Target
In the ENDF or SIGMA interface the word 'target' is used in the usuall nuclear physics meaning, i.e., it specifies a nuclide with which an incident paticle (projectile) is interacting. We choose 'target' instead of the ENDF-6 term 'material' since we believe the former is more familiar and informative to the users with nuclear physics background. We admit, however, that for certain sublibraries, such as decay sublibrary or spontanous fission yields sublibrary, the term 'material' is more appropriate since target implies existence of a projectile, which is not involved in these specific cases.
Reaction
List of reactions is closely related to MT numbers. User typically requests n,*; n,tot; n,g; n,f; n,el; n,inl; n,2n; n,a; n,nu*; p,*;�, where n = neutron, * = anything, tot= total, g = gamma, f = fission, el = elastic, inl = inelastic, a = alpha, nu* = nu-bar with all options (prompt, total, delayed), p = proton.
MT # Range | Reaction Description for Incident Neutrons |
1-100 | Cross sections for emission of neutrons. |
101-150 | Cross sections for emission of photons and charged particles. |
151-200 | Neutron resonance parameters. |
201-450 | Total production cross sections and scattering parameters. |
451-699 | Information in text, fission data, decay data, nu_bar, ... |
600-999 | Partial cross sections for emission of charged particles. |
List of neutron-induced reactions (MT#) is presented below. Neutrons are selected as default projectiles among protons and other particles.
MT # | Reaction | Reaction Description |
1 | n,tot | Neutron total cross sections |
2 | n,el | Elastic neutron scattering cross sections |
3 | n,non | Nonelastic neutron cross sections. Sum of MT=4, 5, 16-18, 22-26, 28-37,41, 42, 102-116. |
4 | n,inl | Inelastic neutron cross sections. Sum of the MT=50-91. |
5 | n,x | Cross section for the sum of all reactions not given explicitly in another MT number. |
6-9 | Not allowed in version 6 except for compatibility with JENDL-3.2 | |
10 | n,tot | Total neutron-induced continuum reactions. |
11 | n,2n+d | Production of two neutrons and a deutron. |
12-15 | Unassigned | |
16 | n,2n | Cross section for producing two neutrons and a residual. |
17 | n,3n | Cross section for producing three neutrons and a residual. |
18 | n,f | Neutron fission cross section, MT=18 is the sum of MT=19-21, 38. |
19 | n,f | First chance neutron fission cross section. |
20 | n,n+f | Second-chance neutron fission cross section. |
21 | n,2n+f | Third-chance neutron fission cross section. |
22 | n,n+a | Cross section for production of a neutron and an alpha particle. |
23 | n,n+3a | Cross section for production of a neutron and three alpha particles. |
24 | n,2n+a | Cross section for production of two neutrons and an alpha particle. |
25 | n,3n+a | Cross section for production of three neutrona and an alpha particle. |
26 | Not allowed in version 6. That was (n,2n) isomeric state cross section in version 5. | |
27 | n,abs | Neutron absorption cross section. Sum of MT=18 and MT=102-116. |
28 | n,n+p | Cross section for production of a neutron and a proton |
29 | n,n+2a | Cross section for production of a neutron and two alpha particles. |
30 | n,2n+2a | Cross section for production of two neutrons and two alpha particles. |
31 | Not allowed in version 6. | |
32 | n,n+d | Cross section for production of a neutron and a deutron. |
33 | n,n+t | Cross section for production of a neutron and a triton. |
34 | n,n+he3 | Cross section for production of a neutron and a 3He. |
35 | n,nd+2a | Cross section for production of a neutron and deutron and two alpha particles. |
36 | n,nt+2a | Cross section for production of a neutron and triton and two alpha particles. |
37 | n,4n | Cross section for production of four neutrons. |
38 | n,3n+f | Fourth chance neutron fission cross section. |
39 | Not allowed in version 6. | |
40 | Not allowed in version 6. | |
41 | n,2n+p | Cross section for production of two neutrons and a proton. |
42 | n,3n+p | Cross section for production of three neutrons and a proton. |
43 | Unassigned. | |
44 | n,n+2p | Cross section for production of a neutron and two protons. |
45 | n,n+pa | Cross section for production of neutron, proton and alpha particle. |
46-49 | Not allowed in version 6. | |
50 | n,n0 | Cross section for production of a neutron, with residual in the ground state. |
51 | n,n1 | Cross section for production of a neutron, with residual in the 1st excited state. |
52 | n,n2 | Cross section for production of a neutron, with residual in the 2nd excited state. |
53-90 | n,n' | Cross section for production of a neutron, with residual in the 3-40th excited state. |
91 | n,nc | Cross section for production of a neutron in the continnum not included in the above discrete representation. |
92-100 | Unassigned. | |
101 | n,dis | Neutron disappearance cross section. |
102 | n,g | Neutron radiative capture cross section. |
103 | n,p | Cross section for production of a proton. If MT=600-649 are present, the MT103 is their sum. |
104 | n,d | Cross section for production of a deutron. If MT=650-699 are present, the MT104 is their sum. |
105 | n,t | Cross section for production of a triton. If MT=700-749 are present, the MT105 is their sum. |
106 | n,3He | Cross section for production of 3He. If MT=750-799 are present, the MT106 is their sum. |
107 | n,a | Cross section for production of an alpha particle. If MT=800-849 are present, the MT107 is their sum. |
108 | n,2a | Cross section for production of two alpha particles. |
109 | n,3a | Cross section for production of tree alpha particles. |
110 | Unassigned. | |
111 | n,2p | Cross section for production of two protons. |
112 | n,p+a | Cross section for for production of a proton and an alpha particle. |
113 | n,t+2a | Cross section for for production of a triton and two alpha particles. |
114 | n,d+2a | Cross section for for production of a deutron and two alpha particles. |
115 | n,p+d | Cross section for for production of a proton and a deutron. |
116 | n,p+t | Cross section for for production of a proton and a triton. |
117 | n,d+a | Cross section for for production of a deutron and an alpha particle. |
118-119 | Unassigned. | |
120 | Not allowed in version 6. | |
121-150 | Unassigned. | |
151 | n,res | Neutron resonance parameters that can be used to calculate cross sections at different temperatures in the resolved and unresolved energy ranges. |
152-200 | Unassigned. | |
201 | n,xn | Total neutron production cross section. |
202 | n,xg | Total gamma production cross section. |
203 | n,xp | Total proton production cross section. |
204 | n,xd | Total deutron production cross section. |
205 | n,xt | Total triton production cross section. |
206 | n,x3He | Total 3He production cross section. |
207 | n,xa | Total alpha particle production cross section. |
208 | n,xpi+ | Total pi+ production cross section. |
209 | n,xpi0 | Total pi0 production cross section. |
210 | n,xpi- | Total pi- production cross section. |
211 | n,xmu+ | Total mu+ production cross section. |
212 | n,xmu- | Total mu- production cross section. |
213 | n,xk+ | Total k+ production cross section. |
214 | n,xk0 | Total k0 (long) production cross section. |
215 | n,xk0 | Total k0 (short) production cross section. |
216 | n,xk- | Total k- production cross section. |
217 | n,xp | Total anti-proton production cross section. |
218 | n,xn | Total anti-neutron production cross section. |
219-250 | Unassigned. | |
251 | n,... | μL (mu bar), the average cosine of scattering angle (laboratory system) for elastic scattering of neutrons. |
252 | n,... | ε, the average logarithmic energy decrement for elastic scattering of neutrons. |
253 | n,... | γ, the average of the square of the logarithmic energy decrement for elastic scattering divided by twice the average logarithmic energy decrement for elastic scattering of neutrons. |
254-300 | Unassigned. | |
301-450 | n,... | Energy release parameters, E, &sigma, for total and partial cross sections. MT=300 plus reaction MT number. |
451 | n,... | Heading or title information; given in File 1 only. |
452 | n,nu_tot | Neutron fission: Nubar (νT), average total (prompt plus delayed) number of neutrons released per fission |
452 | 0,nu_tot | Spontaneous fission: Nubar (νT), average total (prompt plus delayed) number of neutrons released per fission |
453 | Unassigned. | |
454 | n,ind_fy | Neutron fission: Independent fission product yield data |
454 | 0,ind_fy | Spontaneous fission: Independent fission product yield data |
455 | n,nu_d | Neutron fission: Average number of delayed neutrons released per fission event |
456 | n,nu_p | Neutron fission: Average number of prompt neutrons released per fission event |
457 | decay | Radioactive decay data |
458 | e_rel_fis | Neutron fission: Energy release in fission for incident neutrons |
459 | fy_cum | Neutron/spontaneous fission: Cumulative fission product yield data |
460-464 | Unassigned. | |
465-466 | Not allowed in version 6. | |
467-499 | Unassigned. | |
500 | Total charge particle stopping power. | |
501 | Total photon interaction. | |
502 | Photon coherent scattering. | |
503 | Unassigned. | |
504 | Photon incoherent scattering. | |
505 | Imaginary scattering factor. | |
506 | Real scattering factor. | |
507-514 | Unassigned. | |
515 | Pair production, electron field. | |
516 | Pair production, Sum of MT=515,517. | |
517 | Pair production, nuclear field. | |
518 | Not allowed in version 6. | |
519-521 | Unassigned. | |
522 | Photoelectric absorption. | |
523 | Photo-excitation cross section. | |
524-525 | Unassigned. | |
526 | Electro-atomic scattering. | |
527 | Electro-atomic bremsstrahlung. | |
528 | Electro-atomic excitation cross section. | |
529-531 | Unassigned. | |
532 | Not allowed in version 6. | |
533 | Atomic relaxation data. | |
534 | K | (1s1/2) Subshell photoelectric cross section. |
535 | L1 | (2s1/2) Subshell photoelectric cross section. |
536 | L2 | (2p1/2) Subshell photoelectric cross section. |
537 | L3 | (2p3/2) Subshell photoelectric cross section. |
538 | M1 | (3s1/2) Subshell photoelectric cross section. |
539 | M2 | (3p1/2) Subshell photoelectric cross section. |
540 | M3 | (3p3/2) Subshell photoelectric cross section. |
541 | M4 | (3d3/2) Subshell photoelectric cross section. |
542 | M5 | (3d5/2) Subshell photoelectric cross section. |
543 | N1 | (4s1/2) Subshell photoelectric cross section. |
544 | N2 | (4p1/2) Subshell photoelectric cross section. |
545 | N3 | (4p3/2) Subshell photoelectric cross section. |
546 | N4 | (4d3/2) Subshell photoelectric cross section. |
547 | N5 | (4d5/2) Subshell photoelectric cross section. |
548 | N6 | (4f5/2) Subshell photoelectric cross section. |
549 | N7 | (4f7/2) Subshell photoelectric cross section. |
550 | O1 | (5s1/2) Subshell photoelectric cross section. |
551 | O2 | (5p1/2) Subshell photoelectric cross section. |
552 | O3 | (5p3/2) Subshell photoelectric cross section. |
553 | O4 | (5d3/2) Subshell photoelectric cross section. |
554 | O5 | (5d5/2) Subshell photoelectric cross section. |
555 | O6 | (5f5/2) Subshell photoelectric cross section. |
556 | O7 | (5f7/2) Subshell photoelectric cross section. |
557 | O8 | (5g7/2) Subshell photoelectric cross section. |
558 | O9 | (5g9/2) Subshell photoelectric cross section. |
559 | P1 | (6s1/2) Subshell photoelectric cross section. |
560 | P2 | (6p1/2) Subshell photoelectric cross section. |
561 | P3 | (6p3/2) Subshell photoelectric cross section. |
562 | P4 | (6d3/2) Subshell photoelectric cross section. |
563 | P5 | (6d5/2) Subshell photoelectric cross section. |
564 | P6 | (6f5/2) Subshell photoelectric cross section. |
565 | P7 | (6f7/2) Subshell photoelectric cross section. |
566 | P8 | (6g7/2) Subshell photoelectric cross section. |
567 | P9 | (6g9/2) Subshell photoelectric cross section. |
568 | P10 | (6h9/2) Subshell photoelectric cross section. |
569 | P11 | (6h11/2) Subshell photoelectric cross section. |
570 | Q1 | (7s1/2) Subshell photoelectric cross section. |
571 | Q2 | (7p1/2) Subshell photoelectric cross section. |
572 | Q3 | (7p3/2) Subshell photoelectric cross section. |
573-599 | Unassigned. | |
600 | n,p0 | Cross section for production of a proton leaving the residual nucleus in the ground state. |
601 | n,p1 | Cross section for production of a proton leaving the residual nucleus in the 1st excited state. |
602 | n,p2 | Cross section for production of a proton leaving the residual nucleus in the 2nd excited state. |
603-648 | n,p# | Cross section for production of a proton leaving the residual nucleus in the #th excited state. |
649 | n,pc | Cross section for production of a proton in the continuum. |
650 | n,d0 | Cross section for production of a deutron leaving the residual nucleus in the ground state. |
651 | n,d1 | Cross section for production of a deutron leaving the residual nucleus in the 1st excited state. |
652 | n,d2 | Cross section for production of a deutron leaving the residual nucleus in the 2nd excited state. |
653-698 | n,d# | Cross section for production of a deutron leaving the residual nucleus in the #th excited state. |
699 | n,dc | Cross section for production of a deutron in the continuum. |
700 | n,t0 | Cross section for production of a triton leaving the residual nucleus in the ground state. |
701 | n,t1 | Cross section for production of a triton leaving the residual nucleus in the 1st excited state. |
702 | n,t2 | Cross section for production of a triton leaving the residual nucleus in the 2nd excited state. |
703-748 | n,t# | Cross section for production of a triton leaving the residual nucleus in the #th excited state. |
749 | n,tc | Cross section for production of a triton in the continuum. |
750 | n,3He0 | Cross section for production of a 3He particle leaving the residual nucleus in the ground state. |
751 | n,3He1 | Cross section for production of a 3He particle leaving the residual nucleus in the 1st excited state. |
752 | n,3He2 | Cross section for production of a 3He particle leaving the residual nucleus in the 2nd excited state. |
753-798 | n,3He# | Cross section for production of a 3He particle leaving the residual nucleus in the #th excited state. |
799 | n,3Hec | Cross section for production of a 3He particle in the continuum. |
800 | n,a0 | Cross section for production of an alpha particle leaving the residual nucleus in the ground state. |
801 | n,a1 | Cross section for production of an alpha particle leaving the residual nucleus in the 1st excited state. |
802 | n,a2 | Cross section for production of an alpha particle leaving the residual nucleus in the 2nd excited state. |
803-848 | n,a# | Cross section for production of an alpha particle leaving the residual nucleus in the #th excited state. |
849 | n,ac | Cross section for production of an alpha particle in the continuum. |
850 | Unassigned. | |
851-870 | Lumped reaction covariances. | |
871-874 | Unassigned. | |
875 | n,2n0 | Cross section for production of two neutrons leaving the residual nucleus in the ground state. |
876 | n,2n1 | Cross section for production of two neutrons leaving the residual nucleus in the 1st excited state. |
877 | n,2n2 | Cross section for production of two neutrons leaving the residual nucleus in the 2nd excited state. |
878-890 | n,2n# | Cross section for production of two neutrons leaving the residual nucleus in the #th excited state. |
891 | n,2nc | Cross section for production of two neutrons in the continuum. |
892-999 | Unassigned. |
Quantity
List of quantities is defined by MF numbers. User typically requests sig; da; de; res; cov*, where sig = reaction cross-section, de = energy spectrum of emitted particles, da = angular distribution of emitted particles, res = neutron resonance parameters, cov* = covariances with all options (cross sections, nu-bar, neutron resonance parameters).
MF # | Quantity | Quantity Description |
1 | info | General information |
2 | res | Resonance parameters |
3 | sig | Reaction cross section |
4 | da | Angular distribution for emitted particles |
5 | de | Energy distribution for emitted particles |
6 | da/de | Energy-Angle distribution for emitted particles |
7 | sig/ths | Thermal neutron scattering |
8 | rnp | Radioactivity and fission-product yields |
9 | mrnp | Multiplicities for production of radioactive nuclides |
10 | sig/act | Cross section for production of radioactive nuclides |
11 | commg | General comments on photon production |
12 | mult | Multiplicities for photon production |
13 | sigg | Cross sections for photon production |
14 | dag | Angular distribution for emitted photons |
15 | deg | Energy distribution for emitted photons |
23 | sig/sgi | Photo-atomic interaction cross section |
27 | aff | Atomic form factors or scattering functions for photo-atomic interaction |
30 | cov/intr | Covariances (correlated uncertainties) obtained from nuclear reaction model parameters |
31 | cov/nu | Covariances for nubar (average number of neutrons per fission) |
32 | cov/res | Covariances for resonance parameters |
33 | cov/sig | Covariances for reaction cross sections |
34 | cov/da | Covariances for angular distributions of emitted particles |
35 | cov/de | Covariances for energy distributions of emitted particles |
40 | cov/act | Covariances for production cross sections of radioactive nuclides |
Library
Libraries available via the retrieval interface are the major national projects and a few most important special purpose libraries. All these libraries are in the current ENDF-6 format, which elimintaes older versions such as ENDF/B-I, -II, -III, -IV and -V. Most recent versions of the following libraries are included:
# | Library | Description |
1 | ENDF/B | (U.S.) Evaluated Nuclear Data File |
2 | JENDL | Japanese Evaluated Nuclear Data Library |
3 | JEFF | (European) Joint Evaluated Fission & Fussion data library |
4 | BROND | (Russian) Library of Recommended Neutron Data |
5 | CENDL | Chinese Evaluated Neutron Data Library |
6 | ROSFOND | Russian compilation of neutron data selected from different libraries |
7 | IRDF | (IAEA) International Reactor Dosimetry File |
Sub-library (Projectile)
ENDF is organized into sub-libraries that usually represent projectiles such as n, p and g, where n=neutron, p=proton, g=gamma. There are other sub-libraries such as decay=decay data, fission yields, ... The ENDF/B-VII.1 and -VII.0 libraries contain the most extended variety of sublibraries:
# | NSUB | Sub-library name | Short name |
1 | 0 | Photonuclear | g |
2 | 3 | Photo-atomic | photo |
3 | 4 | Radioactive decay | decay |
4 | 5 | Spontaneous fission yields | s/fpy |
5 | 6 | Atomic relaxation | ard |
6 | 10 | Neutron | n |
7 | 11 | Neutron fission yields | n/fpy |
8 | 12 | Thermal scattering | tsl |
9 | 19 | Standards | std |
10 | 113 | Electro-atomic | e |
11 | 10010 | Proton | p |
12 | 10020 | Deutron | d |
13 | 10030 | Triton | t |
14 | 20030 | He3 | he3 |
MT number (Ejectile)
MT numbers provide the most detailed definition of emitted particles (ejectiles): 1=total, 2=elastic, 4=inelastic, 16=2n, 51=inelastic to 1st excited level, 102=gamma, 103=proton, 107=alpha. If neutron is a projectile, then 1 stands for (n,tot), 2=(n,n), 4=(n,n'), 16=(n,2n), 51=(n,n1'), 102=(n,gamma), 103=(n,p), 107=(n,alpha).
MT # Range | Reaction Description for Incident Neutrons |
1-100 | Cross sections for emission of neutrons. |
101-150 | Cross sections for emission of photons and charged particles. |
151-200 | Neutron resonance parameters. |
201-450 | Total production cross sections and scattering parameters. |
451-699 | Information in text, fission data, decay data, nu_bar, ... |
600-999 | Partial cross sections for emission of charged particles. |
List of neutron-induced reactions (MT#) is presented below. Neutrons are selected as default projectiles among protons and other particles.
MT # | Reaction | Reaction Description |
1 | n,tot | Neutron total cross sections |
2 | n,el | Elastic neutron scattering cross sections |
3 | n,non | Nonelastic neutron cross sections. Sum of MT=4, 5, 16-18, 22-26, 28-37,41, 42, 102-116. |
4 | n,inl | Inelastic neutron cross sections. Sum of the MT=50-91. |
5 | n,x | Cross section for the sum of all reactions not given explicitly in another MT number. |
6-9 | Not allowed in version 6 except for compatibility with JENDL-3.2 | |
10 | n,tot | Total neutron-induced continuum reactions. |
11 | n,2n+d | Production of two neutrons and a deutron. |
12-15 | Unassigned | |
16 | n,2n | Cross section for producing two neutrons and a residual. |
17 | n,3n | Cross section for producing three neutrons and a residual. |
18 | n,f | Neutron fission cross section, MT=18 is the sum of MT=19-21, 38. |
19 | n,f | First chance neutron fission cross section. |
20 | n,n+f | Second-chance neutron fission cross section. |
21 | n,2n+f | Third-chance neutron fission cross section. |
22 | n,n+a | Cross section for production of a neutron and an alpha particle. |
23 | n,n+3a | Cross section for production of a neutron and three alpha particles. |
24 | n,2n+a | Cross section for production of two neutrons and an alpha particle. |
25 | n,3n+a | Cross section for production of three neutrona and an alpha particle. |
26 | Not allowed in version 6. That was (n,2n) isomeric state cross section in version 5. | |
27 | n,abs | Neutron absorption cross section. Sum of MT=18 and MT=102-116. |
28 | n,n+p | Cross section for production of a neutron and a proton |
29 | n,n+2a | Cross section for production of a neutron and two alpha particles. |
30 | n,2n+2a | Cross section for production of two neutrons and two alpha particles. |
31 | Not allowed in version 6. | |
32 | n,n+d | Cross section for production of a neutron and a deutron. |
33 | n,n+t | Cross section for production of a neutron and a triton. |
34 | n,n+he3 | Cross section for production of a neutron and a 3He. |
35 | n,nd+2a | Cross section for production of a neutron and deutron and two alpha particles. |
36 | n,nt+2a | Cross section for production of a neutron and triton and two alpha particles. |
37 | n,4n | Cross section for production of four neutrons. |
38 | n,3n+f | Fourth chance neutron fission cross section. |
39 | Not allowed in version 6. | |
40 | Not allowed in version 6. | |
41 | n,2n+p | Cross section for production of two neutrons and a proton. |
42 | n,3n+p | Cross section for production of three neutrons and a proton. |
43 | Unassigned. | |
44 | n,n+2p | Cross section for production of a neutron and two protons. |
45 | n,n+pa | Cross section for production of neutron, proton and alpha particle. |
46-49 | Not allowed in version 6. | |
50 | n,n0 | Cross section for production of a neutron, with residual in the ground state. |
51 | n,n1 | Cross section for production of a neutron, with residual in the 1st excited state. |
52 | n,n2 | Cross section for production of a neutron, with residual in the 2nd excited state. |
53-90 | n,n' | Cross section for production of a neutron, with residual in the 3-40th excited state. |
91 | n,nc | Cross section for production of a neutron in the continnum not included in the above discrete representation. |
92-100 | Unassigned. | |
101 | n,dis | Neutron disappearance cross section. |
102 | n,g | Neutron radiative capture cross section. |
103 | n,p | Cross section for production of a proton. If MT=600-649 are present, the MT103 is their sum. |
104 | n,d | Cross section for production of a deutron. If MT=650-699 are present, the MT104 is their sum. |
105 | n,t | Cross section for production of a triton. If MT=700-749 are present, the MT105 is their sum. |
106 | n,3He | Cross section for production of 3He. If MT=750-799 are present, the MT106 is their sum. |
107 | n,a | Cross section for production of an alpha particle. If MT=800-849 are present, the MT107 is their sum. |
108 | n,2a | Cross section for production of two alpha particles. |
109 | n,3a | Cross section for production of tree alpha particles. |
110 | Unassigned. | |
111 | n,2p | Cross section for production of two protons. |
112 | n,p+a | Cross section for for production of a proton and an alpha particle. |
113 | n,t+2a | Cross section for for production of a triton and two alpha particles. |
114 | n,d+2a | Cross section for for production of a deutron and two alpha particles. |
115 | n,p+d | Cross section for for production of a proton and a deutron. |
116 | n,p+t | Cross section for for production of a proton and a triton. |
117 | n,d+a | Cross section for for production of a deutron and an alpha particle. |
118-119 | Unassigned. | |
120 | Not allowed in version 6. | |
121-150 | Unassigned. | |
151 | n,res | Neutron resonance parameters that can be used to calculate cross sections at different temperatures in the resolved and unresolved energy ranges. |
152-200 | Unassigned. | |
201 | n,xn | Total neutron production cross section. |
202 | n,xg | Total gamma production cross section. |
203 | n,xp | Total proton production cross section. |
204 | n,xd | Total deutron production cross section. |
205 | n,xt | Total triton production cross section. |
206 | n,x3He | Total 3He production cross section. |
207 | n,xa | Total alpha particle production cross section. |
208 | n,xpi+ | Total pi+ production cross section. |
209 | n,xpi0 | Total pi0 production cross section. |
210 | n,xpi- | Total pi- production cross section. |
211 | n,xmu+ | Total mu+ production cross section. |
212 | n,xmu- | Total mu- production cross section. |
213 | n,xk+ | Total k+ production cross section. |
214 | n,xk0 | Total k0 (long) production cross section. |
215 | n,xk0 | Total k0 (short) production cross section. |
216 | n,xk- | Total k- production cross section. |
217 | n,xp | Total anti-proton production cross section. |
218 | n,xn | Total anti-neutron production cross section. |
219-250 | Unassigned. | |
251 | n,... | μL (mu bar), the average cosine of scattering angle (laboratory system) for elastic scattering of neutrons. |
252 | n,... | ε, the average logarithmic energy decrement for elastic scattering of neutrons. |
253 | n,... | γ, the average of the square of the logarithmic energy decrement for elastic scattering divided by twice the average logarithmic energy decrement for elastic scattering of neutrons. |
254-300 | Unassigned. | |
301-450 | n,... | Energy release parameters, E, &sigma, for total and partial cross sections. MT=300 plus reaction MT number. |
451 | n,... | Heading or title information; given in File 1 only. |
452 | n,nu_tot | Neutron fission: Nubar (νT), average total (prompt plus delayed) number of neutrons released per fission |
452 | 0,nu_tot | Spontaneous fission: Nubar (νT), average total (prompt plus delayed) number of neutrons released per fission |
453 | Unassigned. | |
454 | n,ind_fy | Neutron fission: Independent fission product yield data |
454 | 0,ind_fy | Spontaneous fission: Independent fission product yield data |
455 | n,nu_d | Neutron fission: Average number of delayed neutrons released per fission event |
456 | n,nu_p | Neutron fission: Average number of prompt neutrons released per fission event |
457 | decay | Radioactive decay data |
458 | e_rel_fis | Neutron fission: Energy release in fission for incident neutrons |
459 | fy_cum | Neutron/spontaneous fission: Cumulative fission product yield data |
460-464 | Unassigned. | |
465-466 | Not allowed in version 6. | |
467-499 | Unassigned. | |
500 | Total charge particle stopping power. | |
501 | Total photon interaction. | |
502 | Photon coherent scattering. | |
503 | Unassigned. | |
504 | Photon incoherent scattering. | |
505 | Imaginary scattering factor. | |
506 | Real scattering factor. | |
507-514 | Unassigned. | |
515 | Pair production, electron field. | |
516 | Pair production, Sum of MT=515,517. | |
517 | Pair production, nuclear field. | |
518 | Not allowed in version 6. | |
519-521 | Unassigned. | |
522 | Photoelectric absorption. | |
523 | Photo-excitation cross section. | |
524-525 | Unassigned. | |
526 | Electro-atomic scattering. | |
527 | Electro-atomic bremsstrahlung. | |
528 | Electro-atomic excitation cross section. | |
529-531 | Unassigned. | |
532 | Not allowed in version 6. | |
533 | Atomic relaxation data. | |
534 | K | (1s1/2) Subshell photoelectric cross section. |
535 | L1 | (2s1/2) Subshell photoelectric cross section. |
536 | L2 | (2p1/2) Subshell photoelectric cross section. |
537 | L3 | (2p3/2) Subshell photoelectric cross section. |
538 | M1 | (3s1/2) Subshell photoelectric cross section. |
539 | M2 | (3p1/2) Subshell photoelectric cross section. |
540 | M3 | (3p3/2) Subshell photoelectric cross section. |
541 | M4 | (3d3/2) Subshell photoelectric cross section. |
542 | M5 | (3d5/2) Subshell photoelectric cross section. |
543 | N1 | (4s1/2) Subshell photoelectric cross section. |
544 | N2 | (4p1/2) Subshell photoelectric cross section. |
545 | N3 | (4p3/2) Subshell photoelectric cross section. |
546 | N4 | (4d3/2) Subshell photoelectric cross section. |
547 | N5 | (4d5/2) Subshell photoelectric cross section. |
548 | N6 | (4f5/2) Subshell photoelectric cross section. |
549 | N7 | (4f7/2) Subshell photoelectric cross section. |
550 | O1 | (5s1/2) Subshell photoelectric cross section. |
551 | O2 | (5p1/2) Subshell photoelectric cross section. |
552 | O3 | (5p3/2) Subshell photoelectric cross section. |
553 | O4 | (5d3/2) Subshell photoelectric cross section. |
554 | O5 | (5d5/2) Subshell photoelectric cross section. |
555 | O6 | (5f5/2) Subshell photoelectric cross section. |
556 | O7 | (5f7/2) Subshell photoelectric cross section. |
557 | O8 | (5g7/2) Subshell photoelectric cross section. |
558 | O9 | (5g9/2) Subshell photoelectric cross section. |
559 | P1 | (6s1/2) Subshell photoelectric cross section. |
560 | P2 | (6p1/2) Subshell photoelectric cross section. |
561 | P3 | (6p3/2) Subshell photoelectric cross section. |
562 | P4 | (6d3/2) Subshell photoelectric cross section. |
563 | P5 | (6d5/2) Subshell photoelectric cross section. |
564 | P6 | (6f5/2) Subshell photoelectric cross section. |
565 | P7 | (6f7/2) Subshell photoelectric cross section. |
566 | P8 | (6g7/2) Subshell photoelectric cross section. |
567 | P9 | (6g9/2) Subshell photoelectric cross section. |
568 | P10 | (6h9/2) Subshell photoelectric cross section. |
569 | P11 | (6h11/2) Subshell photoelectric cross section. |
570 | Q1 | (7s1/2) Subshell photoelectric cross section. |
571 | Q2 | (7p1/2) Subshell photoelectric cross section. |
572 | Q3 | (7p3/2) Subshell photoelectric cross section. |
573-599 | Unassigned. | |
600 | n,p0 | Cross section for production of a proton leaving the residual nucleus in the ground state. |
601 | n,p1 | Cross section for production of a proton leaving the residual nucleus in the 1st excited state. |
602 | n,p2 | Cross section for production of a proton leaving the residual nucleus in the 2nd excited state. |
603-648 | n,p# | Cross section for production of a proton leaving the residual nucleus in the #th excited state. |
649 | n,pc | Cross section for production of a proton in the continuum. |
650 | n,d0 | Cross section for production of a deutron leaving the residual nucleus in the ground state. |
651 | n,d1 | Cross section for production of a deutron leaving the residual nucleus in the 1st excited state. |
652 | n,d2 | Cross section for production of a deutron leaving the residual nucleus in the 2nd excited state. |
653-698 | n,d# | Cross section for production of a deutron leaving the residual nucleus in the #th excited state. |
699 | n,dc | Cross section for production of a deutron in the continuum. |
700 | n,t0 | Cross section for production of a triton leaving the residual nucleus in the ground state. |
701 | n,t1 | Cross section for production of a triton leaving the residual nucleus in the 1st excited state. |
702 | n,t2 | Cross section for production of a triton leaving the residual nucleus in the 2nd excited state. |
703-748 | n,t# | Cross section for production of a triton leaving the residual nucleus in the #th excited state. |
749 | n,tc | Cross section for production of a triton in the continuum. |
750 | n,3He0 | Cross section for production of a 3He particle leaving the residual nucleus in the ground state. |
751 | n,3He1 | Cross section for production of a 3He particle leaving the residual nucleus in the 1st excited state. |
752 | n,3He2 | Cross section for production of a 3He particle leaving the residual nucleus in the 2nd excited state. |
753-798 | n,3He# | Cross section for production of a 3He particle leaving the residual nucleus in the #th excited state. |
799 | n,3Hec | Cross section for production of a 3He particle in the continuum. |
800 | n,a0 | Cross section for production of an alpha particle leaving the residual nucleus in the ground state. |
801 | n,a1 | Cross section for production of an alpha particle leaving the residual nucleus in the 1st excited state. |
802 | n,a2 | Cross section for production of an alpha particle leaving the residual nucleus in the 2nd excited state. |
803-848 | n,a# | Cross section for production of an alpha particle leaving the residual nucleus in the #th excited state. |
849 | n,ac | Cross section for production of an alpha particle in the continuum. |
850 | Unassigned. | |
851-870 | Lumped reaction covariances. | |
871-874 | Unassigned. | |
875 | n,2n0 | Cross section for production of two neutrons leaving the residual nucleus in the ground state. |
876 | n,2n1 | Cross section for production of two neutrons leaving the residual nucleus in the 1st excited state. |
877 | n,2n2 | Cross section for production of two neutrons leaving the residual nucleus in the 2nd excited state. |
878-890 | n,2n# | Cross section for production of two neutrons leaving the residual nucleus in the #th excited state. |
891 | n,2nc | Cross section for production of two neutrons in the continuum. |
892-999 | Unassigned. |
MF number (Quantity)
MF numbers define quantity, where 1=comment part, 2=neutron resonance parameters, 3=reaction cross section, 4=angular distribution of emitted particles, 5=spectra of emitted particles, 12=photon production, 32=covariances in neutron resonance region, 33=covariances above neutron resonance region.
MF # | Quantity | Quantity Description |
1 | info | General information |
2 | res | Resonance parameters |
3 | sig | Reaction cross section |
4 | da | Angular distribution for emitted particles |
5 | de | Energy distribution for emitted particles |
6 | da/de | Energy-Angle distribution for emitted particles |
7 | sig/ths | Thermal neutron scattering |
8 | rnp | Radioactivity and fission-product yields |
9 | mrnp | Multiplicities for production of radioactive nuclides |
10 | sig/act | Cross section for production of radioactive nuclides |
11 | commg | General comments on photon production |
12 | mult | Multiplicities for photon production |
13 | sigg | Cross sections for photon production |
14 | dag | Angular distribution for emitted photons |
15 | deg | Energy distribution for emitted photons |
23 | sig/sgi | Photo-atomic interaction cross section |
27 | aff | Atomic form factors or scattering functions for photo-atomic interaction |
30 | cov/intr | Covariances (correlated uncertainties) obtained from nuclear reaction model parameters |
31 | cov/nu | Covariances for nubar (average number of neutrons per fission) |
32 | cov/res | Covariances for resonance parameters |
33 | cov/sig | Covariances for reaction cross sections |
34 | cov/da | Covariances for angular distributions of emitted particles |
35 | cov/de | Covariances for energy distributions of emitted particles |
40 | cov/act | Covariances for production cross sections of radioactive nuclides |
Product
Final nucleus produced in a reaction. It is rarely used, though it may be userful when looking for activation cross section producing radioactive nucleus in JEFF-3.1/A (Activation) library.
Laboratory
Laboratories where evaluation was done are given in the comment part of each file (target).
Author(s)
Authors of evaluations are given in the comment part of each file. For example, to retrieve all evaluations by P.G. Young one should ask for *Young*.
Examples
Several examples follow in a graphical form.
Note: some retrieval and plotting capabilities may be limited, including cross-section plots at higher energies (> 20 MeV); examples were prepared using previous versions of the retrieval system, some changes have been introduced since then.
-
Provide input and select library: Target=Fe-56, Reaction=n,g, Quantity=sig and Library=ENDF/B-VII.0;JENDL-3.3
-
Click on the "Plot" button to display data
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Use "Repaint" button to change plot options
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Use "Repaint" button to change plot options again if necessary
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Provide input and select library: Target=99Tc, Reaction=n,inl;n,2n, Quantity=sig and Library=JENDL-3.3
-
Select (n,inl) reaction and click on the "Interpreted" button to display it
-
View numerical data, go back and click on the "Plot" button to plot it
-
Click on the "plotted data" link to see numerical data for (n,inl) reaction
-
Use this data as input for data analysis programs
-
Provide input and select library: Target=Gd*, Reaction=n,tot, Quantity=sig and Library=ENDF/B-VII.0
-
Click on the "Plot" button to display data
-
Click on the "plotted data" link to see numerical data and "Repaint" button to change plot options
- Provide input and select library: Target=11-na-23, Reaction=n,el, Quantity=sig and Library=ENDF/B-VII.0
- Click on the "Plot" button to display data
- Click on the "EXFOR" button to find and add to the plot experimental data from EXFOR database
- Click on the "Retrieve" button to retrieve EXFOR data
- Set X and Y plot parameters and use the "Repaint" button or Zoom option to re-draw the plot
- Click on the "plotted data" link to access numerical data
- Use this data as input for data analysis programs
-
Provide input and select library: Target=56Fe, Reaction=n,el, Quantity=da and Library=ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ENDF/B-VI.8
-
Select the data and click on the "Plot" button to extract ENDF d&sigma/dΩ(θ) data
-
Select the data and click on the "d&sigma/dΩ(θ)" button
-
Select data for incident netron energies and click on the "Plot Selected" button
-
Access ENDF numerical data plots
-
Access ENDF numerical data by clicking on the "plotted data" link
- Provide input and select library: Target=u-238, Reaction=n,n`;n,2n, Quantity=de and Library=ENDF/B-VI.8;JENDL-3.3.
Warning: One should type (n,n`) with backward prime to retrieve continuum part of inelastic neutron spectrum.
To get full spectrum, one should start with continuum spectrum and add transitions to discrete levels - this capability is not yet available.
-
Select inelastic data for uranium and click on the "Plot" button to extract it
-
Select data and click on the "dσ/dE" button
-
Select for incident neutron energies data for uranium and click on the "Plot Selected" button to extract it
-
View plotted data
-
Provide input and select library: Target=235-u, Reaction=n,nu* and Library=ENDF/B-VII.0; JEFF-3.1; JENDL-3.3; ENDF/B-VI.8
-
Select the data and click on the "Interpreted" button for the total fission nubar
-
View data for the total fission nubar, go back and click on the "Interpreted" button for the delayed fission nubar
-
Compare numerical data for delayed and total fission nubars
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select library: Target=239Pu, Reaction=457 and Library=All
-
Select the data and click on the "Interpreted" button to display it
-
View numerical radioactive decay data for 239Pu, ignore word "unknown" in the data file
-
Select "Extended Retrieval" on the ENDF frontpage
-
Provide input and select library basic version: Target=u-235, Sub-library (Projectile) = *fpy* and Library=ENDF/B-VII.0
-
Select the data and click on the "ENDF-6" button to display it
-
View numerical fission product yields data for 235U
-
Provide input and select library: Reaction=n,g, Quantity=sig and Library=ENDF/B-VII.0
-
Select target materials of interest and click on the "Retrieve" button to access it
-
Display links to text and zipped data for targets of interest, click on the "ZIP" link to download data
-
Click on the "Save" link to download data
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select basic version libraries in advanced request view: Target=*, MF# (Quantity)=32, Library=All and set display options (Retrieve: Sections, Display: Evaluations, View: Details, Sort by: Reactions)
-
View the list of covariance data
-
Select "Extended Retrieval" on the ENDF frontpage
-
Provide input and select basic version library: Target=*, MF# (Quantity)=7 and Library=JEFF-3.1
-
Select the data and click on the "Interpreted" button to display it
-
Analyze numerical data for thermal neutron scattering
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select library: Target=63Cu, Projectile=n, Basic Library=ENDF/B-VII.0 and Options (Retrieve:Sections, Display: Evaluations, View: List, Sort by: Evaluations)
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Select the data and click on the "MAT" button to retrieve full file (material) in original ENDF-6 format
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View numerical data for 63Cu, use "Save As" option to save it
How to compare evaluated and experimental data
- Provide input and select library: Target=11-na-23, Reaction=n,el, Quantity=sig and Library=ENDF/B-VII.0
- Click on the "Plot" button to display data
- Click on the "EXFOR" button to find and add to the plot experimental data from EXFOR database
- Click on the "Retrieve" button to retrieve EXFOR data
- Set X and Y plot parameters and use the "Repaint" button to re-draw the plot
- Experimental data are shown in green, click on the "plotted data" link to access numerical data
- Use this data as input for data analysis programs
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select library: Target=63Cu, Projectile=n, Basic Library=ENDF/B-VII.0 and Options (Retrieve:Sections, Display: Evaluations, View: List, Sort by: Evaluations)
-
Select the data and click on the "MAT" button to retrieve full file (material) in original ENDF-6 format
-
View numerical data for 63Cu, use "Save As" option to save it
-
Provide input and select library: Target=56Fe, Reaction=n,el, Quantity=da and Library=ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ENDF/B-VI.8
-
Select the data and click on the "Plot" button to extract ENDF d&sigma/dΩ(θ) data
-
Select the data and click on the "d&sigma/dΩ(θ)" button
-
Select data for incident netron energies and click on the "Plot Selected" button
-
Access ENDF numerical data plots
-
Provide input and select library: Target=u-238, Reaction=n,n`;n,2n, Quantity=de and Library=ENDF/B-VI.8;JENDL-3.3.
Warning: One should type (n,n`) with backward prime to retrieve continuum part of inelastic neutron spectrum.
To get full spectrum, one should start with continuum spectrum and add transitions to discrete levels - this capability is not yet available.
-
Select inelastic data for uranium and click on the "Plot" button to extract it
-
Select data and click on the "dσ/dE" button
-
Select for incident neutron energies data for uranium and click on the "Plot Selected" button to extract it
-
View plotted data
-
Provide input and select library: Target=Fe-56, Reaction=n,g, Quantity=sig and Library=ENDF/B-VII.0;JENDL-3.3
-
Click on the "Plot" button to display data
-
Use "Repaint" button to change plot options
-
Use "Repaint" button to change plot options again if necessary
-
Provide input and select library: Target=Gd*, Reaction=n,tot, Quantity=sig and Library=ENDF/B-VI.8
-
Click on the "Plot" button to display data
-
Click on the "plotted data" link to see numerical data and "Repaint" button or Zoom to change plot options
-
Provide input and select library: Target=56Fe, Reaction=n,el, Quantity=da and Library=ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ENDF/B-VI.8
-
Select the data and click on the "Plot" button to extract ENDF d&sigma/dΩ(θ) data
-
Select the data and click on the "d&sigma/dΩ(θ)" button
-
Select data for incident netron energies and click on the "Plot Selected" button
-
Access ENDF numerical data plots
-
Access ENDF numerical data by clicking on the "plotted data" link
-
Provide input and select library: Target=235-u, Reaction=n,nu* and Library=ENDF/B-VII.0; JEFF-3.1; JENDL-3.3; ENDF/B-VI.8
-
Select the data and click on the "Interpreted" button for the total fission nubar
-
View data for the total fission nubar, go back and click on the "Interpreted" button for the delayed fission nubar
-
Compare numerical data for delayed and total fission nubars
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select basic version libraries in advanced request view: Target=*, MF# (Quantity)=32, Library=All and set display options (Retrieve: Sections, Display: Evaluations, View: Details, Sort by: Reactions)
-
View the list of covariance data
-
Provide input and select library: Target=Fe-56, Reaction=n,g, Quantity=sig and Library=ENDF/B-VII.0;JENDL-3.3
-
Click on the "Plot" button to display data
-
Use "plotted data" link to access plotted data
-
Select "Advanced Retrieval" option on the ENDF frontpage
-
Provide input and select library: Target=239Pu, Reaction=457 and Library=All
-
Select the data and click on the "Interpreted" button to display it
-
View numerical radioactive decay data for 239Pu, ignore word "unknown" in the data file
-
Provide input and select library: Target=u-238, Reaction=n,n`;n,2n, Quantity=de and Library=ENDF/B-VI.8;JENDL-3.3.
Warning: One should type (n,n`) with backward prime to retrieve continuum part of inelastic neutron spectrum.
To get full spectrum, one should start with continuum spectrum and add transitions to discrete levels - this capability is not yet available.
-
Select inelastic data for uranium and click on the "Plot" button to extract it
-
Select data and click on the "dσ/dE" button
-
Select for incident neutron energies data for uranium and click on the "Plot Selected" button to extract it
-
View plotted data
-
Select "Extended Retrieval" on the ENDF frontpage
-
Provide input and select library basic version: Target=u-235, Sub-library (Projectile) = *fpy* and Library=ENDF/B-VII.0
-
Select the data and click on the "Retrieve" button to display it
-
View numerical fission product yields data for 235U
-
Provide input and select library: Target=99Tc, Reaction=n,inl;n,2n, Quantity=sig and Library=JENDL-3.3
-
Select (n,inl) reaction and click on the "Interpreted" button to display it
-
View numerical data, go back and click on the "Plot" button to plot it
-
Click on the "plotted data" link to see numerical data for (n,inl) reaction
-
Use this data as input for data analysis programs
-
Select "Extended Retrieval" on the ENDF frontpage
-
Provide input and select basic version library: Target=*, MF# (Quantity)=7 and Library=JEFF-3.1
-
Select the data and click on the "Interpreted" button to display it
-
Analyze numerical data for thermal neutron scattering
-
Provide input and select library: Reaction=n,g, Quantity=sig and Library=ENDF/B-VII.0
-
Select target materials of interest and click on the "Retrieve" button to access it
-
Display links to text and zipped data for targets of interest, click on the "ZIP" link to download data
-
Click on the "Save" link to download data
-
What is Library 300° K pointwise?
It refers to conversion of resonance parameters into cross sections and to Doppler broadening them by taking into account neutron velocities at room temperature. Applies only to the neutron sublibrary in the resonance region. This pointwise version of the library is used for plotting. -
What is Library basic version?
It is the library in the original ENDF-6 format, usually with the resonance region given in terms of the resonance parameters. This basic version of the library must be processed by codes such as NJOY to provide suitable input for application codes. -
What is ENDF-6?
ENDF-6 is format developed for ENDF/B-VI library. The format is internationally adopted. New ENDF/B-VII library is also stored in this format. -
What is interpreted ENDF?
Data stored in the ENDF-6 format are difficult to understand. The interpretted ENDF, developed by R. MacFarlane in Los Alamos, translates ENDF-6 data into human readable format. -
What is MAT?
It stands for material (isotope, element). In the ENDF-6 format each material is assigned a unique MAT number. For example, Fe-56 has MAT= 2631. -
What is nu-bar?
It is the average number of neutrons per fission.
-
Why can't I plot low-energy neutron capture cross sections?
Neutron reactions at low energies (usually below 100 keV) are stored in form of resonance parameters (MF=2). These parameters must be first converted into cross sections by processing that includes Doppler broadening effect usually at 3000 Kelvin. If you retrieved data from the basic library where no processing was applied, cross sections may not be available and thus cannot be plotted. -
Why can't I retrieve neutron inelastic cross section
Inelastic neutrons are typically stored under several MT numbers (MT=51,52, ... for transitions to discrete levels, and MT=91 for the continuum part). These partials might be summed up and stored as total inelastic neutron cross section under MT=4 that can be retrieved and plotted as Reaction=(n,inl), Quantity=sig. MT=4, however, is redundant and not mandatory in a basic evaluation. If it is not present, in order to plot inelastic scattering one should sum continuum cross section with cross sections to all discrete levels (this capability is not yet available)
To retrieve spectrum (Quantity=de) of inelastic neutrons, one should use Reaction=(n,n`) with backward prime to get continuum spectrum. -
Why can't I retrieve neutron cross-sections for a number of nuclei?
ENDF libraries contain data for nuclei of technological interest, representing a small fraction (~10%) of all known nuclei.