60-Nd-143 JAEA+ EVAL-Dec09 N.Iwamoto,A.Zukeran,K.Shibata
DIST-MAY10 20100119
----JENDL-4.0 MATERIAL 6028
-----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 5 keV
For JENDL-2, resonance energies were adopted from Tellier
/1/, and those not measured by Tellier were taken from
Rohr et al./2/ and Musgrove et al./3/ after
normalization to Tellier's data. Radiation widths were
derived from capture areas measured by Rohr et al. below 2
keV and Musgrove et al. above 2.5 keV; for the resonances
not measured by Tellier, neutron widths were determined from
capture areas by assuming the average radiation widths of
0.077 eV for s-wave resonances and 0.085 eV for p-wave ones.
Scattering radius was determined from systematics of
measured values. A negative resonance was added at -6 eV so
as to reproduce the capture cross section of 325+-10 barns
compiled by Mughabghab et al./4/
For JENDL-3, total spin J of some resonances was estimated
with a random number method.
For JENDL-3.2, these resonance parameters were modified so
as to reproduce the capture area data measured at ORNL, by
taking account of the correction factor (0.9507) announced
by Allen et al./5/ The parameters of a negative
resonance and scattering radius were adjuseted to get better
agreement with recommended thermal cross sections/4/.
In JENDL-4, the data for 55.4 - 446.5 eV were replaced with
the ones obtained by Barry et al./6/
Unresolved resonance region : 5.0 keV - 200.0 keV
The unresolved resonance paramters (URP) were determined by
ASREP code /7/ so as to reproduce the evaluated total and
capture cross sections calculated with optical model code
OPTMAN /8/ and CCONE /9/. 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 4.0821e+02
Elastic 8.3074e+01
n,gamma 3.2511e+02 1.2850e+02
n,alpha 2.2186e-02
----------------------------------------------------------
(*) 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 /9/.
MT= 16 (n,2n) cross section
Calculated with CCONE code /9/.
MT= 17 (n,3n) cross section
Calculated with CCONE code /9/.
MT= 22 (n,na) cross section
Calculated with CCONE code /9/.
MT= 24 (n,2na) cross section
Calculated with CCONE code /9/.
MT= 28 (n,np) cross section
Calculated with CCONE code /9/.
MT= 32 (n,nd) cross section
Calculated with CCONE code /9/.
MT= 41 (n,2np) cross section
Calculated with CCONE code /9/.
MT= 51-91 (n,n') cross section
Calculated with CCONE code /9/.
MT=102 Capture cross section
Calculated with CCONE code /9/.
MT=103 (n,p) cross section
Calculated with CCONE code /9/.
MT=104 (n,d) cross section
Calculated with CCONE code /9/.
MT=105 (n,t) cross section
Calculated with CCONE code /9/.
MT=106 (n,He3) cross section
Calculated with CCONE code /9/.
MT=107 (n,a) cross section
Calculated with CCONE code /9/.
MT=112 (n,pa) cross section
Calculated with CCONE code /9/.
MF= 4 Angular distributions of emitted neutrons
MT= 2 Elastic scattering
Calculated with CCONE code /9/.
MF= 6 Energy-angle distributions of emitted particles
MT= 16 (n,2n) reaction
Calculated with CCONE code /9/.
MT= 17 (n,3n) reaction
Calculated with CCONE code /9/.
MT= 22 (n,na) reaction
Calculated with CCONE code /9/.
MT= 24 (n,2na) reaction
Calculated with CCONE code /9/.
MT= 28 (n,np) reaction
Calculated with CCONE code /9/.
MT= 32 (n,nd) reaction
Calculated with CCONE code /9/.
MT= 41 (n,2np) reaction
Calculated with CCONE code /9/.
MT= 51-91 (n,n') reaction
Calculated with CCONE code /9/.
MT=102 Capture reaction
Calculated with CCONE code /9/.
*****************************************************************
Nuclear Model Calculation with CCONE code /9/
*****************************************************************
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./10/ (+)
proton omp: Koning,A.J. and Delaroche,J.P./11/
deuteron omp: Lohr,J.M. and Haeberli,W./12/
triton omp: Becchetti Jr.,F.D. and Greenlees,G.W./13/
He3 omp: Becchetti Jr.,F.D. and Greenlees,G.W./13/
alpha omp: McFadden,L. and Satchler,G.R./14/ (+)
(+) omp parameters were modified.
2) Two-component exciton model/15/
* Global parametrization of Koning-Duijvestijn/16/
was used.
* Gamma emission channel/17/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Width fluctuation correction/18/ 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/19/.
Parameters are shown in Table 2.
* Gamma-ray strength function of generalized Lorentzian form
/20/,/21/ 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-143
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 7/2 - *
1 0.74205 3/2 -
2 1.22804 13/2 +
3 1.30586 1/2 -
4 1.40708 9/2 - *
5 1.43123 11/2 -
6 1.50600 5/2 +
7 1.55554 5/2 -
8 1.55644 3/2 +
9 1.55880 9/2 +
10 1.60838 1/2 +
11 1.69000 5/2 +
12 1.73921 9/2 -
13 1.77485 1/2 +
14 1.79952 3/2 +
15 1.85150 7/2 -
16 1.85256 3/2 -
17 1.90030 7/2 -
18 1.91081 5/2 -
19 1.92060 5/2 -
20 1.96600 3/2 +
21 1.98822 11/2 -
22 1.99640 5/2 +
23 2.00467 1/2 -
24 2.01130 9/2 +
25 2.01887 15/2 -
26 2.01920 7/2 -
27 2.03560 7/2 -
28 2.06385 9/2 +
29 2.06684 13/2 -
30 2.07400 5/2 +
31 2.07513 11/2 -
32 2.09060 7/2 +
33 2.09439 11/2 -
34 2.10100 5/2 -
35 2.12582 3/2 -
36 2.13443 9/2 -
37 2.13700 3/2 -
38 2.14790 3/2 +
39 2.17358 7/2 +
40 2.18300 9/2 -
-------------------
*) 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-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
Nd-141 17.8113 1.0106 -0.4633 0.5388 0.1405 4.2362
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
Pr-140 16.9753 0.0000 -0.5433 0.5678 -0.9137 3.4023
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
Ce-138 16.8661 2.0430 -0.4123 0.5781 1.0263 5.6162
Ce-137 18.4300 1.0252 0.5020 0.5105 0.0280 4.2432
--------------------------------------------------------
Table 3. Gamma-ray strength function for Nd-144
--------------------------------------------------------
* E1: ER = 15.05 (MeV) EG = 5.28 (MeV) SIG = 317.00 (mb)
* M1: ER = 7.82 (MeV) EG = 4.00 (MeV) SIG = 0.76 (mb)
* E2: ER = 12.02 (MeV) EG = 4.38 (MeV) SIG = 3.40 (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. et al.: "Neutron Cross Sections, Vol. I,
Part A", Academic Press (1981).
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) Kikuchi,Y. et al.: JAERI-Data/Code 99-025 (1999)
[in Japanese].
8) Soukhovitski,E.Sh. et al.: JAERI-Data/Code 2005-002 (2004).
9) Iwamoto,O.: J. Nucl. Sci. Technol., 44, 687 (2007).
10) Kunieda,S. et al.: J. Nucl. Sci. Technol. 44, 838 (2007).
11) Koning,A.J. and Delaroche,J.P.: Nucl. Phys. A713, 231 (2003)
[Global potential].
12) Lohr,J.M. and Haeberli,W.: Nucl. Phys. A232, 381 (1974).
13) Becchetti Jr.,F.D. and Greenlees,G.W.: Ann. Rept.
J.H.Williams Lab., Univ. Minnesota (1969).
14) McFadden,L. and Satchler,G.R.: Nucl. Phys. 84, 177 (1966).
15) Kalbach,C.: Phys. Rev. C33, 818 (1986).
16) Koning,A.J., Duijvestijn,M.C.: Nucl. Phys. A744, 15 (2004).
17) Akkermans,J.M., Gruppelaar,H.: Phys. Lett. 157B, 95 (1985).
18) Moldauer,P.A.: Nucl. Phys. A344, 185 (1980).
19) Mengoni,A. and Nakajima,Y.: J. Nucl. Sci. Technol., 31, 151
(1994).
20) Kopecky,J., Uhl,M.: Phys. Rev. C41, 1941 (1990).
21) Kopecky,J., Uhl,M., Chrien,R.E.: Phys. Rev. C47, 312 (1990).