1

Nuclear Data Evaluation for Actinoid Nuclides

T.Nakagawa, O.Iwamoto, N.Otuka, S.ChibaJapan Atomic Energy Agency

2

Nuclear Data EvaluationFor JENDL Actinoid File (JENDL/AC)Nuclides79 nuclides: Ac-225 to Fm-255

62 nuclides: JENDL-3.3 data were revised.17 nuclides: additional nuclides for JENDL/AC. Their half-lives are longer than 1day.

Neutron energy range1.0e-5 eV to 20 MeV

3

Resolved Resonance Parameters

SAMMY analyses Th-232, U-233, U-238, Pu-241: ENDF/B-VII.0 was adopted.U-235: JENDL-3.3 was adopted up to 500 eVPu-239: JENDL-3.3. New analysis has been done by Derrien.Pu-240: JENDL-3.3 was slightly modified.Np-236: New evaluation was done by Furutaka (JAEA).

Other nuclidesParameters of Multi-Level Breit-Wigner formula were modified to reproduce experimental data.Thermal fission and capture cross sections Evaluated from experimental data if available.

4

-40

-30

-20

-10

0

10

20

30

40

50U-

234

U-23

5

U-23

6

U-23

8

Np-2

37

Pu-2

38

Pu-2

39

Pu-2

40

Pu-2

41

Pu-2

42

Am-2

41

Am-2

42m

Am-2

43

Cm-2

42

Cm-2

43

Cm-2

44

Cm-2

45

(AC-J33)/

J33[%

]

capturefission

Changes from JENDL-3.3.Thermal Cross Sections

-85

-99.

6

5

Thermal cross section (b)JENDL-3.3 639.5Kalebin (1976) 624 ± 20Shinohara+ (1997) 854 ± 58Fioni+ (2001) 696 ± 48Bringer+ (2006) 714 ± 23

Present 697.1

LANL (2007) about 660 ± 25

10-2 10-1 100102

103

104

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

Ju.V.Adamchuk+ ('55) Ju.V.Adamchuk+ ('55) Ju.V.Adamchuk+ ('55) G.G.Slaughter+ ('61) H.Derrien+ ('75) S.M.Kalebin+ ('76)

241Am total

Present JENDL-3.3

Large thermal capture cross section is inconsistent with measured total cross sections.

241Am Capture Cross Section

6

Modification of Resonance Parameters

40 50 60100

101

102

103

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

M.G.Silbert ('76)

243Cm fission

present JENDL-3.3

7102 103 104 10510-1

100

101

102

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

F.D.Brooks+ ('66) G.de Saussure+ ('66) G.de Saussure+ ('66) R.B.Perez+ ('73) R.B.Perez+ ('73) F.Corvi+ ('75) V.N.Kononov+ ('75) R.Gwin+ ('76) G.V.Muradyan+ ('77) F.Corvi+ ('82)

235U capture present JENDL-3.3

235U Capture Cross Section500eV～2.25 keV; pointwise data were given

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UnresolvedResonance Parameters

Option LSSF=1Unresolved resonance parameters are used only for self-shielding calculations. Cross sections are given in MF=3.

Upper boundary energiesSelected to be high enough to avoid discontinuity of self-shielded cross sections. Code for URP evaluationASREP code

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Theoretical Calculation

CCONE codeDeveloped by O. Iwamoto (Nuclear Data Center, JAEA).Based on coupled channel optical model, DWBA, pre-equilibrium exciton model, statistical model.

Fission cross section can be calculated.Programming language: C++

10

Fission Cross Sections calculated with CCONE Code

102 103 104 105 106 107100

101

102

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

P.A.Seeger+ ('67) C.D.Bowman+ ('68) E.F.Fomushkin+ ('81) E.F.Fomushkin+ ('81) E.F.Fomushkin+ ('81) J.W.T.Dabbs+ ('83) J.W.T.Dabbs+ ('83) J.C.Browne+ ('84) J.C.Browne+ ('84) J.C.Browne+ ('84) V.F.Gerasimov+ ('87) V.A.Shigin+ ('91) B.I.Fursov+ ('97) B.I.Fursov+ ('97) V.F.Gerasimov+ ('97) T.Kai+ ('01)

242mAm fission

present CCONE

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238U(n,2n) Cross Section

10 15 200

0.5

1.0

1.5

Neutron Energy (MeV)

Cro

ss S

ectio

n (b

)

L.R.Veeser+ ( '78) H.Kar ius+ ('79) J .Frehaut+ ( '80) J .Frehaut+ ( '80) N.V.Kornilov+ ('80) P .Raics+ ( '80) T .B.Ryves+ ( '80) R.Pepelnik+ ( '85) V.Ya.Golovnya+ ('87) V.Ya.Golovnya+ ('87) C.Konno+ ( '93) A.A.Filatenkov+ ( '99) A.A.Filatenkov+ ( '99)

238U (n,2n)

CCONE (= present) JENDL-3.3 ENDF/B-VII .0

Frehaut+ (befor correction)

12

237Np Capture Cross Section

103 104 105 106 10710-2

10-1

100

101

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

D.C.Stupegia+ ('67) M.Lindner+ ('76) L.W.Weston+ ('81) Ju.N.Trofimov+ ('83) N.N.Buleeva+ ('88) N.N.Buleeva+ ('88) N.N.Buleeva+ ('88) Katsuhei Kobayashi+ ('02) Katsuhei Kobayashi+ ('02) E.I.Esch+ ('05)

237Np capture

present (=CCONE) JENDL-3.3 ENDF/B-VII.0

13

241Am Capture Cross Section

103 104 105 106 10710-3

10-2

10-1

100

101

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

G.N.Flerov+ ('67) D.B.Gayther+ ('77) K.Wisshak+ ('80) K.Wisshak+ ('80) K.Wisshak+ ('80) K.Wisshak+ ('80) K.Wisshak+ ('80) K.Wisshak+ ('82) G.Vanpraet+ ('85)

241Am capture

present (=CCONE) JENDL-3.3 ENDF/B-VII

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Isomeric Ratio of 241Am Capture Cross Section

present IRFBR 0.850PWR 0.878BWR 0.884

10-2 10-1 100 101 102 103 104 105 106 1070.4

0.5

0.6

0.7

0.8

0.9

1.0

Neutron Energy (eV)

Isom

eric

ratio

(gro

und

stat

e)

241Am capture

present JENDL-3.3 ENDF/B-VII JEFF-3.1

Bringer et al. ('06) Fioni et al. ('01) Shinohara et al. ('97) Wisshak et al. ('82) Dovbenko et al. ('71) Gavrilov et al. ('75) Harbour et al. ( '73) Bak et al. ('67)

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Fission Cross Sections

Evaluated with GMA code and experimental dataGMA is a least-squares fitting code developed by Poenitz, and improved by S.Chiba et al.Simultaneous Evaluation for important nuclides

16103 104 105 106 107

10-2

10-1

100

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

K.Wisshak+ ('83) K.Wisshak+ ('83) K.Wisshak+ ('83) K.Wisshak+ ('83) E.F.Fomushkin+ ('84) B.I.Fursov+ ('85) K.Kanda+ ('87) H.Knitter+ ('88) K.Kobayashi+ ('99) V.Ya.Golovnya+ ('99) A.Laptev+ ('04) Baba+ ('07)

243Am fission

present JENDL-3.3. ENDF/B-VII.0

243Am Fission Cross Section

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243Cm Fission Cross Section

103 104 105 106 107100

101

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

R.R.Fullwood+ ('70) M.G.Silbert ('76) E.F.Fomushkin+ ('87) E.F.Fomushkin+ ('90) E.F.Fomushkin+ ('91) B.I.Fursov+ ('97) B.I.Fursov+ ('97)

243Cm fission

Present JENDL-3.3

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235U Fission Cross Section

105 106 1071

1.5

2

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

A.D.Carlson ('83) I.D.Alkhazov+ ('83) A.D.Carlson+ ('84) L.W.Weston+ ('84) C.M.Herbach+ ('85) C.M.Herbach+ ('85) R.Arlt+ ('85) R.Arlt+ ('85) I.D.Alkhazov+ ('86) Li Jingwen+ ('86) J.Rapaport+ ('87)

JENDL-3.3

235U fission I .D.Alkhazov+ ('88) I .D.Alkhazov+ ('88) N.N.Buleeva+ ('88) R.G.Johnson+ ('88) T.Iw asaki+ ('88) K.Merla+ ('91) K.Merla+ ('91) R.G.Johnson+ ('91) V.A.Kalinin+ ('91)

D.Rochman+ ('05)

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235U Fission Cross Section

1061

1.2

1.4

1.6

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

A.D.Carlson ('83) I.D.Alkhazov+ ('83) A.D.Carlson+ ('84) L.W.Weston+ ('84) C.M.Herbach+ ('85) C.M.Herbach+ ('85) R.Arlt+ ('85) R.Arlt+ ('85) I.D.Alkhazov+ ('86) Li Jingwen+ ('86) J.Rapaport+ ('87)

present JENDL-3.3 ENDF/B-VII.0

235U fission I .D.Alkhazov+ ('88) I .D.Alkhazov+ ('88) N.N.Buleeva+ ('88) R.G.Johnson+ ('88) T.Iw asaki+ ('88) K.Merla+ ('91) K.Merla+ ('91) R.G.Johnson+ ('91) V.A.Kalinin+ ('91)

D.Rochman+ ('05)

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NuclidesU-233, U-235, U-238, Pu-239, Pu-240, Pu-241Neutron energy range10 keV to 20 MeVCode used for the evaluationSOK code developed by T. Kawano (LANL)

Simultaneous Evaluation of Fission Cross Sections

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Simultaneous Evaluation of Fission Cross Sections

105 106 107

1

1.5

2

Neutron Energy (eV)

Cro

ss S

ectio

n (b

)

235U fission

JENDL-3.3 JENDL/AC previous SOK

105 106 107

1.5

2

2.5

Neutron Energy (eV)

239Pu fission

JENDL-3.3 JENDL/AC previous SOK

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Adjustment of Model Parameters to Integral Data

Integral data are also experimental data.Model parameters used in the CCONE calculation for important nuclides will be adjusted to integral data.Integral data: k-eff of FBR and small cores at LANL, etc.Preliminary adjustment was successfully done.

Cross-section changes were small and reasonable. C/E values were considerably improved.

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SummaryJENDL/AC is under development.Evaluated data are given to 79 nuclides from Ac to Fm.Results of theoretical calculation with CCONE code are widely adopted.The present evaluation is based on experimental data as much as possible.Integral data are considered as experimental data, and model parameters of CCONE code will be adjusted to them.JENDL/AC will be released in FY2007.Covariance data will be provided for important nuclides in FY2008.