[WPEC SG21: Assessment of Neutron Cross-Section Evaluations for the Bulk
of Fission Products]{.underline}

> 59-Pr-141 Review

Pavel Oblozinsky, BNL, August 2003

### **1. General**

### 

### **Properties:** Stable nucleus, natural abundance 100.00%.

### 

**EXFOR data**

Experimental data are generally available; in particular, total, capture
and (n, 2n) were measured; (n,n') data available around 5 MeV; (n,p) and
(n,) available around 14 MeV.

### **Evaluations**

ENDF/B-VI Thermal and resonance region was re-evaluated in 1999 by Oh
(KAERI) and Mughabghab (BNL), included into ENDF/B-VI release 7. Fast
region datas back to 1980, by Schenter and Schmittroth (ORNL). In 2002,
fast region re-evaluated by Lee (KAERI) and Oblozinsky (BNL), submitted
to ENDF/B-VII

JEF-2 Year of origin 1982, by Gruppelaar (Petten) and Menapace
(Bologna), with some later improvements.

BROND-2 None.

JENDL-3 Year of origin 1990, by JNDC FP Working Group.

CENDL-3 Year of origin 2000, by Songbai Zhang and Qingbiao Shen (CNDC
Beijing).

### **2. Thermal and resonance region**

For summary, see Table 1 and 2.

**ENDF/B-VI**

-   Thermal cross sections re-evaluated in 1999, some changes compared
    to BNL-325 (1981); results close to JEF-2 and JENDL-3.

-   RR region -- re-evaluated in 1999 using MLBW formalism, up to 10.049
    keV.

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-   UR region -- revised in 1999, up to 146.47 keV.

**JEF-2**

-   Thermal cross sections based on BNL-325 (1981).

-   RR region -- parameters taken from BNL-325 (1981).

-   UR region -- no data.

**JENDL-3**

-   Thermal capture cross sections based on BNL-325 (1981).

-   RR region up to 13.226keV, MLBW formalism used with parameters from
    BNL-325 (1981) with some updates.

-   UR region (13.226 keV -- 100 keV) based on BNL-325 (1981).

**CENDL-3**

-   Taken from unspecified release of ENDF/B-VI (before release 7).

**Recommendation/Conclusion**

In the thermal and resonance region the most recent evaluation is
ENDF/B-VI (release 7), it was performed in 1999 under guidance of Said
Mughabghab. **ENDF/B-VI is recommended**.

**[3. Fast neutron region]{.underline}**

#### ENDF/B-VI methodology

-   Failry up-to-date methodology already in 1980.

-   Extra comment: In 2002, completely new evaluation was prepared for
    ENDF/B-VII. It uses up-to-date methodology, based on model
    calculations with the code EMPIRE (statistical model,
    pre-equilibrium model, FKK model) and RIPL input parameter library,
    adjusted to reproduce experimental data.

#### JENDL-3 methodology

-   Methodology fairly up-to-date, all important nuclear reaction
    processes taken into account.

-   Above 100 keV spherical optical model and statistical model code
    CASTHY used. Total cross section above 5 MeV taken from experimental
    data.

-   Pre-equilibrium effects included using code PEGASUS.

-   (n, p) and (n,) normalized at 14.5 MeV to systematic values by
    Forrest. (n,) below 13.2 keV calculated from resonance parameters.

-   Angular distributions are provided, either based on CASTHY or
    considered isotropic.

##### JEF-2 methodology

-   Adopted RCN evaluation, based on statistical model and fits.

-   Charged particle production taken from REAC-ECN-4 library.

##### CENDL-3 methodology

-   Methodology fairly up-to-date, all important nuclear reaction
    processes taken into account.

-   Above 100 keV, spherical optical model used, with own OM parameters.

-   Reaction channels calculated using statistical + pre-equilibrium
    model code SUNF.

-   Direct contributions to (n,n') calculated by DWUCK4.

-   Angular distributions are provided, either based on optical model
    calculations, SUNF or considered isotropic.

**Comparison of evaluations**

-   MT=1 JEF-2 probably best; below 0.2 MeV data in ENDF/B-VI and
    JENDL-3 are

> somewhat high, while CENDL-3 is clearly too high.

-   MT= 4 All evaluations are approximately equivalent and reasonable.

-   MT=16 All files reasonable, JENDL-3 probably the best.

-   MT=102 JENDL-3 and also ENDF/B-VI are the best, JEF-2 and CENDL-3
    tend to

> underestimate data at higher incident energies.

-   MT=103 JENDL-3 is the best, followed by CENDL-3.

-   MT=107 JENDL-3 and CENDL-3 are the best, JEF-2 has unrealistic
    shape.

###### Recommendation/Conclusion

-   In the fast energy region, JENDL-3 is based on up-to-date
    methodology, it is fairly complete evaluation that compares well
    with experimental data in all channels. **JENDL-3 is recommended**.

-   There is new (2002) evaluation submitted for inclusion into
    ENDF/B-VII. In case this new evaluation will be accepted by
    ENDF/B-VII, then **ENDF/B-VII rather than JENDL-3 is recommended**.

Table 1. Thermal cross sections, resonance integral and one-group
capture values for Pr-141.

+----------------+-------+--------+-------+--------------+-----------+
| **\            | #     | # C    | # El  | # Re         | #         |
| Source**       | Total | apture | astic | sonance Capt | One-group |
|                |       |        |       | ure Integral |           |
|                | **(ba | **(b   | **(ba |              | # Capture |
|                | rn)** | arn)** | rn)** | **(barn)**   |           |
|                |       |        |       |              | **(barn)  |
|                |       |        |       |              | \[2\]**   |
+----------------+-------+--------+-------+--------------+-----------+
| #              | #     | # 2.70 | #     | # 1.50E-1    |           |
| ENDF/B-VI14.20 | 11.50 |        |  17.6 |              |           |
+----------------+-------+--------+-------+--------------+-----------+
| JEF-2          | 14.07 | 11.48  | 2.59  | 17.9         | 1.54E-1   |
+----------------+-------+--------+-------+--------------+-----------+
| JENDL-3        | 14.04 | 11.50  | 2.54  | 18.4         | 1.57E-1   |
+----------------+-------+--------+-------+--------------+-----------+
| CENDL-3        | 13.51 | 11.50  | 2.01  | 17.6         | 1.56E-1   |
+----------------+-------+--------+-------+--------------+-----------+
| Mughabghab     |       | 11.    |       | 17.4(2.0)    |           |
| 2003 \[1\]     |       | 5(0.3) |       |              |           |
+----------------+-------+--------+-------+--------------+-----------+

**Table 2**. Summary of resolved and unresolved resonance region for
Pr-141.

<table style="width:100%;">
<colgroup>
<col style="width: 14%" />
<col style="width: 0%" />
<col style="width: 8%" />
<col style="width: 0%" />
<col style="width: 14%" />
<col style="width: 0%" />
<col style="width: 14%" />
<col style="width: 0%" />
<col style="width: 13%" />
<col style="width: 0%" />
<col style="width: 33%" />
</colgroup>
<tbody>
<tr class="odd">
<td><strong>Library</strong></td>
<td colspan="2"><strong>Year</strong></td>
<td colspan="2"><strong>First Author<br />
</strong></td>
<td colspan="2"><strong>RRR<br />
</strong></td>
<td colspan="2"><h1 id="urr">URR</h1></td>
<td colspan="2"><strong>Remarks</strong></td>
</tr>
<tr class="even">
<td>ENDF/B-VI, release 71999</td>
<td colspan="2">R.Q. Wright</td>
<td colspan="2">10.049 keV MLBW</td>
<td colspan="2"><h1 id="kev-146.67-kev">10.049 keV-146.67 keV</h1></td>
<td colspan="2">Re-evaluated under guidance of S. Mughabghab. MLBW
formalism adopted.</td>
<td colspan="2"></td>
</tr>
<tr class="odd">
<td>JEF-2</td>
<td colspan="2">1982</td>
<td colspan="2">Gruppelaar</td>
<td colspan="2">5.77274 keV</td>
<td colspan="2"><h1 id="section-5"></h1></td>
<td colspan="2">Based on BNL-325 (1981).</td>
</tr>
<tr class="even">
<td colspan="2">JENDL-3</td>
<td colspan="2">1990</td>
<td colspan="2"><p>JNDC FP</p>
<p>Work. Group</p></td>
<td colspan="2">13.226 keV<br />
MLBW</td>
<td colspan="2">13.226 keV – 100 keV</td>
<td><p>Taken from JENDL-2 with some</p>
<p>update.</p></td>
</tr>
<tr class="odd">
<td colspan="2">CENDL-3</td>
<td colspan="2">2000</td>
<td colspan="2">S. Zhang</td>
<td colspan="2"></td>
<td colspan="2"></td>
<td>Taken from earlier version of ENDF/B-VI (before release 7).</td>
</tr>
</tbody>
</table>

#### References

1.  S.F. Mughabghab, Thermal neutron capture cross sections, resonance
    integrals and g-factors, Report INDC (NDS)-440 (IAEA, Vienna,
    February 2003).

2.  H. Gruppelaar et al, Status of pseudo-fission-product cross sections
    for fast reactors, Final Report of Subgroup 17, Report NEA/WPEC-17,
    ECN-R-98-014 (NEA, Paris, 1998). Recalculated systematically for all
    211 FP and all 5 evaluated libraries by I. Sirakov (BNL, 2002), see
    <http://www.nndc.bnl.gov/sg21/nucl/acs/accs.htm>.