SCALE 6: RES, THERM, FAST? One Group Cross Section Library Generation using COUPLE

[Matt Z]

Hello All, 

I am attempting to use use ORIGEN-S to perform irradiation and decay calculations for samples that are irradiated in a number of flux traps/facilities in a research reactor. I need to develop a problem-dependent one-group neutron cross section binary library for each of the irradiation facilities in the reactor. 

It seems that previous versions of SCALE had an ORIGEN-S input (4** array) for the THERM, RES, and FAST factors that would adjust the cross section library to better approximate the flux spectrum in a given problem. SCALE 6 does not seem to have that feature, but I instead need to use the 9** array and input multigroup weighting function values (NWGT). Where do I get these NWGTvalues? 

I intend to use an MCNP model of the research reactor to determine the local fluxes in the various locations of interest. Do I simply run a 238 group flux tally on the cell of interest and use these results on the 9** card? Do they have to be normalized to the total flux in the cell? What order do they need to be entered on the 9** card (low energy to high?)?

Can anyone clarify the use of this feature in COUPLE? The input description wasn't entirely clear on these issues. 

Also, does this seem like an appropriate approach to the problem I am trying to solve? If not, what would you recommend?

Thank you for your time.

Matt

[Ian Gauld]

Matt, the 3-group flux weighting parameters have been replaced with a more detailed and accurate energy-dependent fine-group weighting based on the actual flux energy spectrum in the material being irradiated. These values can be calculated using transport modules in SCALE (xsdrn, newt, keno, mavric, etc.) but can just as easily be generated by other codes like MCNP. The only requirement is the flux tally must be in the energy group structure of the ORIGEN library being used (e.g., 44, 200, 238 group etc.). There is no normalization requirement. The weighting is used to determine the RELATIVE importance of each energy group so any normalization can be used. The weighting values in the 9* array are simply the neutron flux values in each group from a flux tally. Per the manual, these entries are in descending energy order (high to low). Note that COUPLE prints the energy groups and the weighting so you can verify this order is correct. Also, a quick way to get the energy boundaries is to run a trial COUPLE run setting the 9* entries to any value (e.g., 9** F1 for fill with "1"s) to get the boundary printout. We routinely use this technique to generate problem-dependent activation libraries for ORIGEN. One word of caution -- the cross sections with this technique are not self shielded (i.e., they are infinite dilute values). If you have heavy actinides or other highly absorbing resonance materials then self shielding can be important (but usually not for most activation problems). This can be easily done by prepending a short CSAS-1X case that will perform the self shielding and these cross sections will be used. This case also performs a transport calculation and this flux solution can be read automatically by COUPLE. If you select the CSAS-MG module the self shielding is performed but not the transport. Below is an input example we have used that describes the CSAS-1X method (example for activation of a pressure vessel configuration) but using a flux solution (weighting flux) supplied by the user. In the example all self shielded isotopes in zone 3 (PV zone) are applied to the ORIGEN library. Hope this is helpful. Regards

Ian Gauld

=csas1x   parm=centrm
problem PV activation (cylindrical geom)
v7-238
read comp
uo2   1 den=10.5 0.3 600 92235 4 92238 96 end
zirc4 1 den=6.56 0.1 600 end
h2o   1 den=1    0.6 600 end
h2o   2 end
ss304 3 end
end comp
read celldata
multiregion cylindrical right_bdy=vacuum end
  1 100.0
  2 120.0
  3 140.0
 end zone
moredata icon=zone wgt=1 end moredata
end celldata
end

=couple
********************************************************************************
*                cross sections from 238-group JEFF-3.0/A                      *
*                  collapsed using pressure vessel spectrum (centrm)           *

0$$ a3 80 a4 21 4 a6 33 e
1$$ a13 3 a18 238 e t

9** 
2.4750E-09 6.9710E-06 1.3400E-05 1.6370E-05 4.3790E-05 5.9970E-04
1.4000E-03 4.7990E-03 1.1730E-02 5.9090E-03 2.4080E-02 2.1140E-02
7.6170E-03 3.3260E-02 3.4640E-02 1.1610E-02 6.2980E-03 5.1060E-03
9.7160E-03 9.1130E-03 2.2890E-02 1.9890E-02 2.2870E-02 4.7890E-03
7.0020E-03 3.0490E-03 1.2440E-02 2.0580E-02 3.3490E-02 3.2600E-03
4.6260E-02 1.6650E-02 1.2950E-02 2.6390E-02 1.4210E-02 1.7480E-02
1.0390E-02 8.9670E-03 5.7100E-02 5.4630E-02 5.7740E-02 5.4680E-02
3.3660E-02 3.1350E-02 1.6440E-02 5.1490E-03 1.3190E-02 2.1550E-03
2.6280E-02 1.1080E-02 2.0000E-03 1.3070E-02 1.8080E-02 2.9850E-02
3.4170E-02 1.1670E-02 1.6410E-02 4.8270E-03 8.5430E-03 1.4980E-02
1.5790E-03 9.8710E-03 6.3660E-03 3.1690E-03 9.9860E-04 8.5190E-03
7.7880E-03 1.7200E-03 1.3810E-02 7.0480E-03 1.4300E-02 8.2160E-04
8.3100E-03 1.6980E-02 2.1000E-03 6.4730E-03 5.2520E-03 4.7200E-04
2.9580E-03 1.3520E-03 1.6420E-02 9.5900E-04 1.3140E-03 2.4070E-03
2.9380E-03 3.9990E-03 3.5110E-03 9.2730E-04 1.9220E-03 2.0190E-03
2.4010E-03 1.3990E-03 2.3460E-03 1.2260E-03 3.6520E-03 9.6910E-04
9.9200E-04 1.0180E-03 6.6860E-04 9.8620E-04 1.4070E-03 9.6740E-04
1.3280E-03 1.2000E-03 1.2370E-03 4.5110E-04 1.0120E-03 9.4410E-04
1.4600E-03 9.0290E-04 8.7180E-04 5.2270E-04 1.6100E-03 5.5220E-04
1.4150E-03 2.9960E-03 3.2490E-03 3.5490E-03 2.3000E-03 1.6130E-03
1.6850E-03 8.7220E-04 2.7430E-03 1.9470E-03 1.8430E-03 1.4990E-03
1.4490E-03 1.9840E-03 2.4780E-03 1.0400E-03 4.1850E-03 2.7640E-03
3.3420E-03 3.4950E-03 5.8370E-04 9.9720E-04 1.0260E-03 1.0570E-03
1.0900E-03 2.7670E-03 1.9770E-03 1.2960E-03 4.2160E-03 1.6590E-03
1.4820E-03 2.3920E-03 7.1630E-04 3.6500E-04 2.6790E-04 8.5920E-04
7.7980E-04 7.8300E-04 8.1640E-04 8.3880E-04 7.7600E-04 7.0660E-04
8.1620E-04 8.3770E-04 1.1640E-03 5.9850E-04 8.1770E-04 9.4940E-04
9.8370E-04 1.0210E-03 1.0650E-03 6.0960E-04 6.2380E-04 6.3900E-04
6.5490E-04 6.7330E-04 3.4330E-04 3.4790E-04 3.5340E-04 3.5830E-04
1.4480E-04 1.4550E-04 1.4620E-04 1.4720E-04 1.4830E-04 1.4900E-04
1.4980E-04 1.5080E-04 1.5190E-04 1.5270E-04 1.5350E-04 1.5460E-04
1.5570E-04 1.5690E-04 1.5760E-04 3.9940E-04 4.0570E-04 4.1240E-04
4.1880E-04 8.6470E-04 9.0010E-04 9.3900E-04 9.8610E-04 1.0360E-03
5.3760E-04 5.5290E-04 1.1530E-03 1.2340E-03 1.3150E-03 1.4250E-03
7.6240E-04 8.0120E-04 8.3950E-04 9.1260E-04 9.8430E-04 1.1370E-03
1.3610E-03 1.7580E-03 2.5510E-03 3.9600E-03 6.3780E-03 1.1070E-02
6.3860E-03 7.9000E-03 9.3450E-03 1.0890E-02 1.2250E-02 1.3010E-02
1.2900E-02 5.4830E-03 1.3750E-02 1.1970E-03 7.9640E-04 2.0980E-04
1.5140E-04 5.5100E-05 4.2230E-05 3.0230E-05 1.2820E-05 6.5010E-06
6.0000E-06 3.8490E-06 2.4350E-06 6.9750E-08 t
done
end

[Jordan A]

Ian, I seem to be running into a similar issue. I'm trying to use my flux, and you note that no normalization is needed, and yet your flux values are all super tiny. I just put in my flux magnitude for each energy group (10^9-13), and COUPLE crashes out without any error messages. I'm guessing that I need to have flux weightings instead, but how do I accomplish this given a grouped set of actual flux values?