WPEC SG21: Assessment of Neutron Cross-Section Evaluations for the Bulk
of Fission Products

> 49-In-115 Review

Pavel Oblozinsky, BNL, March 2003

### **1. General**

### 

### **Properties:** Stable nucleus (T ~1/2~ = 4.41x10^14^y), natural abundance 95.71%.

### 

**EXFOR data**

Experimental data are generally available. In particular, total, capture
and partly also (n,n') are available; (n,p), (n,) and (n,2n) available
around 14 MeV only.

### **Evaluations**

-   ENDF/B-VI partial evaluation (capture), year of origin 1974, by
    Schenter & Schmittroth

> (ORNL); in 1990 production of ^115m^In isomer via (n,n') added by
> Chiba

-   JEF-2 year of origin 1982, by Gruppelaar (Petten) and Menapace
    (Bologna)

-   BROND-2 none

-   JENDL-3 year of origin 1990, author JNDC FP Working Group

-   CENDL-3 year of origin 2001, author Zhang et al (CNDC Beijing)

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

**ENDF/B-VI**

-   Important comment: capture only

-   Thermal cross sections based on BNL-325 (1973).

-   RR region -- parameters taken from BNL-325 (1973).

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-   UR region - no data.

**JEF-2**

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

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

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-   UR region - no data.

**JENDL-3**

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

-   RR region up to 2 keV, MLBW formula used with parameters from
    BNL-325 (1981) with

> some updates.

-   UR region (2 keV -- 100 keV) based on BNL-325 (1981)

**CENDL-3**

-   Taken completely from JENDL-3.2.

**Recommendation/Conclusion**

In the thermal and resonance region, JENDL-3 represents original and
fairly up-to-date evaluation, CENDL-3 is just a copy of it. JENDL-3 is
well documented. **JENDL-3 is recommended**.

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

#### ENDF/B-VI methodology

-   Partial evaluation only: capture and production of isomer ^115m^In

-   Evaluation of dosimetry reaction ^115^In(n, n') ^115m^In was
    performed using experimental data

> and GMA code by Chiba and D.L. Smith in 1990, including covariance
> matrix.

#### 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.

-   Pre-equilibrium effects included using code PEGASUS.

-   (n, p) and (n,) normalized at 14 MeV using systematic values by
    Forrest.

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

##### JEF-2 methodology

-   Data mostly taken from ENDF/B-V (1974, Schenter & Schmittroth) that
    in general judged

> to be obsolete.

-   Charged-particle production data 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 NUNF.

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

-   Angular distributions are provided, either based OM, NUNF or
    considered isotropic.

**Comparison of evaluations**

-   MT=1 CENDL-3 probably best, in particular in 0.1 --3.7 MeV range
    where adjustments

> to experimental data were made.

-   MT= 4 Unrealistic shape for JEF-2, while JENDL-3 and CENDL-3 look
    ok.

-   MT=16 JEF-2 not available, JENDL-3 and CENDL-3 similar.

-   MT=102 Evaluations look equivalent.

-   MT=103 Both JENDL-3 and CENDL-3 probably better justified by physics
    than JEF-2.

-   MT=107 Evaluations look equivalent.

**Recommendation/Conclusion**

In the fast energy region, JENDL-3 and CENDL-3 evaluation are based on
up-to-date methodology, and they are fairly complete. CENDL-3 looks
somewhat better for MT=1, otherwise both evaluations look equivalent.
ENDF/B-VI is of interest in view of the dosimetry cross sections for
^115m^In, including covariance data. **CENDL-3 or JENDL-3 is
recommended**.

Table 1. Thermal cross sections for In-115

+----------------+-------+--------+-------+--------------+-----------+
| **\            | #     | # C    | # El  | # Re         | # O       |
| Source**       | Total | apture | astic | sonance Capt | ne-group  |
|                |       |        |       | ure Integral |           |
|                | **(ba | **(b   | **(ba |              | # Capture |
|                | rn)** | arn)** | rn)** | **(barn)**   |           |
|                |       |        |       |              | **(barn)  |
|                |       |        |       |              | \[2\]**   |
+----------------+-------+--------+-------+--------------+-----------+
| # ENDF/B-VI    | #     | #      | #     | # 5.488E-1   |           |
|                |       |        |  2587 |              |           |
+----------------+-------+--------+-------+--------------+-----------+
| JEF-2          |       | 202    | 2.5   | 3215         | 4.769E-1  |
+----------------+-------+--------+-------+--------------+-----------+
| JENDL-3        | 203.5 | 201    | 2.5   | 3210         | 6.090E-1  |
+----------------+-------+--------+-------+--------------+-----------+
| CENDL-3        | 203.5 | 201    | 2.5   | 3204         | 6.116E-1  |
+----------------+-------+--------+-------+--------------+-----------+
| Mughabghab     |       | 202(2) |       | 3300(100)    |           |
| 2003 \[1\]     |       |        |       |              |           |
+----------------+-------+--------+-------+--------------+-----------+

**Table 2**. Summary of resolved and unresolved resonance region for
In-115

<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: 12%" />
<col style="width: 0%" />
<col style="width: 11%" />
<col style="width: 37%" />
</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><strong>Remarks</strong></td>
</tr>
<tr class="even">
<td>ENDF/B-VI1974</td>
<td colspan="2">Schenter</td>
<td colspan="2"></td>
<td colspan="2"><h1 id="section-5"></h1></td>
<td colspan="2">Based on BNL-325 (1973)</td>
<td></td>
</tr>
<tr class="odd">
<td>JEF-2</td>
<td colspan="2">1982</td>
<td colspan="2">Gruppelaar</td>
<td colspan="2">1.11 keV</td>
<td colspan="2"><h1 id="section-6"></h1></td>
<td>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">2 keV<br />
MLBW</td>
<td>2 keV – 100 keV</td>
<td><p>MLBW formalism adopted. Based on</p>
<p>BNL-325 (1981) with some update.</p></td>
</tr>
<tr class="odd">
<td colspan="2">CENDL-3</td>
<td colspan="2">2001</td>
<td colspan="2">Zhang et al.</td>
<td colspan="2"></td>
<td></td>
<td>Taken from JENDL-3.</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>.