Cross-Section Library Reference
Suffix naming convention
The suffix format is .XXy where XX is a library identifier and y is the data type.
The library identifier is chosen by whoever processed the ACE files, not by the evaluation. The table below gives the LANL numbering that ships with MCNP6, where .80c is ENDF/B-VII.1 and ENDF/B-VIII.0 lives at .00c. Other vendors number differently: KAERI uses .80c for ENDF/B-VI.8, and JAEA ships JENDL-5 at .20c. Your own xsdir file is the only authority for the machine you are running on.
| Type | Description |
|---|---|
| c | Continuous-energy neutron interaction data. The primary data type for transport calculations. |
| t | Thermal scattering S(α,β) data for bound-atom thermal treatment. |
| d | Discrete-energy neutron data. Rarely used in modern work. |
| p | Photoatomic interaction data for photon transport (MODE P). |
| u | Photonuclear data — photons inducing nuclear reactions. |
| y | Dosimetry cross sections. Usable as response functions, never as transport data. |
| e | Electron data. |
| h | Proton data. |
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) brought significant improvements to light nuclei, actinides, and thermal scattering data; ENDF/B-VIII.1 followed in August 2024. LANL processes these into the ACE files MCNP reads and publishes them as Lib80x, ENDF71x, and so on.
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 |
|---|---|---|---|---|---|
| .00c | ENDF/B-VIII.0 (Lib80x) | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Latest US evaluated library at room temperature. Recommended for most new work.(recommended) |
| .01c | ENDF/B-VIII.0 (Lib80x) | 600 K | Continuous-energy neutron | 1e-11 – 20 MeV | Hot-zero-power and coolant temperatures. |
| .02c | ENDF/B-VIII.0 (Lib80x) | 900 K | Continuous-energy neutron | 1e-11 – 20 MeV | Intermediate fuel temperatures. |
| .03c | ENDF/B-VIII.0 (Lib80x) | 1200 K | Continuous-energy neutron | 1e-11 – 20 MeV | Typical operating fuel temperature for a PWR pellet. |
| .04c | ENDF/B-VIII.0 (Lib80x) | 2500 K | Continuous-energy neutron | 1e-11 – 20 MeV | Accident-condition fuel temperatures. |
| .05c | ENDF/B-VIII.0 (Lib80x) | 0.1 K | Continuous-energy neutron | 1e-11 – 20 MeV | Effectively unbroadened data, used by the DBRC treatment. |
| .06c | ENDF/B-VIII.0 (Lib80x) | 250 K | Continuous-energy neutron | 1e-11 – 20 MeV | Below room temperature. |
| .80c | ENDF/B-VII.1 (ENDF71x) | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | The suffix most often seen in published MCNP decks. This is VII.1, not VIII.0.(recommended) |
| .81c | ENDF/B-VII.1 (ENDF71x) | 600 K | Continuous-energy neutron | 1e-11 – 20 MeV | Hot-zero-power and coolant temperatures. |
| .82c | ENDF/B-VII.1 (ENDF71x) | 900 K | Continuous-energy neutron | 1e-11 – 20 MeV | Intermediate fuel temperatures. |
| .83c | ENDF/B-VII.1 (ENDF71x) | 1200 K | Continuous-energy neutron | 1e-11 – 20 MeV | Typical operating fuel temperature for a PWR pellet. |
| .84c | ENDF/B-VII.1 (ENDF71x) | 2500 K | Continuous-energy neutron | 1e-11 – 20 MeV | Accident-condition fuel temperatures. |
| .85c | ENDF/B-VII.1 (ENDF71x) | 0.1 K | Continuous-energy neutron | 1e-11 – 20 MeV | Effectively unbroadened data, used by the DBRC treatment. |
| .86c | ENDF/B-VII.1 (ENDF71x) | 250 K | Continuous-energy neutron | 1e-11 – 20 MeV | Below room temperature. |
| .70c | ENDF/B-VII.0 (ENDF70) | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Superseded by VII.1 and VIII.0, but still the default in older installations. |
| .71c | ENDF/B-VII.0 (ENDF70) | 600 K | Continuous-energy neutron | 1e-11 – 20 MeV | VII.0 at 600 K. Commonly misread as "ENDF/B-VII.1" — the .1 in the suffix is a temperature index, not a release number. |
| .73c | ENDF/B-VII.0 (ENDF70) | 1200 K | Continuous-energy neutron | 1e-11 – 20 MeV | VII.0 at typical fuel temperature. |
| .66c | ENDF/B-VI (ENDF66) | 293.6 K | Continuous-energy neutron | 1e-11 – 20 MeV | Legacy library. Not recommended for new calculations. |
| .14p | eprdata14 | N/A | Photoatomic + electron | 1 keV – 100 GeV | Photoatomic data with atomic relaxation, the current MCNP6 default for photon transport.(recommended) |
| .12p | eprdata12 | N/A | Photoatomic + electron | 1 keV – 100 GeV | Previous eprdata release. |
| .84p | mcplib84 | N/A | Photoatomic | 1 keV – 100 GeV | Older photoatomic library based on ENDF/B-VI.8. No atomic relaxation data. |
| .20t | ENDF71SaB | 293.6 K | Thermal S(α,β) | < 4 eV | Thermal scattering law for bound atoms. Required for accurate thermal-spectrum transport. Pairs with .80c neutron data.(recommended) |
| .21t | ENDF71SaB | 350 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 350 K. |
| .22t | ENDF71SaB | 400 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 400 K. |
| .23t | ENDF71SaB | 450 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 450 K. |
| .24t | ENDF71SaB | 500 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 500 K. |
| .26t | ENDF71SaB | 600 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 600 K — the usual choice for PWR coolant. |
| .28t | ENDF71SaB | 800 K | Thermal S(α,β) | < 4 eV | S(α,β) data at 800 K. |
| .10t | endf70sab | 293.6 K | Thermal S(α,β) | < 4 eV | Older thermal set that pairs with .70c neutron data. |
| .16t | endf70sab | 600 K | Thermal S(α,β) | < 4 eV | endf70sab at 600 K. |
| .80t | ENDF80SaB | 293.6 K | Thermal S(α,β) | < 4 eV | ENDF/B-VIII.0 thermal set. Note it also renames the tables: lwtr became h-h2o, poly became h-poly. |
| varies (c) | JEFF-3.3 | as processed | Continuous-energy neutron | 1e-11 – 20 MeV | Latest JEFF release; strong for structural materials and fission products. JEFF is not distributed by LANL, so the suffix is assigned by whoever processed the ACE files (KAERI's KN-series uses .74c for JEFF-3.2, for example). Read your xsdir. |
| varies (c) | JENDL-5 | as processed | Continuous-energy neutron | 1e-11 – 200 MeV | Comprehensive updates to minor actinides; strong for fast-reactor work. JAEA distributes it as ACE-J50 with its own suffixes (.20c at 300 K, .21c at 600 K, and so on) that collide with other vendors' numbering. Read your xsdir. |
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 table temperature to your material temperature — but read the two columns above separately, because the same two digits mean different temperatures in each. In MCNP the digits are a temperature index on the same ladder as the neutron data: ENDF71SaB, which pairs with .80c, runs .20t at 293.6 K, .22t at 400 K, and .26t at 600 K, so PWR coolant is .26t and not .22t. The older endf70sab set that pairs with .70c covers the same temperatures at .10t–.18t. Serpent's JEFF-processed tables number independently, which is where .11t (~300 K) and .22t (~600 K) come from.
The MCNP names above are the ENDF/B-VII.1 ones. ENDF/B-VIII.0 renamed several tables — lwtr became h-h2o and poly became h-poly — so a VIII.0 deck cannot reuse a VII.1 mt card verbatim. Your xsdir lists the names your install actually has.
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