Cross Section (barns)

235U(n,tot)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
235-U(n,tot)

102

235U(n,tot)

1010-3

10-2

10-1

1

10Incident Neutron Energy (MeV)

Figure 1: 235 U total cross section

Cross Section (barns)

235U(n,el)

235U(n,el)

10

ENDF-VII.1
JEFF-3.1
JENDL-4.0
235-U(n,el)

10-2

10-1

1

Incident Neutron Energy (MeV)

Figure 2: 235 U elastic scattering cross section

1

10

235U(n,f)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
235-U(n,f)

235U(n,f)

101
10-2

10-1

110Incident Neutron Energy (MeV)

Figure 3: 235 U fission cross section
235U(n,f)7235U(n,f)

Cross Section (barns)

65

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
235-U(n,f)

43
2

10-2

10-1

Incident Neutron Energy (MeV)

Figure 4: 235 U fission cross section

2

235U(n,f)

Cross Section (barns)

2.5

235U(n,f)
2.0

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
235-U(n,f)

1.5

1.0
1

10Incident Neutron Energy (MeV)

Figure 5: 235 U fission cross section

3235U(n,g)

Cross Section (barns)

1235U(n,g)

10-1

10-2

ENDF-VII.1 ==> JEFF-3.1, ROSFONDJENDL-4.0
235-U(n,g)

10-3

10-2

10-1

110Incident Neutron Energy (MeV)

Figure 6: 235 U radiative capture cross section
235U(n,g)

Cross Section (barns)

235U(n,g)
10
ENDF-VII.1 ==> JEFF-3.1, ROSFONDJENDL-4.0
235-U(n,g)

1

10-3

10-2

Incident Neutron Energy (MeV)

Figure 7: 235 U radiative capture cross section

410-1

235U(n,2n)

235U(n,2n)

Cross Section (barns)

0.8

0.6

0.4
ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
235-U(n,2n)

0.205101520Incident Neutron Energy (MeV)

Figure 8: 235 U (n,2n) reaction cross section
235U(n,3n)

Cross Section (barns)

0.5235U(n,3n)

0.4

0.3
ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
235-U(n,3n)

0.2

0.1

012

141618

Incident Neutron Energy (MeV)

Figure 9: 235 U (n,3n) reaction cross section

5235U(n,inl)

Cross Section (barns)

2.5235U(n,inl)
2.01.5ENDF-VII.1
JEFF-3.1 ==> ROSFONDJENDL-4.0
235-U(n,inl)

1.00.55101520

Incident Neutron Energy (MeV)

Figure 10: 235 U total inelastic scattering cross section
235U(n,n1)

1

Cross Section (barns)

ENDF-VII.1 ==> JEFF-3.1, ROSFONDJENDL-4.0

235U(n,n1)

10-2

10-4

10-6

10-8
10-4

10-3

10-2

10-1

1Incident Neutron Energy (MeV)

Figure 11: 235 U inelastic to the first excited state610

235U(n,n2)

235U(n,n2)

Cross Section (barns)

10-2

10-3

10-4

ENDF-VII.1 ==> JEFF-3.1, ROSFONDJENDL-4.0

10-5

10-6
246810Incident Neutron Energy (MeV)

Figure 12: 235 U inelastic to the second excited state
235U(n,n3)

Cross Section (barns)

0.3

ENDF-VII.1 ==> JEFF-3.1, ROSFONDJENDL-4.0

235U(n,n3)

0.2

0.1

10-1

1Incident Neutron Energy (MeV)

Figure 13: 235 U inelastic to the third excited state710

235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc = 14.7 MeV

235U(n,f) PFNS E

inc = 14.7 MeV

1ENDF-VII.1
JENDL-4.0235-U(n,f)

10-4

10-3

10-210-1

110

Outgoing Neutron Energy (MeV)

Figure 14: 235 U prompt fission neutron spectrum
235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc=15.0 MeV

ENDF-VII.1
JENDL-4.0
235-U(n,f)

1235U(n,f) PFNS E
10-2

10-1

inc=15.0 MeV
1Outgoing Neutron Energy (MeV)

Figure 15: 235 U prompt fission neutron spectrum

810

235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc=7.5 MeV

235U(n,f) PFNS E =7.5 MeV
inc
ENDF-VII.1
JENDL-4.0
235-U(n,f)

110-410-310-210-1

110Outgoing Neutron Energy (MeV)

Figure 16: 235 U prompt fission neutron spectrum
235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc=5.5 MeV

235U(n,f) PFNS E

inc=5.5 MeV

ENDF-VII.1
JENDL-4.0
235-U(n,f)

110-3

10-2

10-1

1

Outgoing Neutron Energy (MeV)

Figure 17: 235 U prompt fission neutron spectrum

910

235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc=0.5 MeV

1ENDF-VII.1
JENDL-4.0
235-U(n,f)

235U(n,f) PFNS E =0.5 MeV
inc10-110-1110Outgoing Neutron Energy (MeV)

Figure 18: 235 U prompt fission neutron spectrum
235U(n,f) PFNS E

inc=34 keV

Ratio to Maxwellian T=1.382 MeV

1.5
ENDF-VII.1
JENDL-4.0
235-U(n,f)

1.0

0.5

235U(n,f) PFNS E

inc=34 keV

10-210-1

1

Outgoing Neutron Energy (MeV)

Figure 19: 235 U prompt fission neutron spectrum

1010

235U(n,f) PFNS E

Ratio to Maxwellian T=1.382 MeV

inc=0.0253 eV

ENDF-VII.1
JENDL-4.0
235-U(n,f)

1235U(n,f) PFNS E =0.0253 eV
inc
10-1
10-2

10-1

110Outgoing Neutron Energy (MeV)

Figure 20: 235 U prompt fission neutron spectrum
235U µ-bar

1.0

235U µ-bar

mu-bar (barns)

0.8
ENDF-VII.1
JENDL-4.0235-U(MT=251)

0.6

0.4

0.2

10-3

10-2

10-1

1

Incident Neutron Energy (MeV)

Figure 21: 235 U average scattering cosine

1110

235U prompt ν-bar

Prompt nu-bar

5

ENDF-VII.1
JENDL-4.0
235-U(MT=456)

4235U prompt ν-bar
3510152025

Energy (MeV)

Figure 22: 235 U average number of prompt fission neutrons
235U total ν-bar

Total nubarENDF-VII.1
JENDL-4.0
Experiment

235U total ν-bar

510152025

Incident Neutron Energy (MeV)

Figure 23: 235 U average number of total fission neutrons1230

238U(n,tot)

Cross Section (barns)

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,tot)

238U(n,tot)

1010-1

1

10Incident Energy (MeV)

Figure 24: 238 U total cross section
238U(n,el)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
238-U()

238U(n,el)

1010-1

1Incident Energy (MeV)

Figure 25: 238 U elastic scattering cross section

1310

238U(n,f)

10Cross Section (barns)

1

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

238U(n,f)

10-2
10-3
10-4
1

10

Incident Energy (MeV)

Figure 26: 238 U fission cross section
238U(n,f)

Cross Section (barns)

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

10-2

10-3

238U(n,f)10-4
1Incident Energy (MeV)

Figure 27: 238 U fission cross section

14

238U(n,f)

Cross Section (barns)

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,f)

1238U(n,f)

10Incident Energy (MeV)

Figure 28: 238 U fission cross section
238U(n,f)

Cross Section (barns)

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,f)

238U(n,f)

1.5

1.0

0.5

5

10Incident Energy (MeV)

Figure 29: 238 U fission cross section

1515

238U(n,γ)

Cross Section (barns)

1238U(n,γ)
10-1

10-2

10-3

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,g)

10-11Incident Energy (MeV)

Figure 30: 238 U radiative capture cross section

1610

238U(n,2n)

Cross Section (barns)

1.5

238U(n,2n)

1.0

0.506

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,2n)

81012141618

Incident Energy (MeV)

Figure 31: 238 U (n,2n) reaction cross section
238U(n,3n)

1.0

Cross Section (barns)

ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0
238-U(n,3n)

0.8

0.6

238U(n,3n)

0.4

0.2

10121416

Incident Energy (MeV)

Figure 32: 238 U (n,3n) reaction cross section1718

238U(n,inl)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND
238-U(n,inl)

32

238U(n,inl)

1010-1110Incident Energy (MeV)

Figure 33: 238 U total inelastic scattering cross section
238U(n,n )1

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
ROSFOND

238U(n,n )
1

110-1

10-1

1Incident Energy (MeV)

Figure 34: 238 U inelastic to the first excited state

1810

238U(n,n )2

Cross Section (barns)

0.6

ENDF-VII.1 ==> ROSFOND
JEFF-3.1JENDL-4.0

238U(n,n )
20.5
0.4
0.3
0.2
0.1

110Incident Energy (MeV)

Figure 35: 238 U inelastic to the second excited state
238U(n,n )30.06
ENDF-VII.1 ==> ROSFOND
JEFF-3.1
JENDL-4.0

238U(n,n )3Cross Section (barns)

0.05
0.04
0.03
0.02
0.01

110Incident Energy (MeV)

Figure 36: 238 U inelastic to the third excited state

19

238U µ-bar

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(MT=251)

mu-bar110-1

238U µ-bar
10-2

10-3

10-210-1

110Incident Energy (MeV)

Figure 37: 238 U average scattering cosine
238U ν-bar

6.0

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(MT=452)

238U ν-barTotal nubar

5.04.03.051015Incident Energy (MeV)

Figure 38: 238 U average number of prompt fission neutrons2020Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =1.32 MeVn

238U PFNS E =1.32 MeV
n103

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

110Outgoing Neutron Energy (MeV)

Figure 39: 238 U prompt fission neutron spectrum

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =14.3 MeVn238U PFNS E =14.3 MeV
n1.5

ENDF-VII.1 Ei1.43E+7JEFF-3.1 Ei1.43E+7JENDL-4.0 Ei1.43E+7238-U(n,f) Ei1.43E+71.010-1

1Outgoing Neutron Energy (MeV)

Figure 40: 238 U prompt fission neutron spectrum

2110Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =16 MeV
n2.5

238U PFNS E =16 MeV
n2.0

ENDF-VII.1
JEFF-3.1
JENDL-4.0238-U(n,f)

1.5

1.00.510-1

110Outgoing Neutron Energy (MeV)Figure 41: 238 U prompt fission neutron spectrum

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =17.7 MeVn238U PFNS E =17.7 MeVn
ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

110-11Outgoing Neutron Energy (MeV)

Figure 42: 238 U prompt fission neutron spectrum

2210

Ratio to Maxwellian T=1.328 MeV

238U PFNS E =2.9 MeV
n1.2

238U PFNS E =2.9 MeV
n1.0

0.8

0.6

0.4

0.2

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f) Ei2.90E+610-1

110Outgoing Neutron Energy (MeV)

Figure 43: 238 U prompt fission neutron spectrum

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =5.0 MeV
n1.2

238U PFNS E =5.0 MeV
n

1.0

0.8

0.6

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

0.4
10-1

1Outgoing Neutron Energy (MeV)

Figure 44: 238 U prompt fission neutron spectrum

2310

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =6.0 MeV
n238U PFNS E =6.0 MeV
n1

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

110Outgoing Neutron Energy (MeV)Figure 45: 238 U prompt fission neutron spectrum

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =7.0 MeV
n238U PFNS E =7.0 MeV
n1

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-11Outgoing Neutron Energy (MeV)

Figure 46: 238 U prompt fission neutron spectrum

2410

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =8.01 MeV
n

238U PFNS E =8.01 MeV
n1

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

110Outgoing Neutron Energy (MeV)

Figure 47: 238 U prompt fission neutron spectrum

Ratio to Maxwellian at T=1.382 MeV

238U PFNS E =8.94 MeV
n238U PFNS E =8.94 MeV
n1

ENDF-VII.1
JEFF-3.1
JENDL-4.0
238-U(n,f)

10-1

1

Outgoing Neutron Energy (MeV)

Figure 48: 238 U prompt fission neutron spectrum

2510239Pu(n,tot)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,tot)

239Pu(n,tot)

1010-210-1

110Incident Neutron Energy (MeV)

Figure 49: 239 Pu total cross section
239Pu(n,ela)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,ela)

239Pu(n,ela)

1010-210-1

1Incident Neutron Energy (MeV)

Figure 50: 239 Pu elastic scattering cross section

2610

239Pu(n,f)

Cross Section (barns)

3.5

239Pu(n,f)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,f)

3.0

2.5

2.0

1.5
10-2

10-1

1

10Incident Neutron Energy (MeV)

Figure 51: 239 Pu fission cross section
239Pu(n,γ)

239Pu(n,γ)

Cross Section (barns)

110-1

10-2

ENDF-VII.1
JEFF-3.1
JENDL-4.0
1962 Hopkins
1958 Andreev
1956 Spivak
19 Kononov

10-3
10-2

10-1

1Incident Neutron Energy (MeV)

Figure 52: 239 Pu radiative capture cross section

2710

239Pu(n,2n)

Cross Section (barns)

0.6

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,2n)

239Pu(n,2n)

0.4

0.2

010

Incident Neutron Energy (MeV)

Figure 53: 239 Pu (n,2n) reaction cross section
239Pu(n,3n)

Cross Section (barns)

0.4

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,3n)

239Pu(n,3n)
0.3

0.2

0.1

0131415161718Incident Neutron Energy (MeV)

Figure 54: 239 Pu (n,3n) reaction cross section

2819

239Pu(n,inl)

2.0

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0
239-Pu(n,inl)

239Pu(n,inl)
1.5

1.0

0.5

10-2

10-1

1

10Incident Neutron Energy (MeV)

Figure 55: 239 Pu total inelastic scattering cross section

Cross Section (barns)

239Pu(n,n1)

0.6

ENDF-VII.1
JEFF-3.1
JENDL-4.0

239Pu(n,n1)
0.4

0.2

010-2

10-1

1Incident Neutron Energy (MeV)

Figure 56: 239 Pu inelastic to the first excited state2910

239Pu(n,n2)

ENDF-VII.1
JEFF-3.1
JENDL-4.0

Cross Section (barns)

0.5

239Pu(n,n2)

0.4

0.3

0.2

0.1

010-1

110

Incident Neutron Energy (MeV)

Figure 57: 239 Pu inelastic to the second excited state
239Pu(n,n3)

Cross Section (barns)

ENDF-VII.1
JEFF-3.1
JENDL-4.0

239Pu(n,n3)

0.10

0.05

0 -1
101

Incident Neutron Energy (MeV)

Figure 58: 239 Pu inelastic to the third excited state3010

239Pu µ-bar1ENDF/B-VII.1
JEFF-3.1
JENDL-4.0
experimentµ-bar (elastic)

239Pu µ-bar
10-1

10-2

10-510-410-3

10-210-1

110Incident Neutron Energy (MeV)

Figure 59: 239 Pu average scattering cosine
239Pu prompt ν-bar

Prompt nubar

6

ENDF/B-VII.1
JENDL-4.0
experiment

239Pu prompt ν-bar

54351015Incident Neutron Energy (MeV)

Figure 60: 239 Pu average number of prompt fission neutrons

3120

239Pu(n,f) PFNS

Maxwellian T=1.382 MeV

dσ/dE (mbarns)

1200

ENDF/B-VII.1
JENDL-4.0
239-Pu(n,f) Ei=2.53E-2 eV

1000

800

600

239Pu(n,f) PFNS

400

10-210-1

1Energy (MeV)

Figure 61: 239 Pu prompt fission neutron spectrum

3210