Cross-Section Library Reference
MCNP and Serpent use numeric suffixes (e.g. .80c, .71c) to identify cross-section libraries. SCONE reads continuous-energy ACE libraries and lists nuclides in nuclearData.materials as ZAID plus a temperature/table index (e.g. 92235.06) that must exist in your aceLibrary handle—so the suffix tables below still matter when you build or choose an ACE file. OpenMC typically uses element/isotope names in Python APIs rather than ZAID.suffix strings. This page explains the naming convention and lists common libraries with their temperatures and energy ranges.
Suffix naming convention
The suffix format is .XXy where XX is a library identifier and y is the data type.
| Type | Description |
|---|---|
| c | Continuous-energy neutron interaction data. The primary data type for transport calculations. |
| p | Photoatomic interaction data for photon transport (MODE P). |
| t | Thermal scattering S(α,β) data for bound-atom thermal treatment. |
| d | Dosimetry cross sections for activation and dose calculations. |
| u | Unresolved resonance probability tables. |
Library families
ENDF/B — Evaluated Nuclear Data File (US)
NNDC / BNL / LANL
The primary US nuclear data library maintained by the Cross Section Evaluation Working Group (CSEWG). ENDF/B-VIII.0 (2018) is the latest release with significant improvements to light nuclei, actinides, and thermal scattering data.
https://www.nndc.bnl.gov/endf/JEFF — Joint Evaluated Fission and Fusion File (Europe)
NEA Data Bank / OECD
The European evaluated library jointly developed by NEA member countries. JEFF-3.3 (2017) features strong structural material and fission product evaluations, often complementary to ENDF/B.
https://www.oecd-nea.org/dbdata/jeff/JENDL — Japanese Evaluated Nuclear Data Library
JAEA
Japan's comprehensive evaluated library. JENDL-5.0 (2021) includes extensive updates to minor actinides and fission products, particularly strong for fast-reactor applications.
https://wwwndc.jaea.go.jp/jendl/jendl.htmlSuffix table
| Suffix | Library | Temperature | Type | Energy range | Description |
|---|---|---|---|---|---|
| .80c | ENDF/B-VIII.0 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Latest US evaluated library. Recommended for most applications.(recommended) |
| .81c | ENDF/B-VIII.0 | 600 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VIII.0 at 600 K for hot-zero-power conditions. |
| .82c | ENDF/B-VIII.0 | 900 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VIII.0 at 900 K for fuel-temperature conditions. |
| .83c | ENDF/B-VIII.0 | 1200 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VIII.0 at 1200 K for high-temperature fuel. |
| .84c | ENDF/B-VIII.0 | 2500 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VIII.0 at 2500 K for accident-condition fuel temperatures. |
| .71c | ENDF/B-VII.1 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Widely used US library. Default in many MCNP installations.(recommended) |
| .72c | ENDF/B-VII.1 | 600 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VII.1 at 600 K. |
| .73c | ENDF/B-VII.1 | 900 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VII.1 at 900 K. |
| .74c | ENDF/B-VII.1 | 1200 K | Continuous-energy neutron | 1e-11 – 20 MeV | ENDF/B-VII.1 at 1200 K. |
| .70c | ENDF/B-VII.0 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Legacy US library. Superseded by VII.1 and VIII.0. |
| .66c | ENDF/B-VI.8 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Legacy library. Not recommended for new calculations. |
| .84p | ENDF/B-VIII.0 | N/A | Photoatomic | 1 keV – 100 GeV | Photon interaction cross sections (photoelectric, Compton, pair production).(recommended) |
| .70p | ENDF/B-VII.0 | N/A | Photoatomic | 1 keV – 100 GeV | Legacy photon interaction data. |
| .20t | ENDF/B-VIII.0 | 293.6 K | Thermal S(α,β) | < 4 eV | Thermal scattering law for bound atoms. Required for accurate thermal-spectrum transport.(recommended) |
| .21t | ENDF/B-VIII.0 | 350 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 350 K. |
| .22t | ENDF/B-VIII.0 | 400 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 400 K. |
| .23t | ENDF/B-VIII.0 | 500 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 500 K. |
| .24t | ENDF/B-VIII.0 | 600 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 600 K. |
| .25t | ENDF/B-VIII.0 | 800 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 800 K. |
| .10t | ENDF/B-VII.1 | 293.6 K | Thermal S(α,β) | < 4 eV | Thermal scattering law tables from ENDF/B-VII.1. |
| .11t | ENDF/B-VII.1 | 350 K | Thermal S(α,β) | < 4 eV | S(α,β) data from ENDF/B-VII.1 at 350 K. |
| .12t | ENDF/B-VII.1 | 400 K | Thermal S(α,β) | < 4 eV | S(α,β) data from ENDF/B-VII.1 at 400 K. |
| .31c | JEFF-3.1 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | European Joint Evaluated Fission and Fusion File. |
| .32c | JEFF-3.2 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Updated JEFF library with improved actinide evaluations. |
| .33c | JEFF-3.3 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Latest JEFF release. Strong for structural materials and fission products.(recommended) |
| .40j | JENDL-4.0 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Japanese Evaluated Nuclear Data Library. |
| .50j | JENDL-5.0 | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Latest Japanese library with comprehensive updates to minor actinides.(recommended) |
ZAID identifier rules
A ZAID (Z and A IDentifier) uniquely identifies a nuclide: ZZZAAA where ZZZ is the atomic number and AAA is the mass number.
Format
92235→ Z=92 (U), A=2351001→ Z=1 (H), A=18016→ Z=8 (O), A=1694239→ Z=94 (Pu), A=239
Common mistakes
92000= natural uranium (A=000), not U-200- Full ID in MCNP/Serpent:
92235.80c(ZAID + suffix) - SCONE (ACE):
92235.06incomposition— suffix must match a nuclide in your ACE library (see SCONE nuclear data) - OpenMC uses element names:
'U235', not numeric ZAIDs - Metastable states add 400 to A:
95642= Am-242m (95000 + 242 + 400)
Thermal scattering S(α,β) guide
Below ~4 eV, neutron scattering is affected by chemical binding and crystal structure. Free-atom cross-sections are inaccurate — you must apply S(α,β) thermal scattering libraries for bound scatterers.
| Scatterer | MCNP (mt card) | Serpent (therm) | OpenMC |
|---|---|---|---|
| H in light water | lwtr.20t | lwj3.11t / lwj3.22t | c_H_in_H2O |
| D in heavy water | hwtr.20t | hwj3.11t | c_D_in_D2O |
| C in graphite | grph.20t | grj3.11t | c_Graphite |
| H in polyethylene | poly.20t | polj3.11t | c_H_in_CH2 |
| H in ZrH | h-zr.20t | hzrj3.11t | c_H_in_ZrH |
| Be metal | be.20t | bej3.11t | c_Be |
| Zr in ZrH | zr-h.20t | zrzrj3.11t | c_Zr_in_ZrH |
Match the thermal scattering library temperature to your material temperature. Use .11t (~300 K) for room temperature and .22t (~600 K) for reactor operating conditions. Some libraries support interpolation between two temperatures.
SCONE: continuous-energy transport uses whatever thermal and S(α,β) data are present in your processed ACE library. Align material temp and ZAID suffixes with that library’s documentation—the MCNP/Serpent thermal names above describe the same underlying evaluations you often embed in ACE builds.
Common MT reaction numbers
MT numbers identify specific nuclear reactions in cross-section data and tally specifications.
| MT | Reaction | Description |
|---|---|---|
| 1 | (n,total) | Total cross-section |
| 2 | (n,elastic) | Elastic scattering |
| 4 | (n,inelastic) | Total inelastic scattering |
| 16 | (n,2n) | Neutron multiplication |
| 18 | (n,fission) | Total fission |
| 102 | (n,γ) | Radiative capture |
| 103 | (n,p) | Proton production |
| 104 | (n,d) | Deuteron production |
| 105 | (n,t) | Triton production |
| 107 | (n,α) | Alpha production |
| 251 | μ̄ | Average scattering cosine |
| -2 | absorption | Total absorption (MCNP tally multiplier) |
| -6 | fission ν | Total fission × ν (MCNP tally multiplier) |
Material card sanity checks
Before running
- Don't mix atom and weight fractions in the same material (MCNP/Serpent enforce this)
- Fractions don't need to sum to 1 — codes normalize automatically — but ratios must be correct
- Check density sign: negative = g/cm³, positive = atoms/barn-cm (MCNP cell cards; SCONE uses atoms/barn-cm in
composition) - Verify library suffix exists in your xsdir/xsdata for every ZAID (or, for SCONE, that each ZAID.suffix exists in your ACE file)
- Add S(α,β) for any bound scatterer below ~4 eV (water, graphite, poly, ZrH)
Common errors
- UO₂ with wrong O fraction: should be 2 atoms O per 1 atom U (atom ratio), not by weight
- Borated water: 1000 ppm boron ≈ 0.001 weight fraction, not 0.001 atom fraction
- Forgetting that Zirc-4 has Sn, Fe, Cr — not just Zr
- Using room-temperature density (1.0 g/cm³) for hot water at reactor conditions (~0.7 g/cm³)
- Missing thermal scattering for hydrogen — can shift k-eff by 1000+ pcm