. The same website has a link to ENDF/B-VII beta1 library. This new library in the testin stage and contains 20 TSL data sets, you may look at them too.
Unfortunately, we cannot to redistribute NJOY for you guys.
Thank you,

Boris.

Response to Mikhail from Boris, 2/24/2006 (pritychenko at bnl.gov)

Please follow this procedure:
1) http://www.nndc.bnl.gov/exfor3/endf11.htm
2) Projectile textbox: TSL
3) Library radio-buttons: All
4) Clck on "Submit" button
5) You will "ENDF Data Selection" Page. You can select different evaluations and retrieve them in the original or interpreted formats.

Interpreted format was developed by Bob McFarlane (LANL) to make ENDF-6 format more humanly-readable. The main problem of ENDF library that it requires you to know ENDF-6 format. Actually, all evaluated neutron data libraries are using ENDF-6 format. The best way around this problem is to use "interpreted" format from Bob McFarlane.
Thank you,

Boris.

Response to Jaya from Mike, 1/31/2006 (mwherman at bnl.gov)

Dear Dr. Ramanuja,
Unfortunately, I'm not in a position to help you in this case. Thermal and fast neutron flux produced by the proton accelerator depend on a number of deign and construction factors (e.g., type of the target used to produce neutrons, its geometry, shielding, etc.) that are specific to each facility. Therefore, these are not universal quantities that can be treated on the same footing as other nuclear data distributed by the NNDC. Actually, thermal and fast neutron flux can be calculated using basic nuclear data (ENDF library) but it is a complex procedure involving series of codes such as NJOY and MCNP or GEANT. Usually, it is not a trivial task.
Best regards

Mike Herman

Response to Phil from Mike, 1/25/2006 (mwherman at bnl.gov)

In order to get requested fission yields you should select :
Target: U-235
Reaction: N,ind_FY; FY_cum
It is convenient to use help (small triangle after the end of the field) for the reaction quantity. X-135 will be buried in the ENDF-formatted file including all fission products. The two attached screen shots show the request and the result of the retrieval.
Best regards

Mike Herman

Response to Kusche from Boris, 12/01/2005 (pritychenko at bnl.gov)

I will suggest to access ENDF database http://www.nndc.bnl.gov/endf and search all data libraries for targets al-27, ti-48 and reaction p,*.
Thank you,

Boris.

Response to Joe from Dimitri, 9/30/2005 (drochman at bnl.gov)

Hello Joe, These two web pages present the same evaluation (or data) but with different format. At NNDC, it is the ENDF format and at Los Alamos, it is their own format (easier to read...)