Nd-146

Reviewed by T. Nakagawa, JAERI, March 2004

**1. Files**

***ENDF/B-VI***: Evaluation was made by Schenter et al. in 1980.

***JENDL-3.3 and JENDL-3.2***:

New evaluation was made by JNDC FP Nuclear Data WG for JENDL-2 in 1984
\[Ao85\]. Re-evaluation was made in 1990 for JENDL-3.1 \[Ka92\].
JENDL-3.3 and JENDL-3.2 are the same as JENDL-3.1.

***JEFF-3.0 and JEF-2.2***:

RCN-3 evaluation made in 1978 was adopted to JEF-2. Charged particle
emission cross sections were adopted from REAC-ECN-4. JEFF-3.0 is the
same as JEF-2.2.

***CENDL-3***: New evaluation was made by Qing-Biao Shen et al. \[Sh01\]

***BROND-2***: No data are given.

**2. Thermal and resonance region**

***ENDF/B-VI***

Resonance parameters recommended by Mughabghab et al. \[Mu73\] were
adopted. Large background cross section was given to the capture.

No unresolved resonance parameters were given.

***JENDL-3.3 and JENDL-3.2***

Resolved resonance parameters were based on the data of Tellier \[Te71\]
and Musgrove et al. \[Mu77\] A negative resonance was added so as to
reproduce the thermal capture cross section of 1.4 b \[Mu81\].

The unresolved resonance parameters were determined from optical model
and statistical model calculation with CASTHY code.

***JEFF-3.0 and JEF-2.2***

Resolved resonance parameters were based on the data of Musgrove et al.
\[Mu77\] and the recommendation of Mughabghab et al. \[Mu73\].

Unresolved resonance parameters were given up to 50 keV on the basis of
FISPRO-ECN and SASSI-ECN code calculation.

Background cross sections were given to the capture and elastic
scattering cross sections.

***CENDL-3***

Resonance parameters were taken from JENDL-3.2.

**[Recommendation/conclusion for the thermal and resonance
region]{.underline}**

***JENDL-3.3*** is recommended for the RRP. ***JEFF-3.0*** is
recommended for URP.

**3. Fast neutron region**

**(1) Methodology**

***ENDF/B-VI***

Theoretical calculations were performed using OMP of Moldauer \[Mo63\],
and codes of COMNUC-3 for inelastic and capture, and NCAP for capture.
Threshold reaction cross sections were calculated with THRESH code.
Angular distributions of neutrons emitted from inelastic scattering and
other threshold reactions were assumed to be isotropic.

***JENDL-3.3 and JENDL-3.2***

Theoretical model calculations of total, elastic and inelastic
scattering, and capture were made by using the statistical model code
CASTHY. The gamma-ray strength function was adjusted to the capture
cross section of 114 mb at 30 keV \[Na78\]. The direct/semi-direct
capture was calculated from a simple formula of Benzi and Reffo
\[Be69\].

Other reaction cross sections were calculated with the preequilibrium
and multi-step evaporation code PEGASUS. The (n,p) and (n,α) reaction
cross sections were normalized to the recommendation of Forrest \[Fo86\]
at 14.5 MeV. Sum of the threshold reaction cross sections were
considered as a competing reaction cross sections in the CASTHY
calculation.

***JEFF-3.0 and JEF-2.2***

The data of RCN-3 evaluation was adopted. All the cross sections above
the resonance region were calculated with SASSI-ECN code.

***CENDL-3***

Theoretical calculation was performed with SUNF and DWUCK4 codes.

**(2) Total cross section (MT=1)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt1)Nd146.htm#mt1,
Fig. A1

Figure A1 shows comparison of evaluated data with the experimental data
for Nd-146 and natural Nd. The energy dependence of ***JENDL-3.3*** is
not correct. The shape of ***CENDL-3*** is not smooth around 100 keV.
***JEFF-3.0*** seems to reproduce well the experimental data.

**(3) Capture cross section (MT=102)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt102)Nd146.htm#mt102,
Fig. A2

***CENDL-3*** has a structure around 100 keV, and it is too large above
100 keV. ***ENDF/B-VI*** is also too large around 3 MeV. Veerapaspong et
al. \[Ve99\] measured recently the Nd-146 capture cross section. Their
results are close to the data of Wisshak et al. \[Wi97\] and
***JEFF-3.0***.

**(4) Elastic scattering cross section (MT=2)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt2)Nd146.htm#mt2

**(5) Inelastic scattering cross section (MT=4, 51 -- 91)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt4)Nd146.htm#mt4

Discrete levels considered in the evaluations are as follows:

ENDF/B-VI 3 levels up to 1.044 MeV

JENDL-3.3(=JENDL-3.2) 9 levels up to 1.751 MeV

JEFF-3.0 (=JEF-2.2) 13 levels up to 1.812 MeV

CENDL-3 11 levels up to 1.7451 MeV

The direct inelastic was calculated with DWUCK4 code for ***CENDL-3***.

**(6) (n,2n) reaction cross section (MT=16)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt16)Nd146.htm#mt16,
Fig. A3

***CENDL-3*** and ***JENDL-3.3*** reproduce well experimental data.

**(7) (n,p) reaction cross section (MT=103)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt103)Nd146.htm#mt103,
Fig. A4

***JENDL-3.3***, ***JEFF-3.0*** and ***CENDL-3*** are in good agreement
with experimental data. However, ***JEFF-3.0*** is too large above 15
MeV.

**(8) (n,α) reaction cross section (MT=107)**

See Fig:
[http://www.nndc.bnl.gov/sg21/fp21/htm/](http://www.nndc.bnl.gov/sg21/fp21/htm/Nd146.htm#mt107)Nd146.htm#mt107,
Fig. A5

***JENDL-3.3*** and ***CENDL-3*** are in agreement with experimental
data. ***JEFF-3.0*** is too large above 15 MeV.

**(9) Other reactions**

The following reaction cross sections are also given:

ENDF/B-VI (n,3n), (n,nα), (n,np), (n,d), (n,t), (n,^3^He)

JENDL-3.3 (JENDL-3.2) (n,3n), (n,nα), (n,np), (n,nd), (n,nt), (n,d),
(n,t)

JEFF-3.0 (JEF-2.2) (n,d), (n,t), (n,^3^He), (n,2p)

CENDL-3 (n,3n), (n,nα), (n,np), (n,d), (n,t)

**[Recommendation/conclusion for the fast energy region]{.underline}**

For the total and capture cross sections, ***JEFF-3.0*** is recommended.
For the inelastic scattering and other threshold reactions,
***CENDL-3*** is recommended. The elastic scattering cross section
should be calculated as total -- sum of partial cross sections.

**References**

\[Al67\] Alstad J. et al.: *J. Inorg. Nucl. Chem.*, **29**, 2155 (1967),
EXFOR20044.

\[Ao85\] Aoki T. et al.: 1985 Santa Fe, Vol.2, p.1627 (1985).

\[Be69\] Benzi V. and Reffo G.: *CCDN-NW/*10 (1969).

\[Fo86\] Forrest R.A.: *AERE-R* 12419 (1986).

\[Gr76\] Gryntakis E.M.: private communications (1976) EXFOR 20625. See
also *Radiochim. Acta*, **17**, 191 (1972).

\[He78\] Heft R.E.: 1978 Mayag, p.495 (1978), EXFOR 12866.

\[Ka92\] Kawai M. et al.: *J. Nucl. Sci. Technol.*, **29**, 195 (1992).

\[Ko53\] Koehler W. et al.: *Phys. Rev.*, **91**, 597 (1953),
EXFOR12060.

\[Mo63\] Moldauer P.: *Nucl. Phys.*, **47**, 65 (1963).

\[Mo68\] Mowatt R.S. et al.: *EANDC(CAN)*-38, p.4 (1968), EXFOR13492.

\[Mu73\] Mughabghab S.F. and Garber D.I.: "Neutron Cross Sections Vol.
1, Resonance Parameters," *BNL*-325, 3rd Edition Vol.1 (1973).

\[Mu77\] Musgrove A.R. Del et al.: *AEEC/E*401 (1977).

\[Mu81\] Mughabghab S.F. et al.: *Neutron Cross Sections*, Vol.1, Part
A, Academic Press (1981).

\[Mu03\] Mughabghab S.F.: "Thermal Neutron Capture Cross Sections
Resonance Integrals and g-Factors," *INDC(NDS)*-440 (2003)

\[Na78\] Nakajima Y. et al.: 1978 Harwell, p.438 (1978).

\[Ri73\] Ricabarra M.D. et al.: *Can. J. Phys.*, **51**, 1454 (1973),
EXFOR30239.

\[Sh01\] Shen Q.: *Comm. Nucl. Data Prog.*, **25**, 29 (2001).

\[St72\] Steinnes E.: *J. Inorg. Nucl. Chem.*, **34**, 2699 (1972),
EXFOR20188.

\[Ta78\] Takiue M. et al.: *Nucl. Instrum. Methods. Phys. Res.*,
**148**, 157 (1978), EXFOR29853.

\[Te71\] Tellier H.: *CEA-N*-1459 (1971).

\[Va74\] Van der Linden R. et al.: *J. Radioanal. Chem.*, **20**, 695
(1974), EXFOR20645.

\[Ve73\] Vertebnyj V.P. et al.: 1973 Kiev, vol.2, p.104 (1973),
EXFOR40290.

\[Ve99\] Veerapaspong T. et al.: J. Nucl. Sci. Technol., 36, 855 (1999).

\[Wi97\] Wisshak K. et al.: *KfK* 5967 (1997), EXFOR22388.

Table 1 Summary for Resolved and Unresolved Resonance Region: Nd-146

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

Library RRP URP Remarks

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

> ENDF/B-VI 4.1774 keV, SLBW no urp
>
> JENDL-3.3 10 keV, MLBW 10 -- 100 keV same as JENDL-3.2
>
> JEFF-3.0 10 keV, MLBW 10 -- 50 keV same as JEF-2.2
>
> CENDL-3 10 keV, MLBW 10 -- 100 keV same as JENDL-3.2

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

Table 2. Thermal Cross Sections and Resonance Integral: Nd-146

Units: barns

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

Library/Experiment Total Elastic Capture Res.Integ.

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

ENDF/B-VI 3.098 1.698 1.400 3.10

JENDL-3.3 10.895 9.496 1.400 2.91

JEFF-3.0 5.766 4.366 1.400 2.82

CENDL-3 10.895 9.496 1.399 2.97

Mughabghab et al. \[Mu81\] 9.6±0.4 1.4±0.1 3.2±0.5

Mughabghab et al. \[Mu03\] 1.41±0.05 23.2±0.5

Koehler et al. \[Ko53\] 9.5±0.4^\*4^

Alstad et al. \[Al67\] 1.3±0.1^\*1^ 3.0±0.3 (0.4)^\*3^

Mowatt et al. \[Mo68\] 1.8±0.3^\*2^

Steinnes \[St72\] 3.2±0.5 (0.5)

Ricabarra et al. \[Ri73\] 2.58±0.2 (0.6)

Vertebnyj et al. \[Ve73\] 9.5±0.3

Van der Linden et al. \[Va74\] 2.84±0.21 (0.55)

Gryntakis \[Gr76\] 1.31±0.06^\*1^ 2.57±0.14 (0.55)

Takiue et al. \[Ta78\] 1.36±0.05^\*1^

Heft \[He78\] 1.55±0.03^\*2^ 3.65±0.08 (0.5)

\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

\*1 Measured in Maxwellian spectrum.

\*2 Measured in reactor spectrum.

\*3 Cut-off energy (eV)

\*4 Measured at 0.07 eV.

![](media/image1.wmf){width="8.197222222222223in"
height="7.382638888888889in"}

Fig. A1 Nd-146 and Nd-natural total cross section

![](media/image2.wmf){width="8.197222222222223in"
height="7.382638888888889in"}

Fig. A2 Nd-146 capture cross section

![](media/image3.wmf){width="8.370138888888889in"
height="7.382638888888889in"}

Fig. A3 Nd-146(n,2n) cross section

![](media/image4.wmf){width="8.568055555555556in"
height="7.382638888888889in"}

Fig. A4 Nd-146 (n,p) cross section

![](media/image5.wmf){width="8.568055555555556in"
height="7.382638888888889in"}

Fig. A5 Nd-146 (n,α) cross section