60-Nd-145 JAEA+ EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata
DIST-MAY10 20100119
----JENDL-4.0 MATERIAL 6034
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
09-12 The resolved resonance parameters were evaluated by
A.Zukeran,K.Shibata.
The data above the resolved resonance region were evaluated
and compiled by N.Iwamoto.
MF= 1 General information
MT=451 Descriptive data and directory
MF= 2 Resonance parameters
MT=151 Resolved and unresolved resonance parameters
Resolved resonance region (MLBW formula) : below 4 keV
For JENDL-2, resonance energies were taken from Tellier
/1/, and after calibration, data of Rohr et al./2/ and
Musgrove et al./3/ were adopted for the levels not
measured by Tellier. Neutron widths were adopted from
Tellier, and radiation widths were obtained from the
capture areas measured by Rohr et al. and Musgrove et al.
The average radiation width of 0.087 eV was assumed for the
resonances whose capture area was not measured, and to
estimate neutron widths from the capture areas for the
resonances not measured by Tellier. A negative resonance
was added so as to reproduce the thermal capture and total
cross sections given by Mughabghab et al./4/
For JENDL-3, total spin j of some resonances was tentative-
ly estimated with a random number method.
For JENDL-3.2, the capture data measured at ORELA of ORNL
were renormalized (factor = 0.9507)/5/. The neutron width
and/or the radiation width was revised to reproduce the
renormalized capture area for each resonance above 2.592
keV.
In JENDL-4, the data for 4.36 - 497.87 eV were replaced with
the ones obtained by Barry et al./6/ The parameters for
the negative resonance were adjusted so as to reproduce
the thermal capture cross section recommended by Mughabghab
/7/.
Unresolved resonance region : 4.0 keV - 200.0 keV
The unresolved resonance paramters (URP) were determined by
ASREP code /8/ so as to reproduce the evaluated total and
capture cross sections calculated with optical model code
OPTMAN /9/ and CCONE /10/. The unresolved parameters
should be used only for self-shielding calculation.
Thermal cross sections and resonance integrals at 300 K
----------------------------------------------------------
0.0253 eV res. integ. (*)
(barn) (barn)
----------------------------------------------------------
Total 6.9165e+01
Elastic 1.9711e+01
n,gamma 4.9455e+01 2.2269e+02
n,alpha 8.0333e-05
----------------------------------------------------------
(*) Integrated from 0.5 eV to 10 MeV.
MF= 3 Neutron cross sections
MT= 1 Total cross section
Sum of partial cross sections.
MT= 2 Elastic scattering cross section
Obtained by subtracting non-elastic scattering cross sections
from total cross section.
MT= 4 (n,n') cross section
Calculated with CCONE code /10/.
MT= 16 (n,2n) cross section
Calculated with CCONE code /10/.
MT= 17 (n,3n) cross section
Calculated with CCONE code /10/.
MT= 22 (n,na) cross section
Calculated with CCONE code /10/.
MT= 24 (n,2na) cross section
Calculated with CCONE code /10/.
MT= 28 (n,np) cross section
Calculated with CCONE code /10/.
MT= 32 (n,nd) cross section
Calculated with CCONE code /10/.
MT= 33 (n,nt) cross section
Calculated with CCONE code /10/.
MT= 41 (n,2np) cross section
Calculated with CCONE code /10/.
MT= 51-91 (n,n') cross section
Calculated with CCONE code /10/.
MT=102 Capture cross section
Calculated with CCONE code /10/.
MT=103 (n,p) cross section
Calculated with CCONE code /10/.
MT=104 (n,d) cross section
Calculated with CCONE code /10/.
MT=105 (n,t) cross section
Calculated with CCONE code /10/.
MT=106 (n,He3) cross section
Calculated with CCONE code /10/.
MT=107 (n,a) cross section
Calculated with CCONE code /10/.
MT=108 (n,2a) cross section
Calculated with CCONE code /10/.
MF= 4 Angular distributions of emitted neutrons
MT= 2 Elastic scattering
Calculated with CCONE code /10/.
MF= 6 Energy-angle distributions of emitted particles
MT= 16 (n,2n) reaction
Calculated with CCONE code /10/.
MT= 17 (n,3n) reaction
Calculated with CCONE code /10/.
MT= 22 (n,na) reaction
Calculated with CCONE code /10/.
MT= 24 (n,2na) reaction
Calculated with CCONE code /10/.
MT= 28 (n,np) reaction
Calculated with CCONE code /10/.
MT= 32 (n,nd) reaction
Calculated with CCONE code /10/.
MT= 33 (n,nt) reaction
Calculated with CCONE code /10/.
MT= 41 (n,2np) reaction
Calculated with CCONE code /10/.
MT= 51-91 (n,n') reaction
Calculated with CCONE code /10/.
MT=102 Capture reaction
Calculated with CCONE code /10/.
*****************************************************************
Nuclear Model Calculation with CCONE code /10/
*****************************************************************
Models and parameters used in the CCONE calculation
1) Optical model
* coupled channels calculation
coupled levels: 0,4 (see Table 1)
* optical model potential
neutron omp: Kunieda,S. et al./11/ (+)
proton omp: Koning,A.J. and Delaroche,J.P./12/
deuteron omp: Lohr,J.M. and Haeberli,W./13/
triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./14/
He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./14/
alpha omp: McFadden,L. and Satchler,G.R./15/
(+) omp parameters were modified.
2) Two-component exciton model/16/
* Global parametrization of Koning-Duijvestijn/17/
was used.
* Gamma emission channel/18/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Width fluctuation correction/19/ was applied.
* Neutron, proton, deuteron, triton, He3, alpha and gamma
decay channel were taken into account.
* Transmission coefficients of neutrons were taken from
optical model calculation.
* The level scheme of the target is shown in Table 1.
* Level density formula of constant temperature and Fermi-gas
model were used with shell energy correction/20/.
Parameters are shown in Table 2.
* Gamma-ray strength function of generalized Lorentzian form
/21/,/22/ was used for E1 transition.
For M1 and E2 transitions the standard Lorentzian form was
adopted. The prameters are shown in Table 3.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Level Scheme of Nd-145
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 7/2 - *
1 0.06722 3/2 -
2 0.07250 5/2 -
3 0.65767 11/2 -
4 0.74828 9/2 - *
5 0.78045 3/2 -
6 0.91983 1/2 -
7 0.92072 9/2 -
8 0.93705 5/2 -
9 1.01122 11/2 +
10 1.05141 5/2 -
11 1.08525 3/2 +
12 1.11120 13/2 +
13 1.15026 7/2 -
14 1.16105 5/2 +
15 1.16232 9/2 -
16 1.21370 1/2 -
17 1.24973 5/2 -
18 1.28560 5/2 -
19 1.31680 3/2 -
20 1.32630 1/2 +
21 1.33860 7/2 -
22 1.40090 3/2 -
23 1.40130 15/2 -
24 1.40392 5/2 -
25 1.42760 13/2 -
-------------------
*) Coupled levels in CC calculation
Table 2. Level density parameters
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Nd-146 18.1900 1.9863 1.6792 0.5692 0.1138 6.4542
Nd-145 18.5400 0.9965 1.1101 0.5235 -0.2928 4.6189
Nd-144 17.5000 2.0000 0.3419 0.6111 0.2496 6.6190
Nd-143 17.7000 1.0035 -0.4179 0.5516 0.0353 4.4179
Nd-142 15.0000 2.0140 -1.2557 0.6895 0.7987 6.4278
Pr-145 16.8637 0.9965 1.7883 0.6002 -0.8883 5.5766
Pr-144 15.5000 0.0000 0.9153 0.6715 -1.9662 5.0412
Pr-143 16.6639 1.0035 0.4682 0.6161 -0.5920 5.4208
Pr-142 16.4000 0.0000 -0.4377 0.7390 -2.6336 6.4135
Pr-141 16.4637 1.0106 -1.2280 0.6590 -0.3966 5.5793
Ce-144 17.4894 2.0000 1.0129 0.5822 0.3675 6.2813
Ce-143 19.6000 1.0035 0.4100 0.4774 0.1189 3.9645
Ce-142 18.9500 2.0140 -0.3155 0.5558 0.6875 5.9346
Ce-141 17.9000 1.0106 -1.0773 0.4985 0.5829 3.4550
Ce-140 17.0742 2.0284 -1.9470 0.5674 1.4861 4.9920
Ce-139 15.5000 1.0178 -1.1255 0.5922 0.4151 4.0889
--------------------------------------------------------
Table 3. Gamma-ray strength function for Nd-146
--------------------------------------------------------
* E1: ER = 14.74 (MeV) EG = 5.78 (MeV) SIG = 310.00 (mb)
* M1: ER = 7.79 (MeV) EG = 4.00 (MeV) SIG = 1.01 (mb)
* E2: ER = 11.96 (MeV) EG = 4.36 (MeV) SIG = 3.37 (mb)
--------------------------------------------------------
References
1) Tellier, H.: CEA-N-1459 (1971).
2) Rohr, G., et al.: "Proc. 3rd Conf. on Neutron Cross Sections
and Technology, Knoxville 1971", Vol. 2, 743.
3) Musgrove, A.R. de L., et al.: AEEC/E401 (1977).
4) Mughabghab, S.F.: "Neutron Cross Sections, Vol. I, Part B",
Academic Press (1984).
5) Allen, B.J. et al.: Nucl. Sci. Eng., 82, 230 (1982).
6) Barry, D.P. et al.: Nucl. Sci. Eng., 153, 8 (2006).
7) Mughabghab, S.F.: "Atlas of Neutron Resonances", Elsevier
(2006).
8) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)
[in Japanese].
9) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).
10) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).
11) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).
12) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)
[Global potential].
13) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).
14) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.
J.H.Williams Lab., Univ. Minnesota (1969).
15) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966).
16) Kalbach,C.: Phys. Rev. C33, 818 (1986).
17) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).
18) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).
19) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).
20) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151
(1994).
21) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).
22) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).