Gamma Transport w/ Spent Fuel: Nuclide not found

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iankolaja
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Gamma Transport w/ Spent Fuel: Nuclide not found

Post by iankolaja » Tue Jul 07, 2020 3:19 am

Hi! I am currently attempting to do some gamma transport calculations with a canister of spent fuel elements. I produced my spent fuel composition through burn-up calculations using a full core model. My goal is to combine this material definition with a gamma source definition in external source mode, like so: "src spentfuel p sg spentfuel 2".

However, this approach lead to either a segmentation fault while processing the cross sections, or a memory allocation failure. I tried a handful of fixes involving buffer sizes, but that didn't fix it. I read that using the bumat file format can cause some issues, which is how I had extracted the composition of the spent fuel material. So, I attempted to fix the problem by using two separate input files to instead work with binaries.

With the first simulation, I ran a normal criticality simulation with an extremely short burn-up step and zero power. I included "set rfw" in order to "convert" the composition in a binary format in hopes of alleviating the memory issues.

Then for the second input, which was in external source mode with the gamma source defined, I loaded the binary file rather than directly inputting the composition of the spent fuel.

This, however, has produced a different set of issues. Now, I get the following error:

"Nuclide 271581 not found in the data libraries"

I tried setting "fpcut" to 1E-4 on the first input to ensure that I wasn't providing any unnecessary isotopes for which there may be a lack of gamma transport data, but that didn't change anything. Are there any other suggestions for how I can fix this? Let me know if I need to elaborate on anything I did. Thanks!

Ana Jambrina
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by Ana Jambrina » Tue Jul 07, 2020 10:39 am

271581 nuclide corresponds with Co-158m, available at ENDF/B-VII and JEFF3.1.1 cross section data libraries (27358.ID).

Off the top of my head:
- did you use the same libraries for both runs (neutron transport and decay source)? and include all necessary nuclide data libraries?
- which cross section data libraries are you using?

Could you post the input here?
- Ana

iankolaja
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by iankolaja » Tue Jul 21, 2020 3:38 am

Ana Jambrina wrote:
Tue Jul 07, 2020 10:39 am
271581 nuclide corresponds with Co-158m, available at ENDF/B-VII and JEFF3.1.1 cross section data libraries (27358.ID).

Off the top of my head:
- did you use the same libraries for both runs (neutron transport and decay source)? and include all necessary nuclide data libraries?
- which cross section data libraries are you using?

Could you post the input here?
Hey, thanks for the response. I did use the same libraries for both. For testing purposes, I'm working with ENDF/B-VII, JEFF3.2, and photon data a colleague got from from MCNP. For now, I actually went ahead and removed that nuclide, which seemingly fixed the nuclear data issue.

Now I am getting a different error:

Code: Select all

Fatal error in function ProcessMaterials:

Error in temperature: SS316H 6000.84p 3.000000E+02 0.0000000E+02

Simulation aborted.
There are my simulation parameters.

Code: Select all

src spent_fuel p
sg spentfuel 2

set rfr idx 1 canister.wrk
set opti 1
set srcrate 1
set bala 1
set gcu -1
set nps 1000 1

Ana Jambrina
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by Ana Jambrina » Tue Jul 21, 2020 3:39 pm

Are you using material temperature for Doppler-Broadening with material SS316H?

Photon data, from MCNP libraries (i.e., mcplib84, mcplib63) does not set up temperature, as it supposed to. Consequently, when Serpent tries to compare the nuclide temperature and the Doppler-Broadening temperature during (non-coupled) photon transport, there is a mismatch.

Try the following modification --> Serpent 2.1.31, after line 210 in processmaterials.c, add:

Code: Select all

      /* Skip if no neutron transport mode */
      
      if ((long)RDB[DATA_NEUTRON_TRANSPORT_MODE] == NO)   
        break;
- Ana

Nt009
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by Nt009 » Wed Oct 07, 2020 11:36 am

I'm working on spent fuel canister too and at the beginning I need to get the nuclides composition of irradiated fuel rod of VVER-1000 NPP, here is my code.

Code: Select all

[code]set title "VVER-1000"

% --- Fuel pin with central hole, 4.85% enrichment:

pin 1
void   0.075
fuel   0.3785
void   0.3865
clad   0.455
water   

% --- Central tube:

%pin 2 
%water  0.44000
%clad   0.51500
%water   

%---Guide tube

pin 3
water 0.545
clad 0.65
water

% --- Empty lattice position:

pin 4
water   

% --- Lattice (type = 2, pin pitch = 1.23 cm):

lat 10 2 0.0 0.0 23 23 1.275
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
 4 4 4 4 4 4 4 4 4 4 4 1 1 1 1 1 1 1 1 1 1 1 4
  4 4 4 4 4 4 4 4 4 4 1 1 1 1 1 1 1 1 1 1 1 1 4
   4 4 4 4 4 4 4 4 4 1 1 1 1 1 1 1 1 1 1 1 1 1 4
    4 4 4 4 4 4 4 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4
     4 4 4 4 4 4 4 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 4
      4 4 4 4 4 4 1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1 4
       4 4 4 4 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4
        4 4 4 4 1 1 1 1 3 1 1 1 3 1 1 1 1 3 1 1 1 1 4
         4 4 4 1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 4
          4 4 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1 1 4
           4 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 4
            4 1 1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 4 4
             4 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1 4 4 4
              4 1 1 1 1 3 1 1 1 1 3 1 1 1 3 1 1 1 1 4 4 4 4
               4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 4 4 4 4
                4 1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1 4 4 4 4 4 4
                 4 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 4 4 4 4 4 4 4
                  4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 4 4 4 4 4 4 4
                   4 1 1 1 1 1 1 1 1 1 1 1 1 1 4 4 4 4 4 4 4 4 4
                    4 1 1 1 1 1 1 1 1 1 1 1 1 4 4 4 4 4 4 4 4 4 4
                     4 1 1 1 1 1 1 1 1 1 1 1 4 4 4 4 4 4 4 4 4 4 4
                      4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4


% --- Surfaces (assembly pitch = 14.7 cm):

surf 1  hexyc    0.0  0.0  11.675  % Shroud tube inner radius
surf 2  hexyc    0.0  0.0  11.74  % Shroud tube outer radius
surf 3  hexyc    0.0  0.0  11.85  % Outer boundary

% --- Cells:

cell  1  0  fill 10  -1        % Pin lattice                 
cell  4  0  tube      1    -2  % Shroud tube                
cell  5  0  water     2    -3  % Water in channel                
cell 99  0  outside   3        % Outside world                

% --- UO2 fuel enriched to 3.6 wt-% U-235:

mat fuel   -10.45600
92235.09c   -0.0418
92238.09c   -0.8396
 8016.09c   -0.11860

% --- Zr-Nb cladding and shroud tube:

mat clad    -6.55000
40000.06c   -0.99000
41093.06c   -0.01000

mat tube    -6.58000
40000.06c   -0.97500
41093.06c   -0.02500

% --- Water:

mat water1   -0.7207  moder lwtr 1001
 1001.06c    2.0
 8016.06c    1.0

% --- Thermal scattering data for light water:

therm lwtr lwj3.11t

% --- Natural boron (used as soluble absorber):

mat boron    1.0
 5010.06c    0.2
 5011.06c    0.8

% --- 650 ppm soluble absorber in water:

mix water
water1 -0.999350
boron -650E-6

% --- Cross section library file path:

set acelib "/home/kaf1/Desktop/serpent/xsdata/jeff311/sss_jeff31u.xsdata"
set declib "/home/kaf1/Desktop/serpent/xsdata/jeff311/sss_jeff31.dec"
set nfylib "/home/kaf1/Desktop/serpent/xsdata/jeff311/sss_jeff31.nfy"

% --- Periodic boundary condition:

set bc 3

% --- Group constant generation:

% universe = 0 (homogenization over all space)
% symmetry = 12
% 2-group structure (group boundary at 0.625 eV)

set gcu  0
set sym  12
set nfg  2  0.625E-6

% --- Neutron population and criticality cycles:

set pop 2000 500 20

%---Paramter included in coefficient output
set coefpara 1
B1_RABSXS            % Reduced absorption cross section
B1_NSF               % Fission neutron production cross section
B1_DIFFCOEF          % Diffusion coefficient
B1_S0                % Scattering matrix
B1_KAPPA             % Energy deposited per fission (MeV)
B1_FISS              % Fission cross section
B1_INVV              % Inverse neutron speed
B1_I135_MICRO_ABS    % Microscopic absorption cross section of I-135
B1_XE135_MICRO_ABS   % Microscopic absorption cross section of Xe-135
B1_XE135_MACRO_ABS   % Macroscopic absorption cross section of Xe-135
B1_PM149_MICRO_ABS   % Microscopic absorption cross section of Pm-149
B1_I135_YIELD        % Fission yield of I-135
B1_XE135_YIELD       % Fission yield of Xe-135
B1_PM149_YIELD       % Fission yield of Pm-149
B1_SM149_YIELD       % Fission yield of Sm-149
B1_SM149_MICRO_ABS   % Microscopic absorption cross section of Sm-149
B1_SM149_MACRO_ABS   % Macroscopic absorption cross section of Sm-149
BETA_EFF             % Effective delayed neutron fraction
LAMBDA               % Delayed neutron decay constants
DF_SURF_DF           % Surface discontinuity factors
DF_CORN_DF           % Corner discontinuity factors

% --- History variables:

set his 1

branch HIS
var LIB ENDFB7
var BHI 500
var VHI 550
%set rfr 10 vver.wrk

set rfw 1
% --- Geometry and mesh plots:

plot 3 500 500
mesh 3 500 500
% --- Flux per lethargy:

set powdens 42.5E-03
dep butot
30
40
50

dep decstep
365
365
365
365
365



set inventory
922340
922350
922360
922380
922390
932360
932370
932380
932390
942360
942380
942390
942400
942410
942420
942430
952410
952421
952430
962420
962430
962440
962450
962460
962470
962480
962490
972490
972500
982490
982500
982510
982520
360830
451030
451050
471090
531350
541310
541350
551330
551340
551350
551370
561400
571400
601430
601450
611470
611480
611490
611481
621470
621490
621500
621510
621520
631530
631540
631550
631560
641520
641540
641550
641560
641570
641580
641600
[/code]
As I know the composition can be read from the <input>_dep.m but i can't get this file, could you please help me to find out, what is the error here? Thanks in advance
Last edited by Nt009 on Thu Oct 08, 2020 6:01 pm, edited 1 time in total.

Ana Jambrina
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by Ana Jambrina » Wed Oct 07, 2020 12:03 pm

For starters, currently you are performing a criticality calculation: you are not depleting any material: add ‘burn 1’ flag to the material/s you want to deplete, i.e. fuel (see: http://serpent.vtt.fi/mediawiki/index.p ... l#mat_burn)
You might want to consider to divide the burnable materials (see: http://serpent.vtt.fi/mediawiki/index.p ... e_division), and the role of volumes in depletion calculations (see: http://serpent.vtt.fi/mediawiki/index.p ... al_volumes).
Regarding the depletion output, ‘_dep.m’, aside the ‘inventory’ option, you might want to check ‘depout’ (see: http://serpent.vtt.fi/mediawiki/index.p ... set_depout).
- Ana

Ville Valtavirta
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Re: Gamma Transport w/ Spent Fuel: Nuclide not found

Post by Ville Valtavirta » Wed Oct 07, 2020 12:38 pm

Hi,

it might be a good idea to start with the Serpent tutorial first to get to know the basics of Serpent (http://serpent.vtt.fi/mediawiki/index.php/Tutorial), details on burnup calculations are presented in part 4 of the tutorial (http://serpent.vtt.fi/mediawiki/index.p ... alculation).

-Ville

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