SCALE Software Coordinator
This topic has been migrated from the SCALE 5 ORIGEN notebook.
Date: Wed May 9 17:38:02 2007
Dear Sir
Using ORIGEN-S, ARP in Scale5.1.
I attached 3 inputfile of ORIGEN.
(1) ORIGEN-ARP fuel burnup calculation with POWER input
(2) ORIGEN-ARP fuel burnup calculation with FLUX input
(3) ORIGEN-S fuel burnup calculation with FLUX & FAST,RES,THERM input Material is UO2 and some nuclides(Co59,Sn124).
After 500day irradiation, weight(gram) is
--------------------------------------
(1) (2) (3) (4)
--------------------------------------
Co60 4.3e-2 4.3e-2 4.6e-2 3.5e-2
Sb125 1.6e-3 1.6e-3 6.3e-5 1.5e-3
Pu239 3.5e+3 3.5e+3 4.3e+3 3.5e+3
--------------------------------------
( (4) is results of ORIGEN-2(JENDL3.2 Lib) )
Results of (1),(2) are almost equal.
Co60 & Pu239 are near between (1) & (3).
But, results of Sb125 weight(gram) have large difference between (1) & (3).
THERM=0.5, RSE=4.0, FAST=2.0 at (3) are evaluated by another
1-D transport calculation.
Why is Sb125 of (3) different from others?
Advice, Please.
Thanks,
Yusuke, M
======(1) ORIGEN-ARP (input POWER) ========
'This SCALE input file was generated by
'OrigenArp Version 5.1 October 27, 2006
=arp
ge8x8-4
3.44
1
500
24.9
1
0.44
ft33f001
end
#origens
0$$ a4 33 a11 71 e
t
test
3$$ 33 a3 1 a16 2 e
t
35$$ 0
t
56$$ 5 5 a13 4 a15 3 a18 1 e
57** 0 a3 1e-08 1 e
t
test
test
58** 5r24.9
60** 100 200 300 400 500
66$$ a1 2 a5 2 a9 2 e
73$$ 922350 922380 270590 501240
74** 34400 965600 1 1
75$$ 2 2 1 1
t
56$$ f0
t
end
======(2) ORIGEN-ARP (input FLUX) ========
'This SCALE input file was generated by
'OrigenArp Version 5.1 October 27, 2006
=arp
ge8x8-4
3.44
1
500
24.9
1
0.44
ft33f001
end
#origens
0$$ a4 33 a11 71 e
t
test
3$$ 33 a3 1 a16 2 e
t
35$$ 0
t
56$$ 5 5 1 a13 4 a15 3 a18 1 e
57** 0 a3 1e-08 1 e
t
test
test
59** 5r3.58E+13
60** 100 200 300 400 500
66$$ a1 2 a5 2 a9 2 e
73$$ 922350 922380 270590 501240
74** 34400 965600 1 1
75$$ 2 2 1 1
t
56$$ f0
t
end
======(3) ORIGEN-S (input FLUX & THERM,RES,FAST) ========
=origens
0$$ a5 28 e
1$$ 1
t
test
3$$ a16 2 e
4** 0.5 4.0 2.0 e
t
35$$ 0
t
56$$ 5 5 1 a13 4 a15 3 a18 1 e
57** 0 a3 1e-08 1 e
t
test
test
59** 5r3.58E+13
60** 100 200 300 400 500
66$$ a1 2 a5 2 a9 2 e
73$$ 922350 922380 270590 501240
74** 34400 965600 1 1
75$$ 2 2 1 1
t
56$$ f0
t
end
SCALE Software Coordinator
Dear M Yusuke
Some differences like that observed for Pu239 are expected since the use of the card-image (3-group) library is very approximate, whereas the GE8X8-4 library is self shielded and collapsed with a problem-dependent spectrum. However, the differences you observed for Sb-125 are much larger than I would expect. The values for THERM, RES, and FAST you are using a very reasonable. Can you confirm that the cause is the capture cross section for Sn124? This can be done by editing the cross section library using nn3=1 (3$ array). This should identify if the cause is the thermal or resonance values. If this is the cause of difference, then it is possible that the source of the evaluated cross sections is different. Note that the basic 3-group cross sections in card image library (see file origen.rev02.pwrlib.data that indicates source) for Sn124 is ENDF/B-VI. The SCALE transport library used to generate problem-dependent cross sections for ORIGEN-ARP libraries is the 44-group library, based on ENDF/B-V. It is possible there were significant revisions in the ENDF/B-VI release, but this needs to be confirmed. Thank you for your assistance in tracking down the differences.
Best regards
Ian Gauld
ORIGEN-S Code Manager
SCALE Software Coordinator
Dear Ian Gauld
Thanks for your advice.
I confirmed Sb124 capture cross section by using nn1=1(3$ array) for 2 cases.
(2) ORIGEN-ARP fuel burnup calculation with FLUX input
(3) ORIGEN-S fuel burnup calculation with FLUX & FAST,RES,THERM input
I attach one text file, that is a part of output file.
In case(2), Sn124(n,g)Sn125 cross section 0.03683
Sn124(n,g)Sn125m cross section 1.195
In case(3), Sn124 capture cross section 0.0479
And, in 3 group cross section in card image,
ID=501240 MT=102 ,
sig(0)=0.0038887 , sig(res) =0.010541 , sig(mve)=0.0018867
And, in S.F. Mughabghab :"Neutron Cross section"
average cross section for thermal flux (maxwell distibution) is
Sn124(n,g)Sn125 cross section 0.004
Sn124(n,g)Sn125m cross section 0.130
This 0.004 is similar to sig(0)=0.0038887 of 3 group cross section.
Does this sig(0) include Sn124(n,g)Sn125m cross section ?
Best regards
Yusuke, M
SCALE Software Coordinator
Dear Yusuke,
Thanks you for your helpful analysis. I believe you have identified the problem. The card-image library cross sections for Sn-124 appear to include only the transition to the ground state. It should include the total capture cross section to both states. The problem occurred in the processing of pointwise cross section evaluations from FENDL-2.0 library for the LWR card-image library due to use of a non-standard ENDF-6 format to identify different reaction channels. This resulted in partial cross sections to some metastable states being missed. For the ORIGEN-ARP 8x8-4 and other binary libraries, the cross sections have been further updated for the roughly 250 isotopes in ENDF/B-V, and most problematic FENDL data (including Sn-124) are therefore replaced with the correct data. However, there is a potential that some other materials (not in ENDF/B evaluations) could be incorrect when transitions to metastable states are involved. The data are currently being reviewed and will be updated in the near future.
Thank you for identifying the deficiency and bringing it to our attention.
Best regards
Ian