**1. FILE 1. GENERAL INFORMATION**

File 1 is the first part of any set of evaluated cross-section data for
a material. Each material must have a File 1 that contains at least one
section. This required section provides a brief documentation of how the
data were evaluated and a directory that summarizes the files and
sections contained in the material. In the case of fissionable
materials, File 1 may contain up to four additional sections giving
fission neutron yields and energy release information. Each section has
been assigned an MT number (see below), and the sections are arranged in
order of increasing MT number. A section always starts with a HEAD
record and ends with a SEND record. The end of File 1 (and all other
files) is indicated by a FEND record. These record types are defined in
detail in Section 0.7.

**1.1. Descriptive Data and Directory (MT=451)**

This section is always the first section of any material and has two
parts:

1)  a brief documentation of the cross-section data, and

2)  a directory of the files and sections used for this material.

In the first part, a brief description of the evaluated data set is
given. This information should include the significant experimental
results used to obtain the evaluated data, descriptions of any nuclear
models used, a clear specification of all the MT numbers defined to
identify reactions, the history of the evaluation, and references. The
descriptive information is given as a series of records, each record
containing up to 66 characters.

The first three records of the descriptive information contain a
standardized presentation of information on the material, projectile,
evaluators, and modification status. The following quantities are
defined for MF=1, MT=451:

+-----------+----------------------------------------------------------+
| *         | Standard material charge and mass parameters.            |
| *ZA,AWR** |                                                          |
+-----------+----------------------------------------------------------+
| **LRP**   | Indicates whether resolved and/or unresolved resonance   |
|           | parameters are given in File 2:                          |
|           |                                                          |
|           | LRP=1, no File 2 is given (not allowed for incident      |
|           | neutrons);                                               |
|           |                                                          |
|           | LRP=0, no resonance parameter data are given, but a File |
|           | 2 is present containing the effective scattering radius; |
|           |                                                          |
|           | LRP=1, resolved and/or unresolved parameter data are     |
|           | given in File 2 and cross sections computed from them    |
|           | must be added[^1] to background cross sections given in  |
|           | File 3;                                                  |
|           |                                                          |
|           | LRP=2, parameters are given in File 2, but cross         |
|           | sections derived from them are not to be added to the    |
|           | cross sections in File 3. The option LRP=2 is to be used |
|           | for derived files only.                                  |
+-----------+----------------------------------------------------------+
| **LFI**   | Indicates whether this material fissions:                |
|           |                                                          |
|           | LFI=0, this material does not fission;                   |
|           |                                                          |
|           | LFI=1, this material fissions.                           |
+-----------+----------------------------------------------------------+
| **NLIB**  | Library identifier: NLIB= 0 for ENDF/B.                  |
|           |                                                          |
|           | Additional values have been assigned to identify other   |
|           | libraries using ENDF format.                             |
+-----------+----------------------------------------------------------+
| **NMOD**  | Modification number for this material:                   |
|           |                                                          |
|           | NMOD=0, evaluation converted from a previous version;    |
|           |                                                          |
|           | NMOD=1, new or revised evaluation for the current        |
|           | library version;                                         |
|           |                                                          |
|           | NMOD2, for successive modifications.                     |
+-----------+----------------------------------------------------------+
| **ELIS**  | Excitation energy of the target nucleus relative to 0.0  |
|           | for the ground state.                                    |
+-----------+----------------------------------------------------------+
| **STA**   | Target stability flag:                                   |
|           |                                                          |
|           | STA=0, stable nucleus;                                   |
|           |                                                          |
|           | STA=1 unstable nucleus. If the target is unstable,       |
|           | radioactive decay data should be given in the decay data |
|           | sub-library (NSUB=4).                                    |
+-----------+----------------------------------------------------------+
| **LIS**   | State number of the target nucleus. The ground state is  |
|           | indicated by LIS=0.                                      |
+-----------+----------------------------------------------------------+
| **LISO**  | Isomeric state number. The ground state is indicated by  |
|           | LISO=0. LIS is greater than or equal to LISO.            |
+-----------+----------------------------------------------------------+
| **NFOR**  | Library format.                                          |
|           |                                                          |
|           | NFOR=6 for all libraries prepared according to the       |
|           | specifications given in this manual.                     |
+-----------+----------------------------------------------------------+
| **AWI**   | Mass of the projectile in neutron units. For incident    |
|           | photons or decay data sub-libraries, use AWI=0.0.        |
+-----------+----------------------------------------------------------+
| **EMAX**  | Upper limit of energy range for evaluation.              |
+-----------+----------------------------------------------------------+
| **LREL**  | Release number.                                          |
+-----------+----------------------------------------------------------+
| **NSUB**  | Sub-library number. See Section 0.4 for a description of |
|           | sub-libraries.                                           |
+-----------+----------------------------------------------------------+
| **NVER**  | Library version number; for example, NVER=6 for version  |
|           | VI.                                                      |
+-----------+----------------------------------------------------------+
| **TEMP**  | Target temperature (Kelvin) for data that have been      |
|           | generated by Doppler broadening. For derived data only;  |
|           | use TEMP=0.0 for all primary evaluations.                |
+-----------+----------------------------------------------------------+
| **LDRV**  | Special derived material flag that distinguishes between |
|           | different evaluations with the same material keys        |
|           | (*i.e.*, MAT, NMOD, NSUB): LDRV=0, primary evaluation:   |
|           |                                                          |
|           | LDRV1, special derived evaluation (for example, a        |
|           | dosimetry evaluation using sections (MT) extracted from  |
|           | the primary evaluation).                                 |
+-----------+----------------------------------------------------------+
| **NWD**   | Number of records used to describe the data set for this |
|           | material. Each record can contain up to 66 characters.   |
+-----------+----------------------------------------------------------+
| **NXC**   | Number of records in the directory for this material.    |
|           | Each section (MT) in the material has a corresponding    |
|           | line in the directory that contains MF, MT, NC, and MOD. |
|           | NC is a count of the number of records in the section    |
|           | (not including SEND), and MOD is the modification flag   |
|           | (see below).                                             |
+-----------+----------------------------------------------------------+
| *         | Character representation of the material charge,         |
| *ZSYNAM** | chemical symbol, atomic mass number, and metastable      |
|           | state in the form Z-cc-AM with                           |
|           |                                                          |
|           | > Z, right justified in col. 1 to 3,                     |
|           | >                                                        |
|           | > \- (hyphen) in col. 4,                                 |
|           | >                                                        |
|           | > two-character chemical name left justified in col. 5   |
|           | > and 6,                                                 |
|           | >                                                        |
|           | > \- (hyphen) in col. 7,                                 |
|           | >                                                        |
|           | > A, right justified in col. 8 to 10 or blank,           |
|           | >                                                        |
|           | > M for the indication of a metastable state in col. 11, |
|           | >                                                        |
|           | > for example, 94-PU-239, 1-H - 2, *etc*.                |
+-----------+----------------------------------------------------------+
| **ALAB**  | Mnemonic for the originating laboratory(s) left adjusted |
|           | in col. 12-22.                                           |
+-----------+----------------------------------------------------------+
| **EDATE** | Date of evaluation given in the form \"EVAL-DEC74\" in   |
|           | col. 23-32.                                              |
+-----------+----------------------------------------------------------+
| **AUTH**  | Author(s) name(s) left adjusted in col. 34-66.           |
+-----------+----------------------------------------------------------+
| **REF**   | Primary reference for the evaluation left adjusted in    |
|           | col. 2-22.                                               |
+-----------+----------------------------------------------------------+
| **DDATE** | Original distribution date given in the form             |
|           | \"DIST-DEC74\" in col. 23-32.                            |
+-----------+----------------------------------------------------------+
| **RDATE** | Date and number of the last revision to this evaluation  |
|           | in col. 34-43 in the form \"REV2-DEC74", where \"2\" is  |
|           | the revision number IREV; IREV is computer retrievable.  |
+-----------+----------------------------------------------------------+
| *         | Master File entry date in the form *yyyymmdd* right      |
| *ENDATE** | adjusted in col. 56-63. The Master File entry date will  |
|           | be assigned by NNDC for ENDF/B-VI.                       |
+-----------+----------------------------------------------------------+
| **HSUB**  | Identifier for the library contained on three successive |
|           | records.                                                 |
|           |                                                          |
|           | The first record contains four dashes starting in col.   |
|           | 1, directly followed by the library type (NLIB) and      |
|           | version (NVER). For example,                             |
|           |                                                          |
|           | \"\-\-\-- ENDF/B-VI\", followed by MATERIAL XXXX         |
|           | starting in col. 23 where XXXX is the MAT number, and    |
|           | REVISION 2 (starting in col. 45 only if required) where  |
|           | \"2\" is the revision number IREV.                       |
|           |                                                          |
|           | The second record contains five dashes starting in col.  |
|           | 1, and followed by the sub-library identifier (see Table |
|           | 0.1). For example,                                       |
|           |                                                          |
|           | \"\-\-\-\-- DECAY DATA,\"                                |
|           |                                                          |
|           | \"\-\-\-\-- PHOTO-ATOMIC INTERACTION DATA,\" or          |
|           |                                                          |
|           | \"\-\-\-\-- INCIDENT NEUTRON DATA.\"                     |
|           |                                                          |
|           | The third record contains six dashes starting in col. 1  |
|           | and followed by ENDF-6 where \"6\" is the library format |
|           | type (NFOR).                                             |
|           |                                                          |
|           | Note: the three HSUB records can be generated by a       |
|           | utility program.                                         |
+-----------+----------------------------------------------------------+
| **MF~n~** | MF of the n^th^ section.                                 |
+-----------+----------------------------------------------------------+
| **MT~n~** | MT of the n^th^ section.                                 |
+-----------+----------------------------------------------------------+
| **NC~n~** | Number of records in the nth section. This count does    |
|           | not include the SEND record.                             |
+-----------+----------------------------------------------------------+
| *         | Modification indicator for the nth section. The value of |
| *MOD~n~** | MOD~n~ is equal to NMOD if the corresponding section was |
|           | changed in this revision. MOD~n~ must always be less     |
|           | than or equal to NMOD.                                   |
+-----------+----------------------------------------------------------+

**1.1.1. Formats**

The structure of this section is

\[MAT, 1,451/ ZA, AWR, LRP, LFI, NLIB, NMOD\]HEAD

\[MAT, 1,451/ ELIS, STA, LIS, LISO, 0, NFOR\]CONT

\[MAT, 1,451/ AWI, EMAX, LREL, 0, NSUB, NVER\]CONT

\[MAT, 1,451/ TEMP, 0.0, LDRV, 0, NWD, NXC\]CONT

\[MAT, 1,451/ZSYMAM, ALAB, EDATE, AUTH \]TEXT

\[MAT, 1,451/ REF, DDATE, RDATE, ENDATE \]TEXT

\[MAT, 1,451/ HSUB \]TEXT

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

continue for the rest of the NWD descriptive records

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

\[MAT, 1,451/ blank, blank, MF1, MT1, NC1, MOD1\]CONT

\[MAT, 1,451/ blank, blank, MF2, MT2, NC2, MOD2\]CONT

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

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

\[MAT, 1,451/ blank, blank, MFNXC, MTNXC,NCNXC,MODNXC\]CONT

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

**1.1.2. Procedures**

Note that the parameters NLIB, NVER, NSUB, MAT, NMOD, LDRV, and
sometimes TEMP define a unique set of \"keys\" that identifies a
particular evaluation or \"material\" in the ENDF system. These keys can
be used to access materials in a formal data base management system if
desired.

The flag LRP indicates whether resolved and/or unresolved resonance
parameter data are to be found in File 2 (Resonance Parameters) and how
these data are to be used with File 3 to compute the net cross section.
For incident neutrons, every material will have a File 2. If LRP=0, the
file contains only the effective scattering radius; the potential cross
section corresponding to this scattering radius has already been
included in the File 3 cross sections. If LRP=1, File 2 contains
resolved and/or unresolved resonance parameters. Cross sections or
self-shielding factors computed from these parameters are to be combined
with any cross sections found in File 3 to obtain the correct net cross
section. For other sub-libraries (decay data, incident photons, incident
charged particles, fission product yields), File 2 can be omitted (use
LRP=-1). A number of processing codes exist which reconstruct
resonance-region cross sections from the parameters in File 2 and output
the results in ENDF format. Such a code can set LRP=2 and copy the
original File 2 to its output ENDF tape. Other processing codes using
such a tape will know that resonance reconstruction has already been
performed, but the codes will still have easy access to the resonance
parameters if needed. The LRP=2 option is not allowed in primary
evaluations.

The flag LFI indicates that this material fissions in the context of the
present sub-library. In this case, a section specifying the total number
of neutrons emitted per fission,(E), *must* be given as MF=1, MT=452.
Sections *may* also be given that specify the number of delayed neutrons
per fission (MT=455) and the number of prompt neutrons per fission
(MT=456), and that specify the components of energy release in fission
(MT=458).

The flag LDRV indicates that this material was derived in some way from
another evaluation; for example, it could represent an activation
reaction extracted from a more complete evaluation, it could be part of
a gas production library containing production cross sections computed
from more fundamental reactions, it could represent a reconstructed
library with resonance parameters expanded into detailed point-wise
cross sections, and so on.

The data in the descriptive section must be given for every material.
The first three records are used to construct titles for listings,
plots, *etc*., and the format should be followed closely. The remaining
records give a verbal description of the evaluated data set for the
material. The description should mention the important experimental
results upon which the recommended cross sections are based, the
evaluation procedures and nuclear models used, a brief history and
origin of the evaluation, important limitations of the data set,
estimated uncertainties and covariances, references, and any other
remarks that will assist the user in understanding the data. For
incident neutron evaluations, the 2200 m/s cross sections contained in
the data should be tabulated, along with the infinite dilution resonance
integrals for capture and fission (if applicable). For charged-particle
and high-energy reactions, the meaning of each MT should be carefully
explained using the notation of Section 0.5.

**1.2. Number of Neutrons per Fission,, (MT=452)**

If the material fissions (LFI=1), then a section specifying the average
*total* number of neutrons per fission, (MT=452), must be given. This
format applies to both particle induced

![](media/image1.wmf)

and spontaneous fission, each in its designated sub-library. Values of
may be tabulated as a function of energy or coefficients provided for
the following polynomial expansion of $\overline{\nu}$(E),

> where (E) = the average total (prompt plus delayed) number of neutrons
> per fission produced by neutrons of incident energy E(eV),
>
> C~n~ = the n^th^ coefficient, and
>
> NC = the number of terms in the polynomial.

MT=452 for an energy-dependent neutron multiplicity cannot be
represented by a polynomial expansion when MT=455 and MT=456 are
utilized in the file.

**1.2.1. Formats**

The structure of this section depends on whether values of (E) are
tabulated as a function of energy or represented by a polynomial. The
following quantities are defined:

+-----------+----------------------------------------------------------+
| **LNU**   | Test that indicates what representation of               |
|           | $\overline{\nu}$ (E) has been used:                      |
|           |                                                          |
|           | LNU=1, polynomial representation has been used;          |
|           |                                                          |
|           | LNU=2, tabulated representation.                         |
+-----------+----------------------------------------------------------+
| **NC**    | Count of the number of terms used in the polynomial      |
|           | expansion. (NC4).                                        |
+-----------+----------------------------------------------------------+
| **C~n~**  | Coefficients of the polynomial. There are NC             |
|           | coefficients given.                                      |
+-----------+----------------------------------------------------------+
| **NR**    | Number of interpolation ranges used to tabulate values   |
|           | of (E). (See 0.6.2.)                                     |
+-----------+----------------------------------------------------------+
| **NP**    | Total number of energy points used to tabulate(E).       |
+-----------+----------------------------------------------------------+
| *         | Interpolation scheme (see 0.6.2 for details).            |
| *E~int~** |                                                          |
+-----------+----------------------------------------------------------+
| **(E)**   | Average number of neutrons per fission.                  |
+-----------+----------------------------------------------------------+

If **LNU=1**, the structure of the section is

\[MAT, 1, 452/ ZA, AWR, 0, LNU, 0, 0\]HEAD (LNU=1)

\[MAT, 1, 452/ 0.0, 0.0, 0, 0, NC, 0/ C1, C2, \...CNC\]LIST

\[MAT, 1,99999/ 0.0, 0.0, 0, 0, 0, 0\]SEND

If **LNU=2**, the structure of the section is

\[MAT, 1, 452/ ZA, AWR, 0, LNU, 0, 0\]HEAD (LNU=2)

\[MAT, 1, 452/ 0.0, 0.0, 0, 0, NR, NP/E~int~ /(E)\]TAB1

\[MAT, 1,99999/ 0.0, 0.0, 0, 0, 0, 0\]SEND

**1.2.2. Procedures**

If a polynomial representation (LNU=1) has been used to specify (E),
this representation is valid over any range in which the fission cross
section is specified (as given in Files 2 and 3). When using a
polynomial to fit (E), the fit shall be limited to a third-degree
polynomial (NC4). If such a fit does not reproduce the recommended
values of (E), a tabulated form (LNU=2) should be used.

If tabulated values of (E) are specified (LNU=2), then pairs of energy-
values are given. Values of (E) should be given that cover any energy
range in which the fission cross section is given in File 2 and/or File
3.

The values of (E) given in this section are for the average total number
of neutrons produced per fission event. When another section (MT=455)
that specifies the delayed neutrons from fission is given, the average
number of delayed neutrons per fission (~d~) must be added to ~p~ given
in section MT=456 and included in the value of(E) given in this section
(MT=452). In this case, only LNU=2 is allowed for MT=452.

For spontaneous fission, the polynomial representation (LNU=1) is used
with NC=1 and Cl = ~total~. There is no energy dependence.

**1.3. Delayed Neutron Data, ~d~, (MT=455)**

This section describes the delayed neutrons resulting from either
particle induced or spontaneous fission. The average total number of
delayed neutron precursors emitted per fission,~d~, is given, along with
the decay constants, ~i~, for each precursor family. The fraction of~d~
generated for each family is given in File 5 (section 5 of this report).
The energy distributions of the neutrons associated with each precursor
family are also given in File 5.

For particle-induced fission, the total number of delayed neutrons is
given as a function of energy in tabulated form (LNU=2). The energy
dependence is specified by tabulating~d~(E) at a series of neutron
energies using the same format as for MT=452. For spontaneous fission
LNU=1 is used with NC=1 and C~1~=~d~ as for MT=452.

![](media/image2.wmf)The total number of delayed neutron precursors
emitted per fission event, at incident energy E, is given in this file
and is defined as the sum of the number of neutrons emitted for each of
the precursor families,

$P_{i}(E) = \frac{{\overline{\nu}}_{i}(E)}{{\overline{\nu}}_{d}(E)}$where
*NNF* is the number of precursor families. The fraction of the total,
P~i~(E), emitted for each family is given in File 5 (see section 5) and
is defined as

**1.3.1. Formats**

The following quantities are defined.

+------------+---------------------------------------------------------+
| **LNU**    | Test indicating which representation is used:           |
|            |                                                         |
|            | LNU=1 means that polynomial expansion is used;          |
|            |                                                         |
|            | LNU=2 means that a tabulated representation is used.    |
+------------+---------------------------------------------------------+
| **NC**     | Number of terms in the polynomial expansion. (NC4).     |
+------------+---------------------------------------------------------+
| **NR**     | Number of interpolation ranges used. (NR20).            |
+------------+---------------------------------------------------------+
| **NP**     | Total number of energy points used in the tabulation of |
|            | (E)                                                     |
+------------+---------------------------------------------------------+
| **E~int~** | Interpolation scheme (See Section 0.6.2)                |
+------------+---------------------------------------------------------+
| $\overl    | Total average number of delayed neutrons formed per     |
| ine{\nu_{d | fission event.                                          |
| }}$**(E)** |                                                         |
+------------+---------------------------------------------------------+
| **NNF**    | Number of precursor families considered.                |
+------------+---------------------------------------------------------+
| **~i~**    | Decay constant (sec^-1^) for the i^th^ precursor.       |
+------------+---------------------------------------------------------+

The structure when values of~d~ are tabulated (LNU=2) is:

\[MAT, 1,455/ ZA, AWR, 0, LNU, 0, 0\]HEAD (LNU=2)

\[MAT, 1,455/ 0.0, 0.0, 0, 0, NNF, 0/1,2,\...~NNF~\]LIST

\[MAT, 1,455/ 0.0, 0.0, 0, 0, NR, NP/ E~int~ /~d~(E)\]TAB1

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

If LNU=1 (spontaneous fission) the structure of the section is:

\[MAT, 1,455/ ZA, AWR, 0, LNU, 0, 0\]HEAD (LNU=1)

\[MAT, 1,455/ 0.0, 0.0, 0, 0, NNF, 0/~1~,~2~,\...~NNF~\]LIST

\[MAT, 1,455/ 0.0, 0.0, 0, 0, 1, 0/~d~\]LIST

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

**1.3.2. Procedures**

When tabulated values of~d~(E) are specified, as is required for
particle-induced fission by Section 1.2., they should be given for the
same energy range as that used to specify the fission cross section.

The probability of producing the precursors for each family and the
energy distributions of neutrons produced by each precursor family are
given in File 5 (section 5 of this report). It is extremely important
that the same precursor families be given in File 5 as are given in File
1 (MT=455), and the ordering of the families should be the same in both
files. It is recommended that the families be ordered by decreasing
half-lives (~1~\<~2~\<...\<~NNF~).

For spontaneous fission, the polynomial form (LNU=1) is used with only
one term (NC=1, C~1~=~d~).

If MT=455 is used, then MT=456 must also be used as well as MT=452.

**1.4. Number of Prompt Neutrons per Fission,~p~, (MT=456)**

If the material fissions (LFI=1), a section specifying the average
number of prompt neutrons per fission,~p~, (MT=456) can be given using
formats identical to MT=452. For particle-induced fission,~p~ is given
as a function of energy. The prompt for spontaneous fission can also be
given using MT=456, but there is no energy dependence.

**1.4.1. Formats**

The following quantities are defined:

+-----------+----------------------------------------------------------+
| **LNU**   | Indicates what representation of~p~(E) has been used:    |
|           |                                                          |
|           | LNU=1, polynomial representation has been used;          |
|           |                                                          |
|           | LNU=2, tabulated representation.                         |
+-----------+----------------------------------------------------------+
| **NC**    | Count of the number of terms used in the polynomial      |
|           | expansion. (NC4)                                         |
+-----------+----------------------------------------------------------+
| **NR**    | Number of interpolation ranges used to tabulate values   |
|           | of~p~(E). (See 0.7.7.)                                   |
+-----------+----------------------------------------------------------+
| **NP**    | Total number of energy points used to tabulate~p~(E).    |
+-----------+----------------------------------------------------------+
| *         | Interpolation scheme (see section 0.6.2.)                |
| *E~int~** |                                                          |
+-----------+----------------------------------------------------------+
| *         | Average number of prompt neutrons per fission.           |
| *~p~(E)** |                                                          |
+-----------+----------------------------------------------------------+

If LNU=2, (tabulated values of $\overline{\nu}$), the structure of the
section is:

\[MAT, 1,456/ ZA, AWR, 0, LNU, 0, 0\] HEAD (LNU=2)

\[MAT, 1,456/ 0.0, 0.0, 0, 0, NR, NP/E~int~/~p~(E)\]TAB1

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

If LNU=1 (spontaneous fission) the structure of the section is:

\[MAT, 1,456/ ZA, AWR, 0, LNU, 0, 0\]HEAD (LNU=1)

\[MAT, 1,456/ 0.0, 0.0, 0, 0, 1, 0/~p~\]LIST

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

**1.4.2. Procedures**

If tabulated values of~p~(E) are specified (LNU=2), then pairs of
energy-$\overline{\nu}$ values are given. Values of ~p~(E) should be
given that cover any energy range in which the fission cross section is
given in File 2 and/or File 3. The values of ~p~(E) given in this
section are for the average number of prompt neutrons produced per
fission event. The energy independent~p~ for spontaneous fission is
given using LNU=1 with NC=1 and C~1~=${\overline{\nu}}_{p}$as described
for MT = 452.

If MT=456 is specified, then MT=455 must also be specified as well as
MT=452.

**1.5. Components of Energy Release Due to Fission (MT=458)**

The energy released in fission is carried by fission fragments,
neutrons, gammas, betas (+ and --), and neutrinos and anti-neutrinos.
The term fragments includes all charged particles that are emitted
promptly, since for energy-deposition calculations, all such particles
have short ranges and are usually considered to lose their energy
locally. Neutrons and gammas transport their energy elsewhere and need
to be considered separately. In addition, some gammas and neutrons are
delayed, and in a shut-down assembly, one needs to know the amount of
energy tied up in these particles and the rate at which it is released
from the metastable nuclides or precursors. The neutrino energy is lost
completely in most applications, but is part of the Q-value. As far as
the betas are concerned, prompt betas, being charged, deposit their
energy locally with the fragments, and their prompt energies are
correctly included with the fragment energies.

  ------------ ----------------------------------------------------------
  **ET**       Sum of all the partial energies that follow. This sum is
               the total energy release per fission and equals the Q
               value.

  **EFR**      Kinetic energy of the fragments.

  **ENP**      Kinetic energy of the \"prompt\" fission neutrons.

  **END**      Kinetic energy of the delayed fission neutrons.

  **EGP**      Total energy released by the emission of \"prompt\" rays.

  **EGD**      Total energy released by the emission of delayed rays.

  **EB**       Total energy released by delayed \'s.

  **ENU**      Energy carried away by the neutrinos.

  **ER**       Total energy less the energy of the neutrinos (ET - ENU);
               equal to the pseudo-Q in File 3 for MT=18.
  ------------ ----------------------------------------------------------

All of these energies are given for an incident energy of zero.[^2]

![](media/image3.wmf)

where E~i~ is any of the energy release components;

> E~i~(0) is the value at E~inc~ = 0; E~inc~ = 0 is fictitious and
> represents an artifice by which it is possible to recover the values
> at any E~inc~.

E~i~(E~inc~) is the value at incident energy E~inc~.

![](media/image4.wmf)The E~i~\'s are given by the following:

**1.5.1. Formats**

The structure of this section always starts with a HEAD record and end
with a SEND record. The section contains no subsections and only one
LIST record.

The structure of a section is:

\[MAT, 1,458/ ZA, AWR, 0, 0, 0, 0\]HEAD

\[MAT, 1,458/ 0.0, 0.0, 0, 0, 18, 9/

EFR, EFR, ENP, ENP, END, END,

EGP, EGP, EGD, EGD, EB, EB,

ENU, ENU, ER, ER, ET, ET\]LIST

\[MAT, 1, 0/ 0.0, 0.0, 0, 0, 0, 0\]SEND

where the \'s allow the error estimates on the quantities listed above.

**1.5.2. Procedures**

This section should be used for fertile and fissile isotopes only.

Consistency should be maintained between the Q values in File 3, the
energies calculated from File 5 and 15 and the energies listed in
File 1. Note that ER = the pseudo-Q for fission (MT=18) in File 3.

Other components are not so readily determined or checked. The procedure
should be that File 5 and File 15 data take precedent, whenever
available. That is, \"prompt\" fission neutron energy calculated from
File 5 spectra from MT=18 should be used in File 1; the same holds true
for the delayed neutron spectra given in File 5, MT=455. The \"prompt\"
gamma energy calculated from File 15 (MT=18 for fission) should be input
into File 1, that is the prompt gammas due to the fission process.

These quantities should be calculated at the lowest energy given in the
Files for MT=18 except for fissile isotopes for which the thermal
spectra should be used. For fertile materials, the spectrum given at
threshold would be appropriate. Note that the File 5 spectra for MT=18
should be used with prompt (not total) for the fission neutrons. MT=455
in File 5 contains the delayed fission neutron spectra.

In many reactor applications, time dependent energy deposition rates are
required rather than the components of the total energy per fission
which are the values given in this MT. Time-dependent energy deposition
parameters can be obtained from the six-group spectra in File 5 (MT=455)
for delayed neutrons. Codes such as CINDER, RIBD, and ORIGEN must be
used, however, to obtain more detailed information on the delayed
neutrons and all time-dependent parameters for the betas and the gammas
due to the fission process.

The time-integrated energies for delayed neutrons, delayed gammas, and
delayed betas as calculated from the codes listed above may not always
agree with the energy components given in File 1. The File 1 components
must sum to ET (the total energy released per fission).

In heating calculations, the energy released in all nuclear reactions
besides fission, principally the gamma-energy released in neutron
radiative capture, enters analogously to the various fission energy
components. Thus the (n,) energy-release would be equal to the Q-value
in File 3, MT=102, of the capturing nuclide. The capture gammas can be
prompt or delayed, if branching to isomeric states is involved, and this
is relevant to various fission- and burnup-product calculations. The
\"sensible energy\" in a heating calculation is the sum of ER, defined
previously, and the energy released in all other reactions.

[^1]: In the unresolved region, it is also possible to compute a
    self-shielding factor from File 2 and multiply it by a complete
    unshielded cross section in File 3.

[^2]: Taken from R. Sher and C. Beck, Fission Energy Release for 16
    Fissioning Nuclides, EPRI-NP-1771 (1981).