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Dear Sirs,
I've been experimenting with the new retrieval and plotting capabilities
of exfor and endf files available through your website at
http://www.nndc.bnl.gov/exfor/endf00.htm and http://www.nndc.bnl.gov/sigma/
I found them very useful indeed. However I was wondering if it could be
added an option to plot (and then to get the linearly interpolable data) a
given endf file Doppler-broadened at a temperature different from 300 K
(or 293.5 K). I understand that codes like NJOY and Prepro require some
finite time to process the data to a given temperature, but perhaps
waiting some seconds online would not be tremenduous.
The options in NJOY and Prepro could be set at a fairly low accuracy level
(even thinning options could be turned on) so to decrease the computation
time.
This feature of Doppler-broadening by user-defined temperature would
improve a lot the usefulness of your retrieval systems and would spare a
lot of time to those who don't want to spend a lot of time to process data
(as required for doing transport calculations) but simply to reason about
physics, teach, compare evaluations etc.
Do you think that such a feature could be implemented in the future?
Thanks in advance,
Yours,
Federico
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-- Submitted by: Federico on 7/15/2008 (federico.rocchi at unibo.it)
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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
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Mr. Pritychenko,
I accessed “Sigma, ENDF Retrieval and Plotting”,
Then clicked on hydrogen (1-H-1) in the periodic table browse,
And then looked at the plot and the interpreted table for (n, total) cross section.
The table gives 20.8 b at 0.0253 eV, but the plot gives about 31 b at the same energy.
I expected them to be the same. Why are they different? What am I missing?
Thanks in advance for your help.
Daniel
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-- Submitted by: Daniel on 5/29/2008 (36D at bechteljacobs.org)
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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
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Response to Alejandro from Daniel, 5/30/2008 (36D at bechteljacobs.org)
Thank you so much for the explanation.
Daniel
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Dear Sir,
We are doing fission cross section measurements relative to the 238U
one. In the fast neutron energy range ( From 1MeV to 20 MeV) what is
(are) the uncertainty(ies) on the evaluated 238U(n,f) reaction cross
section?
Best regards
G. BARREAU
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-- Submitted by: Gerard on 3/26/2008 (barreau at cenbq.in2p3.fr)
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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
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Hi,
I'm a bit confused about the different data sets in the interactive web
tool. Both http://www.nndc.bnl.gov/exfor/endf00.htm and
http://www.nndc.bnl.gov/exfor7/endf00.htm claim to be interpreted ENDF/B-VII.0 data on the front page but I seem to get different datasets if I select say 12C and n,* without selecting anything in the "library" box ( or click Reset ).
Also, in case of the exfor/endf00.htm the ENDF/HE-VI: C-12(N,XG) cross section must be wrong - it is larger than the TOTal ( ? )
Peter Skensved
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-- Submitted by: Peter on 1/14/2008 (peter at SNO.Phy.QueensU.CA )
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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
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Boris:
Where can I find the cross section for (alpha,n) reaction in oxygen-17 as a function of alpha energy?
Sincerely,
David
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-- Submitted by: David on 8/20/2007 (D_Brinegar at msn.com)
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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.
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Boris:
CapGam shows 2,223.25 Kev gamma is produced when hydrogen absorbs a thermal neutron. Are higher energy gammas produced when higher energy neutrons are absorbed?
Neutrons of 2Mev to 3 Mev should leak from a general purpose heat source. How can I estimate the gamma energy that would be produced if light hydrogen absorbs a 2 Mev neutron and a 3 Mev neutron?
Sincerely,
David
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-- Submitted by: David on 8/16/2007 (D_Brinegar at msn.com)
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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.
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I am using the endf database for the first time. I am looking for
fission product yield data for U235 fission. I was able to select the
MF=8 and find the FPY files, but how do I interpret them? I assume
there is a file description somewhere, but I have not found it. Could
you point me in the right direction?
Larry Wetzel
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-- Submitted by: Larry on 7/24/2007 (LLWetzel at bwxt.com)
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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
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Dear Dr. Herman,
I am a Senior Engineer at Westinghouse, in the Radiation Analysis Group.
One of the work I am involved in here is on cross sections. I have a
question on ENDF/B-VI, if you could help. I tried to find the
ENDF/B-VI.3 photonuclear data from BNL and LANL nuclear data web
sites, but could not locate it. I saw IAEA and ENDF/B-VII photonuclear
data, though. I was wondering how I could obtain the ENDF/B-VI.3
photon interaction cross sections.
Arzu
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-- Submitted by: Arzu on 7/23/2007 (alpanfa at westinghouse.com)
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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
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Dear Dr. Herman,
Thank you for your prompt response. If ENDF/B-VI.3 photon library does not exist, would you know which photon
ENDF was used in generating BUGLE-96? My understanding was that ENDF/B-VI.3 photon data were processed to obtain
the 20-group gamma's of the 47-neutron, 20-gamma group BUGLE-96 library.
Arzu
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-- Submitted by: Arzu on 7/23/2007 (alpanfa at westinghouse.com)
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Hello, Mr. Herman.
My name is William Kang, and I am an employee at NSTec.
I was wondering if you could provide some documentation or explanation
regarding the formatting of output data in text form on ENDF; as it
is, the text form is rather confusing, and eludes interpretation.
Thank you.
Sincerely,
William Kang
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-- Submitted by: William on 7/9/2007 (KangWB at nv.doe.gov)
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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
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I am trying to get the angular distribution (da) for elastic neutron
scattering off of Xe nuclei. My search parameters on the ENDF site
are (Xe*; n,el; da), and then I use the "Interpreted" results, but I'm
having some confusion when it comes to interpretation of these
results, which I do not find answers for in the help section.
First, in the help section it says "da" is "Angular distribution for
emitted particles". Does this mean (d2sigma)/(dtheta dphi) ?
The table says "Legendre polynomial coefficients given". But I seem
to get the most reasonable results if I assume the list of
coefficients starts at L=1, and not L=0. Is this correct?
If so,
what do I use as the vertical offset? Also, do these coefficients
already contain the (2L+1)/2 factor that is standard for Legendre
sums?
Thank you,
Aaron
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-- Submitted by: Aaron on 6/27/2007 (aaronm at phys.ufl.edu)
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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
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That reply was quick, and I think you cleared up all confusions I had. Thanks!
Cheers,
Aaron
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-- Submitted by: Aaron on 6/27/2007 (aaronm at phys.ufl.edu)
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Hi,
I would like to upload my data in sigma to watch the plot and compare it to another existing library.
I wonder in wich format I have to enter my data?
Thank you,
Best regards
Kevin
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-- Submitted by: Kevin on 6/21/2007 (kevin.chtioui at tsl.uu.se)
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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
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Dear all,
I am a French exchange student in Uppsala in Sweden.
I work in simulations with mcnpx. I would like to change some individual
cross sections and check the effect on my MCNPX results. I have aquired
NJOY99, which can translate an endf library in an ace library (mcnpx
format). I wonder now if there is a way to change the endf library.
For instance I work with neutrons and I would like to double the elastic
scattering cross section of the isotope Fe56. Is it possible and How could
I do it?
Thank you.
Best regards.
Kevin
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-- Submitted by: Kevin on 6/13/2007 (kevin.chtioui at tsl.uu.se)
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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
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Hi.
Just a simple question about Sigma: Are the energies for cross sections given in the
laboratory system or the center-of-mass system? I assume they are in the
lab system since experimental data is given but I just want to be sure.
Thanks in advance!
Hans Peter Loens
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-- Submitted by: Hans on 6/5/2007 (h.p.loens at gsi.de)
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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
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I want to reference ENDF/B-VI on a paper I am submitting but I cannot
find that information anywhere. There is an ENDF/B-VII reference on
the website. Is there a B-VI reference somewhere, or could you send me one?
Thanks,
Eric Edwards
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-- Submitted by: Eric on 5/10/2007 (ejedwards at wisc.edu)
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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
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Dear Dr. Arcilla,
I need to generate ENDF/B-VII.0-ACE library (neutron and thermal scattering) at different temperatures
with NJOY for our applications. Could you please send me a general NJOY input deck? I have requested a
DVD from RSICC but haven't got it yet.
Thanks in advance.
Qi Ao
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-- Submitted by: Qi on 1/24/2007 (Qi.Ao at ge.com)
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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.
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Dear Dr. Arcilla,
Thanks again.
Do you have a RUNJOY script for MS Windows platform, such as RUNJOY.BAT? We don't run LINUX.
Best regards,
Qi Ao
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-- Submitted by: Qi on 1/24/2007 (Qi.Ao at ge.com)
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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
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Dear Dr. Arcilla,
When using your NJOY input deck (runjoy) to process H-1 in ENDF/B-VII, I got the following error:
acer... 0.6s
***error in convr***only law=1 allowed for endf6 file6 photons.
Do you have any suggestions.
Thanks,
Qi
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-- Submitted by: Qi on 1/31/2007 (Qi.Ao at ge.com)
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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
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Ramon,
I used the old version and will upgrade to 99.161.
Thanks,
Qi
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-- Submitted by: Qi on 1/31/2007 (Qi.Ao at ge.com)
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Ramon,
I upgraded our NJOY to 99.161 and tried with your input deck for a couple of ENDF/B-VII nuclides (H1,O16, F19, U235 and U238) at different temperatures. Unfortunately, all failed except for those at 0 K. The following is the error message from NJOY:
broadr...
---message from initds---chebyshev series too short for specified accuracy
---message from initds---chebyshev series too short for specified accuracy
---message from initds---chebyshev series too short for specified accuracy
Any suggestions?
Best regards,
Qi
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-- Submitted by: Qi on 2/1/2007 (Qi.Ao at ge.com)
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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
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Hi Ramon,
Yes, I am running NJOY under MS Windows XP. Your comments are helpful.
Can you send me Skip Kahler's contact?
Thanks,
Qi
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-- Submitted by: Qi on 2/1/2007 (Qi.Ao at ge.com)
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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
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Dear Dr. Albert Kahler,
I was referred to you by Dr. Ramon Arcilla of BNL.
I am currently working on the NJOY process of ENDF/B-VII for our applications at different temperatures with MCNP . NJOY99.161 was run on MS Windows XP and generated the following error:
broadr...
---message from initds---chebyshev series too short for specified accuracy
---message from initds---chebyshev series too short for specified accuracy
---message from initds---chebyshev series too short for specified accuracy
I will appreciate your comments and suggestions. For your reference, the message from Dr. Arcilla is included below.
Best regards,
Qi Ao
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-- Submitted by: Qi on 2/2/2007 (Qi.Ao at ge.com)
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Dear Pavel,
Thank you for your quick response. I am clarifying my first question.
When I check for data from all evaluations for H-3 on the NNDC website I get these available evaluations:
ENDF/B-VII.0
ENDF/B-VII.0 300
JEFF-3.1
JEFF-3.1 300
BROND-2.2
CENDL-2
JEF-3.0
JEFF-3.1/A 293
I need to compare plots for ENDF/B-VI.8 (that was used in FENDL-2.1) to the newly released ENDF/B-VII.0.
However, ENDF/B-VI.8 for H-3 is not available on the website. My question is where can
I acess it on the website for the comparison.
Best regards,
Mohamed
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-- Submitted by: Mohamed on 1/26/2007 (sawan at engr.wisc.edu)
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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.
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Mike,
Today, Jan 25, at 7:55 AM, I downloaded the Cl-35 ENDF VII.0 file from the NNDC website.
A few records are copied below.
....
The above file is VERY different from the Cl-35 ENDF VII.0 file that I retrieved
two days ago. It is much smaller - in part because the energy mesh
in the previous file was extremely fine in certain sections, e.g. File 3, section 103.
The energy mesh in the new VII.0 file appears to be much more reasonable.
I ran the new VII.0 file through checkr and fizcon - no errors.
Note, however, that some File 1 NC values do NOT correspond to the
number of records
in the corresponding sections. For example, here is an excerpt from File 1:
3 4 79 01725 1451 444
3 102 6683 11725 1451 480
3 103 2014 11725 1451 481
3 104 59 01725 1451 482
and here are the actual number of records
3 4 25
3 102 9230
3 103 1194
3 104 20
Apparently the codes do not check the NC values.
As you know, I am in the process of redoing the merge of resonance
parameters and covariance matrices into the Cl-35 and Cl-37 ENDF VII.0
files.
Please let me know when the VII.0 files are stable so I won't be presented with the proverbial
moving target.
Thank you,
Royce
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-- Submitted by: Royce on 1/25/2007 (sayerro at ornl.gov)
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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
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Good afternoon Mr. Oblozinsky and A HAPPY NEW YEAR !
I'm Iosif Prodea from INR Pitesti (Romania) and I have been pleased to
meet you in Antwerp (April,5-7,2005) at The "Nuclear Data Needs for
Gen.IV Nuclear Energy Systems" Workshop.
As the final version of ENDF/B-VII has been released, I tried to
transfer (via ENJOY code) some newer data (for U235 and U238).
into the WIMS library. Unfortunatelly, despite of many selected
reactions and other options, I couldn't run NJOY without errors.
I always received something related to "an MT data missing"
or insufficient data in the downloaded file from the site:
http://www.nndc.bnl.gov/exfor/endf00.htm
Therefore I kindly ask your help.
Could you tell me how can I get (on CD or by downloading) a full
version of the newest ENDF/B-VII files? And, what conditions, please.
I apologize for inconvenience and thank you in advance.
BEST WISHES for The New Year.
Iosif Prodea
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-- Submitted by: Iosif on 1/12/2007 (vprodea at gmail.com)
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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.
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Response from Iosif to Ramon, 1/17/2007 (vprodea at gmail.com)
Dear Dr. Arcilla Jr.,
Thank you very much for your prompt, detailed and usefull answer.
I hope to be able to carry out at least one NJOY task with these sublibraries.
I shall inform you about the further developments.
Sincerelly,
Iosif Prodea
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Hi Pavel,
Where do I look up Indium neutron activation cross sections?
Thanks for your help.
Gunter
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-- Submitted by: Gunter on 12/27/2006 (Gunter_Kegel at uml.edu)
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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
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I just discovered that the evaluations at the NNDC for ENDF/B-VI R8
are a little different from the ones on Bob's T2 Data website. It
seems that the NNDC evaluations have been run through a linearizing
code, with the result that they are much longer and are not the
actual evaluations that were sent to the NNDC. It probably does not
make any difference for integral calculations but I find it strange
that they made this change.
Regards,
Phil
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-- Submitted by: Phil on 12/20/2006 (pgy at lanl.gov)
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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
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Response to Mike from Phil, 12/21/2006 (pgy at lanl.gov)
Thanks Mike.
That is certainly reasonable. And, it is convenient to have an easy source for making plots in the resonance region.
Best wishes and Merry Christmas,
Phil
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Dear colleagues,
Congratulations to the official release of the ENDF/B-VII.0 library!
We deeply appreciate the great effort and dedication of so many, that was
invested in making this culminating event possible. May I suggest that this release be announced
also to the participants of the ENDF discussion-group mailing list? Is ENDF/B-VII.0 identical to endfb7.beta3?
With kind regards and all the best for the holiday season,
Reuven L. Perel
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-- Submitted by: Reuven on 12/20/2006 (perel at huji.ac.il)
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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
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Dear Pavel
Congratulations, you all gave yourself a nice Christmas present.
Is there any chance that some Nucl Data sheets reprints of the
Big Paper are left? Have a nice christmas,
Arjan
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-- Submitted by: Arjan on 12/19/2006 (koning at nrg-nl.com)
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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
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Dear Pavel
Congratulations for a job well done. The significant new evaluations within ENDF/B-VII are certainly a major step
forward in providing users with extremely valid and reliable data. NNDC played the key role in ensuring that all
deadlines were comfortably met, as well as generating a significant fraction of the valuable new data.
All your staff and yourself in particular should take great pride in your achievement.
Best wishes
Alan
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-- Submitted by: Alan on 12/19/2006 (A.L.Nichols at iaea.org)
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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
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I like the new web interface. How can I use other libraries with the
new interface?
Dr. Yaron Danon
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-- Submitted by: Yaron on 12/18/2006 (danony at rpi.edu)
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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
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Pavel:
Congratulations on this major achievement! Best wishes for the Holiday season and the NEW YEAR!
Balraj
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-- Submitted by: Balraj on 12/18/2006 (ndgroup at univmail.cis.mcmaster.ca)
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Response to Balraj from Pavel, 12/19/2006 (oblozinsky at bnl.gov)
Thanks Balraj,
Pavel
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A couple of minor comments on the web site:
1. We all know this, but the ENDF6 format tutorial says file 1
contains miscellaneous data. Nu-bar is hardly miscellaneous, and
should be mentioned.
2. Isn't JAERI now JAEA?
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-- Submitted by: Roger on 12/18/2006 (rnblomquist at anl.gov)
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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
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Dear colleague,
The National Nuclear Data Center (NNDC) at Brookhaven National Laboratory (BNL)
will soon be releasing the official version of the Evaluated Nuclear Data File,
ENDF/B-VII.0 (scheduled for Dec. 15) - the first major revision of ENDF/B since 1990.
By all accounts, it is a significant advancement with noticeable improvements,
particularly for LEU light water critical benchmarks. Its release will be accompanied
by the publication of a special issue of Nuclear Data Sheets containing two long papers:
"ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology"
by Chadwick et al. and a companion paper by Steven van der Marck of NRG, describing the results
of an extensive set of independent criticality and other benchmark tests of ENDF/B-VII.
In recognition of this event, BNL has ordered a large number of copies of the special issue
and would be pleased to distribute them free of charge to key individuals and institutions
where they will have the most impact. In particular, Pavel Oblozinsky of NNDC has asked myself
and other CSEWG (Cross Section Evaluation Working Group) members to identify about 100
or so members from the global reactor end-user community who would benefit from this offer.
In particular, we would like to reach those "early adapters" who would start using ENDF/B-VII.0
in their day-to-day work and would provide feedback to NNDC on any issues that may arise.
Individuals who were actively involved in the nuclear data evaluation associated with
the earlier beta releases of ENDF/B-VII are already on Pavel's distribution list and will
be receiving complimentary copies of this issue.
At Pavel's request, I have assembled a consolidated list of names and mailing addresses of
individuals who have responded to an earlier message and additional people. Please review
the attachment to ensure that I have stated your name and mailing address correctly
and notify Pavel (oblozinsky@bnl.gov) of any necessary changes, or advise him of additional
persons in your respective organizations, and elsewhere, who would benefit most from receiving
this special issue.
Best regards,
Ken Kozier
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-- Submitted by: Ken on 11/28/2006 (KozierK at aecl.ca)
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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
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Arnie,
We will be happy to send a copy to each of your colleagues,
Pavel
PS: Marion, pl. add to the distribution list.
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-- Submitted by: Pavel on 2/4/2007 (oblozinsky at bnl.gov)
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Hello Dr. Smith,
I got your email address off the CSEWG website because I saw that you were the C
hair for the Physics and Basic Measurement part of the group.
I am working on a radiation transport project where I am running MCNPX and
I have a few quick questions for you. If someone has new data that could be
included in ENDF, how do they go about submitting it? How long would such a process take?
And would that data go through an evaluation process before being added to the files?
Also, obviously ENDF includes data and cross sections for fissionable materials.
Is all of this data online or are some of the ENDF files restricted access?
I have the MCNPX code software which includes the data libraries, but
I'm wondering if all of the ENDF files are available publicly.
Thanks so much for your help.
Sincerely,
Emily Wolfing
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-- Submitted by: Emily on 8/22/2006 (ewolfing at directedtechnologies.com)
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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
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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
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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
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Dear Colleagues,
After the corresponding e-mail from Dr. Sartori this morning, I started
implementing these resources on my LINUX PC. I am encountering the
following problems:
1) Problem with download of library zip files. Download on LINUX system
with new version of FireFox lead to error that zip file is not valid.
Transferring this file by standard ssh filetransfer to WINDOWS XP system
produced same error on XP PC for zip file.
Downloading on XP PC with recent version of FireFox resulted in useable
zip files. After transfer to LINUX machine data is useable there too. My
LINUX system is SuSE9.3 based.
2) Installing the njoy patch 125 on LINUX with LaHey Fujitsu FORTRAN
compiler version 6.20d gives errors in routine acer:
- warning line 14454 "nsyso is used but never set"
- error line 14455 and 14507 : "Invalid number of actual arguments for
intrinsic subroutine 'ERROR'"
The warning problem can be corrected easily by properly defining nsyso
(I did not yet look on that).
The "ERROR" problem is well known to me: different compiler have
different intrinsic ERROR handling routines. In my opinion it should be
recommended not to introduce these routines in new coding, but to use a
new routine with alternative name. In the current case probably a simple
print command (with succeeding stop???) could be applied.
Please give me your opinion on these comments.
With best regards,
C. Broeders
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-- Submitted by: Cornelis on 6/28/2006 (broeders at irs.fzk.de)
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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
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Hi,
I am looking for the cross sections for the H(n,p) reaction in the energy range from 1 meV up to 10 MeV but
failed to find an appropriate database. Please can you give me a hint which database I should use (and where I can find it).
Best regards and many thanks for your help
Thomas Bücherl
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-- Submitted by: Thomas on 3/9/2006 (buc at rad.chemie.tu-muenchen.de)
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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.
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Dear Sir/Madam,
I am a graduate student of the Chinese University of Hong Kong who is
working on a neutrino experiment. I would like to know the detailed information
about neutron capture by Gd. May I have this reference?
best regards,
Daniel Ngai
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-- Submitted by: Daniel on 3/7/2006 (wkngai at sun1.phy.cuhk.edu.hk)
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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.
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New interface is prettier, but seemingly less useful.
Go into endf and fill out target projectile and product, get nice
listing of citations, but the “plot” button doesn’t come up.
On earlier version one could chose whether to get data, plots, or
both.
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-- Hagen, Chris, 3/3/2006 (HagenEC at nv.doe.gov)
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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
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Response from Chris to Mike, 3/6/2006 (HagenEC at nv.doe.gov)
Thank you!
Fewer mouse clicks and a cleaner interface is always better.
Chris Hagen
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The links that are after the "target" field have invisible pointers to
the isotope selection pane ... they are invisible on the new version,
which was sort of confusing. Please put an icon there or something.
The version of browser that I am using is IE 6.0
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-- Meehan, Bernard, 2/24/2006 (MEEHANBT at NV.DOE.GOV)
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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
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Response from Meehan to Mike, 2/27/2006 (MEEHANBT at NV.DOE.GOV)
Thanks for the replies ... We like the looks of the new interface
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Dear colleagues,
I am beginning to deal with modeling of thermal neutrons,
and in order to start with
I consider necessary to study and use
thermal scattering group matrices sigma(Egroup_i -> Egroup_k) for
different elements\compounds.
Unfortunately, I can not find free access proper data.
Could you, please, provide an advise
where to find the data?
Yours respectfully,
Mikhail Fedorin,
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-- Submitted by: Mikhail on 2/24/2006 (mikeph at mail.ru)
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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.
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Response from Mikhail to Boris, 2/24/2006 (mikeph at mail.ru)
OK,
thanks!
I had tried to use this site,
but I failed.
After your recomendation and guidlines, I am sure to get progress.
Yours respectfully,
thanks again,
Mikhail Fedorin
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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.
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Dear Dr. Hermann,
Quite some time ago, you were very helpful to me in providing me with
cross sections for production of Hg-199m. I would really appreciate it
if you can provide me with info on thermal neutron flux and fast
neutron flux that causes activation of Ag-110(245 day half life) in a
30 MeV proton accelerator. I would like to calculate the radiation
level on silver target(natural mix of isotopes of silver) following 15
hours of irradiation. I have info on target mass, beam current, target
mass, thickness and area of irradiation. I would appreciate any
reference material as well. Thank you.
Sincerely,
Jaya Ramanuja,
Health Physicist
Mallinckrodt Inc.
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-- Submitted by: Jaya, 1/31/2006 (JayaRamanuja at tycohealthcare.com)
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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
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I have been trying (unsuccessfully) to obtain from the ENDF web page
individual and cumulative fission yield data (for slow neutron
fission of U-235) for several known fission product nuclei. Obviously, I have
not been filling in the search parameters in the standard request form
correctly, since I always receive a "no data found" message. What do
I place in the parameter fields to obtain the desired fission product
yields for, say, Xe-135?
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-- Submitted by: Phil 1/25/2006
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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
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Could you please provide (or suggest alternative sources for) recent sets of data giving the activation
cross sections and the reaction products for low energy (~10MeV) protons incident upon aluminum and titanium?
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-- Submitted by: Kusche 11/30/05 (kusche at bnl.gov)
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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.
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Can you tell me whether your ENDF discrete-line decay-gamma spectra data,
e.g., http://www.nndc.bnl.gov/exfor/x4data/E4R1413_e4.txt, are consistent with or identical to the data located on the LANL site,
e.g., http://t2.lanl.gov/cgi-bin/decay?203,3950? The file formats differ, so I am having difficulty assessing similarities and
differences. Thanks!
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-- Submitted by: Joe Durkee 09/30/05 (jdurkee at lanl.gov)
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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...)
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