Number of Depletion Regions effecting Max NPG / MPI ratio

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cgentry7
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Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Thu Nov 05, 2015 9:02 pm

I've been attempting to run some PWR lattice cases on TITAN and I'm encountering a sort of odd behavior. I find I have a case that runs fine with a given NPG setting and the problem distributed over X amount of MPI tasks, but when I increase the number of depletion zones I find that the runs don't complete and I get the following error message:

Error in normalization: zero fission power

Is this a behavior that I should expect and if so why? Does this mean that some depletion zone is seeing no neutrons and therefore can't deplete?

Thanks,
Cole

cgentry7
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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Fri Nov 06, 2015 3:31 pm

Sorry, I attached an input but forgot to actually list the code in the text.

Cole

Code: Select all


set title "TEST LATTICE"

%  -----------------------------------------------------------------------------
%  Geometry
%  -----------------------------------------------------------------------------

% Symmetry Hole -----------
pin 1
mod

% Fuel Pins ---------------
pin 2
U320     0.409575
he       0.41783
zirlo    0.47498
mod

% Guide Tube --------------
pin 3
mod      0.56134
zirlo    0.61214
mod

% Instrument Tube ---------
pin 4
mod      0.56134
zirlo    0.61214
mod

% Lattice ---------------------------------------------
%     UID  Lattice Type  X0  Y0    NX  NY    Pin Pitch
lat   10        1        0.0 0.0   17  17      1.26

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 4 2 2 3 2 2 3 2 2
1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2
1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2
1 1 1 1 1 1 1 1 1 1 1 3 2 2 3 2 2
1 1 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2
1 1 1 1 1 1 1 1 1 1 1 1 1 3 2 2 2
1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2 2
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2

% --- Geometry Symmetry ----
%Use the following if you wish to apply 1/8 symmetry

%SERPENT 2 FORMAT
%Symmetry axis: x=1 y=2 z=3
%Boundary cond: refl=2 per=3
%t0 - angle where symmetry starts
%tw - width of angle
%         UNIV  AXIS  BC  X0   Y0   t0  tw
set usym   10     3    2  0.0  0.0  0  45

%SERPENT 1 FORMAT
%Symmetry Type is based on number so for 
%half symmetry use 2, for 1/8 use 8 etc.
%         UNIV TYPE  X0   Y0
%set usym   10    8   0.0  0.0


% Global Universe --------------------------------------------------------
% Surface Definitions
%     SID   Type    X0   Y0     Half Pitch
surf   1    sqc     0.0  0.0      10.71         %Lattice Surface
surf   2    sqc     0.0  0.0      10.75         %Assembly Pitch Surface

% Cell Definitions -------------------------------------------------------
% (NOTE: negative indicates inside surface and positive outside surface EX: -2  1  inside surface 2 outside surfae 2 outside surface 1)
%     CID  UID   Cell Material    Cell Geomtry
cell   1    0      fill 10           -1         %Lattice  Cell  (filled by lattice 10)
cell   2    0      mod               -2  1      %Assembly Pitch Cell (fills remainder of assembly cell not filled with lattice 10 with coolant)
cell   3    0      outside            2         %Boundary

% Boundary Conditions ----------------------------------------------------
% 1 - Vacuum
% 2 - Reflective
% 3 - Periodic
%        X   Y   Z
set bc   2   2   2

%  -----------------------------------------------------------------------------
%  Materials
%  -----------------------------------------------------------------------------
% Materials are specified by cross-section library IDs  AAAAA.BBc where AAAAA is the
% isotopic ID and BB the temperature ID.  SERPENT 1 allows for doppler broadening
% correction, and SERPENT 2.1.24 allows for S(a,b) treatment of moderators.
% (NOTE:  "sum"    designates user will provide isotopic number densities in (#/b-cm) )
% (NOTE:  "burn 1" sets the material as a depletable material)
% (NOTE:  "div"  Automatic division of depletion zones
%  div <mat> [sep <lvl>] [subz <nz> <zmin> <zmax>] [subr <nr> <rmin> <rmax>] [subs <ns> <s0>]
%  <lvl> - number of levels counted backwards from the last one (<lvl> = 1 being the last)
%  subz  - Axial subdivision        self explanatory
%  subr  - Radial subdivision       self explanatory
%  subs  - Angular subdivisions     ns - number of angular sectors  s0 - zero position angle
%  EX:  div U31 sep 1  subr 5 0.0 0.41 subs 4 45.0

mat U320    sum  burn 1
92234.09c   6.459018719580362E-06   
92235.09c   7.429822108101351E-04   
92236.09c   1.387179826880089E-07   
92238.09c   2.218483237698008E-02   
8016.09c    4.586859154646811E-02    
div U320 sep 1

mat he     sum
8016.06c   6.726208901497765E-06 

mat zirlo  sum
8016.06c   2.960134488565085E-04
26054.06c  4.128493623884771E-06
26056.06c  6.480854104947671E-05
26057.06c  1.496712136802813E-06
26058.06c  1.991849969909548E-07
40090.06c  2.175317625852084E-02
40091.06c  4.743829090086294E-03
40092.06c  7.251009003390216E-03
40094.06c  7.348253852095207E-03
40096.06c  1.183840596224298E-03
41093.06c  4.245675533096247E-04
50112.06c  3.223206287660899E-06
50114.06c  2.193129202116124E-06
50115.06c  1.129751947104008E-06
50116.06c  4.831349162963977E-05
50117.06c  2.551913131856568E-05
50118.06c  8.047833295931593E-05
50119.06c  2.854286076669262E-05
50120.06c  1.082568595160416E-04
50122.06c  1.538457983881192E-05
50124.06c  1.923903774988889E-05

%(NOTE: "therm" specifies a particular ID, in this case "lwtr", as representing a thermal
% scattering library.  Here it represents the library lwe7.12t, which is the ENDF-7
% Light Water thermal scattering library at 600 K.  Yes, it does end with 12t.
% S(a,b) is available if desired.  "moder" indicates in the material card which cross-section
% should be adjusted by the thermal scattering library, in this case the hydrogen 1001.

therm   lwtr  lwe7.12t
mat mod   sum   moder lwtr 1001
1001.06c  4.970905912557790E-02
8016.06c  2.485452956278895E-02
5010.06c  1.089654094925088E-05
5011.06c  4.385994623291432E-05


%  -----------------------------------------------------------------------------
%  Running Parameters
%  -----------------------------------------------------------------------------
% --- Cross section library file path -----------
set acelib "/lustre/atlas/proj-shared/nfu101/xsdata/sss_endfb7u.xsdata"
set declib "/lustre/atlas/proj-shared/nfu101/xsdata/sss_endfb7.dec"
set nfylib "/lustre/atlas/proj-shared/nfu101/xsdata/sss_endfb7.nfy"

% Optimization Mode -----------------------------
% 4 - (default) highest nemory fastest runtime
% 3 - Good for short burnup cases (some memory improvement)
% 2 - Good for long burnups and many depletion regions (Not good for homogenized xsec generation)
% 1 - lowest memory slowest runtime
set opti 1


% Neutron Histories -----------------------------
%         NPG     ACT   NSK
set pop 4000000   250   100

% Delta Tracking Optimization -------------------
set dt 0.9

% Unresolved Resonance Treatment ----------------
set ures 1

% I Xe Pm Sm XSec Calculation -------------------
set poi 1

% 2D Plots --------------------------------------
% Plot Type Numbers
% 1 - YZ Plot
% 2 - XZ Plot
% 3 - XY Plot
%     Plot Type   X-Pixels  Y-Pixels
plot      3         5000      5000
mesh      3         5000      5000

% Pin Power Output ------------------------------
set ppw  0   10          %NOTE:  0 - Global Universe  10 - Assembly Lattice UID

% Fission Product Yield Cutoff  -----------------
set  fpcut 1.000000e-06

% Stability Cutoff  -----------------------------
set  stabcut 1.000000e-12

% Burnup Mode  ----------------------------------
set bumode 2

% Predictor-Corrector Flag  ---------------------
set pcc 1

% Transmutation X-sec Generation  ---------------
set xscalc 2

% Print depleted material compositions  ---------
set printm 1

% Power Density  (kW/g) -------------------------
set powdens  0.0402 

% Depletion Steps  (GWd/MTHM) -------------------
dep butot

 0.1
 0.5
 1.0
 1.5
 2.0
 2.5
 3.0
 3.5
 4.0
 4.5
 5.0
 5.5
 6.0
 6.5
 7.0
 7.5
 8.0
 8.5
 9.0
 9.5
10.0
11.0
12.0
13.0
14.0
15.0
16.0
17.0
18.0
19.0
20.0
21.0
22.0
23.0
24.0
25.0
26.0
27.0
28.0
29.0
30.0
31.0
32.0
33.0
34.0
35.0
36.0
37.0
38.0
39.0
40.0
42.5
45.0
47.5
50.0
52.5
55.0
57.5
60.0
62.5
65.0
67.5
70.0


% --- Nuclide inventory (Isotopes to Print Out in Output)--------

set inventory
922350 % 235U      % the main fissile nuclides
942390 % 239Pu
541350 % 135Xe     %major fp
621490 % 149Sm     %major fp
551370 % 137Cs     % major medium lived fp
641550 % 155Gd     %Gd fission poisons
641570 % 157Gd
641580 % 158Gd
922340 % 234U  -- other actinides --
922360 % 236U
922370 % 237U
922380 % 238U
922390 % 239U
932390 % 239Np
942380 % 238Pu
942400 % 240Pu
942410 % 241Pu
942420 % 242Pu
952410 % 241Am
952421 % 242mAm
952430 % 243Am
962420 % 242Cm
962440 % 244Cm
962450 % 245Cm
441010 % 101Ru          -- fr poisons--                         hl: stable
451030 % 103Rh                                                  hl: stable
551330 % 133Cs                                                  hl: stable
430990 % 99Tc                                                   hl 211k a
390900 % 90Y            --short-lived products--                hl 64h
420990 % 99Mo                                                   hl: 2.75 d
531310 % 131I                                                   hl: 8d
561371 % 137mBa                                                 hl: 2.6 min, 137Ba is stable
380900 % 90Sr           --medium-lived products--               hl 28.8 a
551340 % 134Cs                                                  hl: 2.07 a
621510 % 151Sm                                                  hl: 90 a   -fission poison-
340790 % 79Se           --long-lived products-                  hl: 327k a
400930 % 93Zr                                                   hl: 1.53M a
531290 % 129I                                                   hl: 15.7M a
551350 % 135Cs                                                  hl  2.3M a
471090 % 109Ag          --very long-lived products--            hl: stable
601430 % 143Nd                                                  hl: stable
621470 % 147Sm                                                  hl: 10000M a
621500 % 150Sm                                                  hl: stable


%  -----------------------------------------------------------------------------
%  Detectors
%  -----------------------------------------------------------------------------
%  Although PPW can provide the pin powers, you can also use detectors.  Detectors
%  are nice because you can designate a variety of response functions and are not
%  just limited to pin power.  They results are by default volume integrated.
%
%  Common Response Functions:
%
%  -1 Total
%  -2 Total Capture
%  -3 Total Elastic
%  -5 Total (n,2n)
%  -6 Total Fission
%  -7 Total Fission Neutron Production
%  -8 Total Fission Energy Deposition (gamma contribution approximated)
% -15 Total Neutrons (analogous to flux)
%
%  Parameters:
%
%  dr  Reaction Multiplier (Response Function Designator)
%  dv  Detector Volume
%  dc  Detector Cell
%  du  Detector Universe
%  dm  Detector Material
%  dl  Detector Lattice
%  de  Detector Energy Grid
%  dx / dy / dz  Detector Mesh
%  dt  Detector Type
%  ds  Surface Current Detector

% Fast / Thermal Energy Group Structure
ene 1  1  1.000000e-12 0.625E-6 2.000000e+01

% SCALE-238 Group Structure for Spectrum Comparison
ene 2  4  scale238

det 1  de 1  dl 10  dr -6   void   % Total Fission Energy Generated in Fuel Pin 3 in each lattice cell of Lattice 10 with Fast / Thermal Energy bins
det 2  de 1  dl 10  dr -8   void   % Total Fission Energy Deposited in Fuel Pin 3 in each lattice cell of Lattice 10 with Fast / Thermal Energy bins
det 3  de 1  dl 10  dr -15  void   % Total Neutrons in Fuel Pin 3 in each lattice cell of Lattice 10 with Fast / Thermal Energy bins
det 4  dt -2 de 2                         % Flux per MeV
det 5  dt -3 de 2                         % Flux per Lethargy


% Volumes of depleted materials must best set manually due to symmetry
set mvol

U320           39 2.10804E+00
U320           38 4.21608E+00
U320           37 2.10804E+00
U320           36 4.21608E+00
U320           35 4.21608E+00
U320           34 2.10804E+00
U320           33 4.21608E+00
U320           32 4.21608E+00
U320           31 4.21608E+00
U320           30 4.21608E+00
U320           29 4.21608E+00
U320           28 4.21608E+00
U320           27 4.21382E+00
U320           26 2.10804E+00
U320           25 4.21608E+00
U320           24 4.21608E+00
U320           23 4.21608E+00
U320           22 4.21608E+00
U320           21 4.21608E+00
U320           20 4.21608E+00
U320           19 4.21608E+00
U320           18 4.21608E+00
U320           17 4.21608E+00
U320           16 4.21608E+00
U320           15 2.10804E+00
U320           14 4.21608E+00
U320           13 4.21608E+00
U320           12 4.21608E+00
U320           11 4.21608E+00
U320           10 4.21608E+00
U320            9 4.21608E+00
U320            8 4.21608E+00
U320            7 2.10804E+00
U320            6 2.10804E+00
U320            5 2.10804E+00
U320            4 2.10804E+00
U320            3 2.10804E+00
U320            2 2.10804E+00
U320            1 2.10804E+00


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Jaakko Leppänen
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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by Jaakko Leppänen » Fri Nov 06, 2015 6:36 pm

No, you should get this error message only if there are no contributions at all to fission rate.

Have you tried running the calculation in debug mode? Have you checked the material volumes? (see related post)
- Jaakko

cgentry7
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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Sun Nov 08, 2015 4:27 pm

Yes, I just tried running in debug mode, and the volumes I originally approximated using the approximator and then corrected to the precise value post-hoc. So the volumes should be good.

Cole

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Jaakko Leppänen
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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by Jaakko Leppänen » Mon Nov 09, 2015 1:04 am

Does the calculation run OK with fewer MPI tasks?
- Jaakko

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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Mon Nov 09, 2015 3:24 pm

Yes, it does. The results appear to be appropriate. If I turn depletion off I can run with many more MPI tasks. It is only when I'm depleting that this becomes a problem.

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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by Jaakko Leppänen » Wed Nov 11, 2015 1:54 am

Can you check the material compositions before the calculation terminates? I think you can get this error if the fissile content falls too low, which can happen if there's some problem with volumes or normalization.

Or do you get this error right from the first step?
- Jaakko

cgentry7
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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Wed Nov 11, 2015 4:16 pm

I get this immediately. Usually before 20 or so skipped generations. If I run about 500 nodes with 16 cores per node it runs fine. When I go back up to 1200 nodes with 16 cores per node, the simulation dies before getting through the skipped generations. If I run with no depletion, I can run 1200 nodes just fine. It's only when depletion is turned on, and the job dies before any depletion has occurred.

Cole

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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by Jaakko Leppänen » Wed Nov 11, 2015 6:17 pm

OK, so it's not related to MPI communication because the tasks don't exchange data until the end of the transport cycle. What kind of depletion zone subdivision are you using when the calculation crashes?
- Jaakko

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Re: Number of Depletion Regions effecting Max NPG / MPI ratio

Post by cgentry7 » Wed Nov 11, 2015 8:12 pm

I'm using "sep 1", so only single depletion regions for each pin, but each pin depleted independently.

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