93-Np-237
93-Np-237 JAEA+ EVAL-JAN10 O.Iwamoto,T.Nakagawa,S.Chiba,+
DIST-MAY10 20100524
----JENDL-4.0 MATERIAL 9346
-----INCIDENT NEUTRON DATA
------ENDF-6 FORMAT
History
06-02 Fission cross section was evaluated with GMA code.
07-05 New calculation was made with CCONE code.
07-07 Re-calculation was made with CCONE code.
07-08 Fission cross section was revised.
07-11 Isomeric ratio of the (n,2n) reaction was given.
07-12 Resonance parameters
08-01 Fission cross section was revised.
08-02 Fission cross section and nu-p were revised.
CCONE calculation was made with revised parameters.
08-03 Data were compiled as JENDL/AC-2008/1/.
09-03 Resonance parameters and fission cross section were
modified.
09-08 (MF1,MT458) was evaluated.
10-01 Data of prompt gamma rays due to fission were given.
10-02 Covariance data were given.
MF= 1
MT=452 Total neutron per fission
Sum of MT=455 and 456.
MT=455 Delayed neutrons
(same as JENDL-3.3)
Nu-d was based on the experimental data of Saleh et al./2/,
Charlton et al./3/, Piksaikin et al./4/ and Zeinalov et
al./5/ Decay constants were detemined from the experimental
data of Piksaikin et al./4/.
MT=456 Prompt neutrons per fission
The following data were fitted by GMA code /6/:
Nu-p measured by Veeser/7/, Frehaut et al./8/, Malinovskii
et al./9/, and nu-total by Boikov et al./10/, Thierens et
al./11/, Mueller et al./12/ and Khokhlov et al./13/
MT=458 Components of energy release due to fission
Total energy and prompt energy were calculated from mass
balance using JENDL-4 fission yields data and mass excess
evaluation. Mass excess values were from Audi's 2009
evaluation/14/. Delayed energy values were calculated from
the energy release for infinite irradiation using JENDL FP
Decay Data File 2000 and JENDL-4 yields data. For delayed
neutron energy, as the JENDL FP Decay Data File 2000/15/ does
not include average neutron energy values, the average values
were calculated using the formula shown in the report by
T.R. England/16/. The fractions of prompt energy were
calculated using the fractions of Sher's evaluation/17/ when
they were provided. When the fractions were not given by Sher,
averaged fractions were used.
MF= 2 Resonance parameters
MT=151
Resolved resonance parameters (below 500eV)
The parameters given in JENDL-3.3 were modified:
Parameters of the resonances below 3.9 eV were modified so
as to reproduce the experimental data of Esch et al./18/ and
Tovesson and Hill/19/, and the thermal capture cross section
of 178.3 b and fission of 0.0202 b.
Background cross section was given to the capture to get
agreement with the average cross section of Esch et al./18/
Thermal cross sections at 0.0253 eV were based on:
fission:
Wagemans et al./20,21/, Kozharin et al./22/
capture:
Kobayashi et al./23/, Katoh et al./24/, Harada et
al./25/, Bringer et al./26/, Esch et al./18/, and
others.
Unresolved resonance parameters (500eV - 30keV)
Parameters were determined with ASREP code/27/ so as to
reproduce the cross sections in the energy range from 500 eV
to 30 keV. They are used only for self-shielding calculations.
Thermal cross sections and resonance integrals (at 300K)
-------------------------------------------------------
0.0253 eV reson. integ. (*)
(barns) (barns)
-------------------------------------------------------
total 192.57
elastic 14.47
fission 0.0202 5.38
capture 178.08 696
-------------------------------------------------------
(*) In the energy range from 0.5 eV to 10 MeV.
MF= 3 Neutron cross sections
Cross sections above the resolved resonance region except for
the elastic scattering (MT=2) and fission cross sections (MT=18,
19, 20, 21, 38) were calculated with CCONE code/28/.
The model parameters were determined by considering integral
experimental data as well as measured cross-section data.
MT= 1 Total cross section
The cross section was calculated with CC OMP of Soukhovitskii
et al./29/
MT= 2 Elastic scattering cross section
Calculated as total - non-elastic scattering cross sections.
MT=16 (n,2n) cross section
Calculated with CCONE code. The experimental data of Gromova
et al./30/, Nishi et al./31/ and Landrum et al./32/ were
used to determine the model parameters for the CCONE
calculation.
MT=18 Fission cross section
The following experimental data reported after 1982 were
analyzed with the GMA code /6/:
Behrens+/33/, Cance+/34/, Alkhazov+/35/, Meadows/36/,
Wu+/37/, Garlea+/38/, Goverdovskij+/39/, Goverdovskij+/40/,
Zasadny+/41/, Goverdovskij+/42/, Kanda+/43/, Kovalenko+/44/,
Alkhazov+/45/, Gul+/46/, Terayama+/47/, Meadows /48/,
Desdin+/49/, Merla+/50/, Garlea+/51/, Shcherbakov+/52/,
Furman+/53/, Baba+/54/, Tovesson+/55/.
The data measured relatively to U235 fission were transformed
to cross sections by using the U235 fission cross section of
JENDL/AC-2008.
The results of GMA were used to determine the parameters in
the CCONE calculation.
Further modification was made for JENDL-4.0 in the energy
range from 500eV to 200keV, by eye-guiding the experimental
data of Tovesson and Hill /19/.
MT=19, 20, 21, 38 Multi-chance fission cross sections
Calculated with CCONE code, and renormalized to the total
fission cross section (MT=18).
MT=102 Capture cross section
Calculted with CCONE code. The experimental data of Esch et
al./56/, Kobayashi et al./57/, Weston and Todd/58/, Linder
et al./59/, and Buleeva et al./60/ were used to determine
the parameters in the CCONE calculation.
MF= 4 Angular distributions of secondary neutrons
MT=2 Elastic scattering
Calculated with CCONE code.
MT=18 Fission
Isotropic distributions in the laboratory system were assumed.
MF= 5 Energy distributions of secondary neutrons
MT=18 Prompt neutrons
Calculated with CCONE code.
MT=455 Delayed neutrons
Calculated by Brady and England/61/.
MF= 6 Energy-angle distributions
Calculated with CCONE code.
Distributions from fission (MT=18) are not included.
MF= 8 Radioactive Decay
MT=16
Decay data of Np-236: taken from ENDF (as of 2007) /62/
MF= 9 Multiplicities for Production of Radioactive Nuclides
MT=16
Meta-stable state (T-1/2 =22.5H) production was assumed to
be 70% from the cross sections measured around 14 MeV.
MF=12 Photon production multiplicities
MT=18 Fission
Calculated from the total energy released by the prompt
gamma-rays due to fission given in MF=1/MT=458 and the
average energy of gamma-rays.
MF=14 Photon angular distributions
MT=18 Fission
Isotoropic distributions were assumed.
MF=15 Continuous photon energy spectra
MT=18 Fission
Experimental data measured by Verbinski et al./63/ for
Pu-239 thermal fission were adopted.
MF=31 Covariances of average number of neutrons per fission
MT=452 Number of neutrons per fission
Combination of covariances for MT=455 and MT=456.
MT=455
Error of 4% and 10 % was assumed below 2 MeV and above 6 MeV,
respectively. Between 2 and 6 MeV, 7% was assumed./64/
MT=456
Covariance was obtained by GMA fitting to the experimental
data (see MF1,MT456). Obtained standard deviation was
multiplied by a factor of 3, because it was too small.
MF=32 Covariances of resonance parameters
Format of LCOMP=0 was adopted.
Standard diviations of resonance parameters were taken from
JENDL-3.3 covariance file /64/, which were estimated from
errors reported in Refs./65,66,67,68/.
MF=33 Covariances of neutron cross sections
Covariances were given to all the cross sections by using
KALMAN code/69/ and the covariances of model parameters
used in the theoretical calculations.
For the following cross sections, covariances were determined
by different methods.
MT=1, 2 Total and elastic scattering cross sections
In the resonance region (below 500 eV), standard deviation
(SD) of 8 % was added to the contributions from resonance
parameters.
Above 500 eV, covariances were obtaibed with CCONE and
KALMAN codes, and experimental data.
MT=18 Fission cross section
SD of 4% was added in the energy region up to 1 eV, and 15%
from 1 eV to 500 eV.
Above 500 eV, covariances were obtained with GMA code/6/.
SD was multiplied by a factor of 2.0.
MT=102 Capture cross section
Additional SD of 4% was given from 0.1 eV 500 eV.
MF=34 Covariances for Angular Distributions
MT=2 Elastic scattering
Covariances were given only to P1 components.
MF=35 Covariances for Energy Distributions
MT=18 Fission spectra
Estimated with CCONE and KALMAN codes.
*****************************************************************
Calculation with CCONE code
*****************************************************************
Models and parameters used in the CCONE/28/ calculation
1) Coupled channel optical model
Levels in the rotational band were included. Optical model
potential and coupled levels are shown in Table 1.
2) Two-component exciton model/70/
* Global parametrization of Koning-Duijvestijn/71/
was used.
* Gamma emission channel/72/ was added to simulate direct
and semi-direct capture reaction.
3) Hauser-Feshbach statistical model
* Moldauer width fluctuation correction/73/ was included.
* Neutron, gamma and fission decay channel were included.
* Transmission coefficients of neutrons were taken from
coupled channel calculation in Table 1.
* The level scheme of the target is shown in Table 2.
* Level density formula of constant temperature and Fermi-gas
model were used with shell energy correction and collective
enhancement factor. Parameters are shown in Table 3.
* Fission channel:
Double humped fission barriers were assumed.
Fission barrier penetrabilities were calculated with
Hill-Wheler formula/74/. Fission barrier parameters were
shown in Table 4. Transition state model was used and
continuum levels are assumed above the saddles. The level
density parameters for inner and outer saddles are shown in
Tables 5 and 6, respectively.
* Gamma-ray strength function of Kopecky et al/75/,/76/
was used. The prameters are shown in Table 7.
------------------------------------------------------------------
Tables
------------------------------------------------------------------
Table 1. Coupled channel calculation
--------------------------------------------------
* rigid rotor model was applied
* coupled levels = 0,1,3,5,7 (see Table 2)
* optical potential parameters /29/
Volume:
V_0 = 49.8581 MeV
lambda_HF = 0.01004 1/MeV
C_viso = 15.9 MeV
A_v = 12.04 MeV
B_v = 81.36 MeV
E_a = 385 MeV
r_v = 1.25 fm
a_v = 0.57 fm
Surface:
W_0 = 17.1839 MeV
B_s = 11.19 MeV
C_s = 0.01361 1/MeV
C_wiso = 23.5 MeV
r_s = 1.1803 fm
a_s = 0.601 fm
Spin-orbit:
V_so = 5.75 MeV
lambda_so = 0.005 1/MeV
W_so = -3.1 MeV
B_so = 160 MeV
r_so = 1.1214 fm
a_so = 0.59 fm
Coulomb:
C_coul = 1.3
r_c = 1.2452 fm
a_c = 0.545 fm
Deformation:
beta_2 = 0.212764
beta_4 = 0.066
beta_6 = 0.0015
* Calculated strength function
S0= 0.99e-4 S1= 2.49e-4 R'= 9.76 fm (En=1 keV)
--------------------------------------------------
Table 2. Level Scheme of Np-237
-------------------
No. Ex(MeV) J PI
-------------------
0 0.00000 5/2 + *
1 0.03320 7/2 + *
2 0.05954 5/2 -
3 0.07592 9/2 + *
4 0.10296 7/2 -
5 0.13000 11/2 + *
6 0.15851 9/2 -
7 0.19146 13/2 + *
8 0.22596 11/2 -
9 0.26754 3/2 -
10 0.26990 15/2 +
11 0.28135 1/2 -
12 0.30506 13/2 -
13 0.31680 7/2 +
14 0.32442 7/2 -
15 0.33236 1/2 +
16 0.34850 17/2 +
17 0.35970 5/2 -
18 0.36859 5/2 +
19 0.37093 3/2 +
20 0.39552 15/2 -
-------------------
*) Coupled levels in CC calculation
Table 3. Level density parameters
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Np-238 18.3635 0.0000 2.2742 0.3206 -0.9887 1.4167
Np-237 18.2971 0.7795 2.4371 0.3964 -0.9743 3.1574
Np-236 18.2307 0.0000 2.1332 0.3000 -0.7998 1.1669
Np-235 18.1643 0.7828 2.2924 0.3974 -0.9420 3.1307
Np-234 18.0979 0.0000 2.1332 0.2845 -0.6773 1.0000
--------------------------------------------------------
Table 4. Fission barrier parameters
----------------------------------------
Nuclide V_A hw_A V_B hw_B
MeV MeV MeV MeV
----------------------------------------
Np-238 6.199 0.460 5.848 0.370
Np-237 6.250 0.950 5.200 0.600
Np-236 5.500 0.600 5.200 0.400
Np-235 6.250 0.950 5.200 0.600
Np-234 6.200 0.460 5.850 0.370
----------------------------------------
Table 5. Level density above inner saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Np-238 20.5728 0.0000 2.6000 0.3280 -2.4114 2.0000
Np-237 20.4984 0.9094 2.6000 0.3286 -1.5020 2.9094
Np-236 17.3240 0.0000 2.6000 0.3613 -2.5262 2.0000
Np-235 20.3497 0.9133 2.6000 0.3299 -1.4981 2.9133
Np-234 20.2753 0.0000 2.6000 0.3306 -2.4114 2.0000
--------------------------------------------------------
Table 6. Level density above outer saddle
--------------------------------------------------------
Nuclide a* Pair Eshell T E0 Ematch
1/MeV MeV MeV MeV MeV MeV
--------------------------------------------------------
Np-238 20.5728 0.0000 0.2800 0.3661 -1.7021 2.0000
Np-237 20.4984 0.9094 0.2400 0.3674 -0.7921 2.9094
Np-236 17.3240 0.0000 0.2000 0.4067 -1.7431 2.0000
Np-235 20.3497 0.9133 0.1600 0.3699 -0.7870 2.9133
Np-234 20.2753 0.0000 0.1200 0.3712 -1.6997 2.0000
--------------------------------------------------------
Table 7. Gamma-ray strength function for Np-238
--------------------------------------------------------
K0 = 1.300 E0 = 4.500 (MeV)
* E1: ER = 10.98 (MeV) EG = 2.17 (MeV) SIG = 309.44 (mb)
ER = 14.08 (MeV) EG = 4.66 (MeV) SIG = 540.00 (mb)
* M1: ER = 6.62 (MeV) EG = 4.00 (MeV) SIG = 2.08 (mb)
* E2: ER = 10.17 (MeV) EG = 3.25 (MeV) SIG = 6.65 (mb)
--------------------------------------------------------
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