A-6.
Average number of neutrons emitted per fission
 
 
     Nuclide       		  
  Type		             Total-neutron Yield             Delayed-neutron Yield
 
   νt		  
  Source
 		  
   νd		  
  Source 
 
  90-Th-232
  fast
  2.456 ± 0.018
  ENDF/B-VII
  0.0499 ± 0.0019
  ENDF/B-VII(1)
  92-U-233
  thermal
  2.4968 ± 0.0035
  IAEA-CRP-STD
  0.0067 ± 0.0003
  JEFF-3.1(2)
  92-U-235
  thermal
  2.4355 ± 0.0023
  IAEA-CRP-STD
  0.0162 ± 0.0005
  JEFF-3.1(3)
  92-U-238
  fast
  2.819 ± 0.020
  ENDF/B-VII(4)
  0.0465 ± 0.0024
  JEFF-3.1(3)
  94-Pu-238
  fast
  3.00 ± 0.14
  JEFF-3.1(2)
  0.0047 ± 0.0005
  JEFF-3.1(1)
  94-Pu-239
  thermal
  2.8836 ± 0.0047
  IAEA-CRP-STD
  0.0065 ± 0.0003
  JEFF-3.1(3)
  94-Pu-240
  fast
  3.086 ± 0.025
  JEFF-3.1(2)
  0.0090 ± 0.0004
  JEFF-3.1(1)
  94-Pu-241
  thermal
  2.9479 ± 0.0055
  IAEA-CRP-STD
  0.0160 ± 0.0008
  JEFF-3.1(2)
  94-Pu-242
  fast
  3.189 ± 0.035
  JEFF-3.1(2)
  0.0183 ± 0.0010
  JEFF-3.1(1)
  95-Am-241
  thermal
  3.239 ± 0.024
  JEFF-3.1(2)
  0.0043 ± 0.0006
  JEFF-3.1(2)
  96-Cm-242
  spontaneous
  2.529 ± 0.017
  JEFF-3.1(5)
  0.0013 ± 0.0003
  Mills(1995)
  96-Cm-243
  thermal
  3.433 ± 0.047
  JEFF-3.1(2)
  0.0030 ± 0.0003
  JEFF-3.1(2)
  96-Cm-244
  spontaneous
  2.691 ± 0.012
  JEFF-3.1(5)
  0.0033 ± 0.0010
  Mills(1995)
  96-Cm-245
  thermal
  3.60 ± 0.13
  JEFF-3.1(2)
  0.0064 ± 0.0014
  JEFF-3.1(2)
  98-Cf-252
  spontaneous
  3.7692 ± 0.0047
  IAEA-CRP-STD
  0.0086 ± 0.0010
  Tuttle(1979)
fast = fast spectrum, thermal = thermal spectrum, spontaneous = spontaneous fission.
 
References
 
 
 IAEA-CRP-STD
S. A. Badikov, C. Zhenpeng, A. D. Carlson, E. V. Gai, G. M. Hale, F.-J. Hambsh, H. M. Hofmann, T. Kawano, N. M. Larson, V. G. Pronyaev, D.L. Smith, Soo-Youl Ho, S. Tagesen, H. K. Vonach, H.L. Nichols, IAEA CRP "International Evaluation of Neutron Cross-Section Standards", IAEA Scientific and Technical Report STI/PUB/1292, November 2007, International Atomic Energy Agency, Vienna, Austria, ISBN 90-0-100807-4.
http://www-nds.iaea.org/standards/.

 ENDF/B-VII
U.S. Evaluated Nuclear Data Library ENDF/B-VII β1, http://www.nndc.bnl.gov/exfor/endf00.htm,
3 October 2006;
see also M. B. Chadwick et al., ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology, Nucl. Data Sheets 107 (2006) 2931.

  JEFF-3.1
Joint Evaluated Fission and Fusion File, Incident-neutron data,
http://www-nds.iaea.org/exfor/endf00.htm, 26 February 2006;
see also A. Koning, R. Forrest, M. Kellett, R. Mills, H. Henriksson, Y. Rugama, The JEFF-3.1 Nuclear Data Library, JEFF Report 21, OECD/NEA, Paris, France, 2006, ISBN 92-64-02314-3.

  NEA/WPEC-6
G. Rudstam, Ph. Finck, A. Filip, A. D'Angelo, R.D. McKnight, Delayed Neutron Data for the Major Actinides, Volume 6, NEA/WPEC-6, NEA/OECD, Paris, France, 2002.

  EXFOR
Experimental Nuclear Reaction Data
http://www-nds.iaea.org/exfor/exfor00.htm, 27 March 2006.

  P&I(1998)
V. M. Piksaikin, S. G. Isaev, Correlation properties of delayed neutrons from fast neutron induced fission, pp. 1-13 in INDC(CCP)-415, October 1998, IAEA, Vienna, Austria.

  Mills(1995)
R. W. Mills, Fission product yield evaluation, PhD thesis, University of Birmingham, UK, March 1995.

 Tuttle(1979)
R. J. Tuttle, Delayed-neutron yields in nuclear fission, pp. 29-67 in Proc. Consultants' Meeting on Delayed Neutron Properties, 26-30 March 1979, INDC(NDS)-107 (1979) 29, IAEA, Vienna, Austria.
 
Notes
 
(1) Uncertainties estimated from selected experimental data reported by P&I (1998).

(2) Uncertainties estimated from selected experimental data available in EXFOR.

(3) Delayed-neutron data adopted from NEA/WPEC-6.

(4) Prompt-neutron yield adopted from ENDF/B-VII β3; uncertainty in prompt-neutron yield estimated from the
U-238 covariance files included in the ENDF/B-VII β1 library (modification flag 5E for material 9237); total neutron yield calculated as the sum of prompt- and delayed-neutron yields.

(5) Prompt-neutron yield adopted from the JEFF-3.1 radioactive decay data library; uncertainty in prompt-neutron yield estimated from selected experimental data available in EXFOR; total spontaneous neutron yield calculated as the sum of prompt- and delayed-neutron yields.

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