Nuclear data needs for

transmutation system

Kazufumi TSUJIMOTO

JAEA

16th ASRC International Workshop " Nuclear Fission and Structure of Exotic Nuclei "

JAEA Tokai, Japan 18-20.March.2014

2

Contents

Background

Benefit and role of Partitioning and Transmutatin

Accelerator-Driven System (ADS) for minor actinide (MA)

transmutation

Current status of calculation accuracy of ADS

Benchmark calculation in IAEA-CRP

Uncertainty analysis utilizing covariance data of nuclear data for

calculation results

Summary

3

Benefit and role of P&T

Benefits of P&T on Management of High-Level Radioactive Wastes (HLW):

Reduction of long-term radiological toxicity

Reduction of dose for future inhabitants

Reduction of amount of HLW

Reduction of repository size

Recovery of valuable materials from wastes, and so on.

To mitigate difficulties

caused by long-term

nature of radioactivity

To extend capacity of a repository

Steady implementation of High Level Waste (HLW) disposal is one of the

most important issues even though we select to reduce dependency on

nuclear energy.

Partitioning and Transmutation (P&T) will be a key technology to reduce

the environmental burden of HLW.

Conceptual Design of ADS in JAEA

4

ガードベッセル

炉心支持構造物

ビーム窓

原子炉容器

炉心槽

主循環ポンプ

ビームダクト

内　筒

蒸気発生器

Guard Vessel

Inner tube

Beam Duct

Window

Core Vessel

Core Support

Steam

Generator Main Pump

Support

Structure

• Proton beam : 1.5GeV ~20MW

• Spallation target : LBE

• Coolant : LBE

• Subcriticality : keff = 0.97

• Thermal output : 800MWt

• Core height : 1000mm

• Core diameter : 2440 mm

• Fuel inventory : 4.2t (MA:2.5t)

• Fuel composition :

(MA + Pu)N+ZrN (Mono-nitride)

Inner : 70%MA+30%Pu

Outer : 54%MA+42%Pu

• Transmutation rate :

250kg(MA) / 300EFPD K. Tsujimoto, H.Oigawa, K.Kikuchi, et. al, “Feasibility of Lead-Bismuth-

Cooled accelerator-Driven System for Minor-Actinide Transmutation”,

Nucl. Tech. 161, 315-328 (2008).

Purpose : MA transmutation

Allowable maximum k-eff for ADS

5

Accidental insertion reactivity %k/k

Beam tube filled with Pb-Bi1) 0.32

Unusual temperature rise of

coolant Pb-Bi2)

0.69

Uncertainty for measurement3) 0.50

Uncertainty for calculation 1.00

Total 2.51

1) Pb-Bi entry to the vacuum beam

tube by failure of the beam

window

2) Coolant temperature rise of 1000K

in only fuel region by coast down

3) Accuracy of βeff influences directly

Allowable maximum k-eff was set at 0.97

High k-eff value implies low proton beam current and small power peaking, but

risk of approaching criticality under accidental conditions will increase.

The subcriticality must be set to adequate level considering accidental

insertion of reactivity and uncertainties for calculation and measured

reactivity.

K. Tsujimoto, T.Sasa, K.Nishihara, et. al, “Neutronics

Design for Lead-Mismuth Cooled Accelerator-Driven

System for Transmutation of Minor Actinide”, J. Nucl.

Sci. Tech., 41, 21-26 (2004).

Benchmark calculation

6

Participant Code Library

JAEA (Japan)

PHITS, NMTC or MCNPX (MC code for proton and

neutron >20MeV)

SLAROM (Cross section code)

TWODANT (Deterministic neutron transport code)

ORIGEN (Burn-up code)

JENDL-3.3

JENDL-4.0

JENDL-3.2

ENDF/B-VI

JEFF-3.0

CIEMAT (Spain)

EVOLCODE2 (MCNPX-based burnup code) JEFF-3.0

(JEFF-3.1) a)

(ENDF/B-VI) a)

KIT (Germany)

High energy particles are not analyzed.

C4P, ZMIX (Cross section code)

DANTSYS (Deterministic neutron tansport code)

TRAIN (Burn-up code)

ENDF/B-VII

JEFF-3.1

a) Library in parenthesis is only for the beginning of cycle (BOC).

To survey current status of neutronics design accuracy for ADS,

benchmark calculations were performed in framework of IAEA-CRP. Benchmark

problem was based on the ADS design proposed by JAEA.

Current status of neutronics design of ADS (1/2)

7

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, E

ND

F

JAEA EB6

CIEMAT EB6

KIT EB7

EB6

EB7

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, JE

FF

JAEAJEFF3.0

CIEMATJEFF3.0

CIEMATJEFF3.1

KITJEFF3.1

JF3.1

JF3.0

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, JE

ND

L

JAEA J3.3

JAEA J3.2

JAEA J4.0

J4.0

J3.2, J3.3

About 2% discrepancies in k-eff were found among the different nuclear

data libraries in a IAEA-CRP benchmark proposed to survey current status of

calculation accuracy of ADS by JAEA.

JENDL ENDF JEFF

Calculated results for IAEC-CRP benchmark proposed by JAEA (Burnup calculation for the first

burnup cycle of 600 EFPD with 800MWth ADS)

A. Stanculescu, “The IAEA Coordinated Research Project (CRP) on “Analytical and experimental benchmark analyses

of accelerator driven systems,” OECD/NEA 11th Information Exchange Meeting, San Francisco, 1-4 Nov. 2010. (2010)

Current status of neutronics design of ADS (2/2)

8

A. Stanculescu, “The IAEA Coordinated Research Project (CRP) on “Analytical and experimental benchmark analyses

of accelerator driven systems,” OECD/NEA 11th Information Exchange Meeting, San Francisco, 1-4 Nov. 2010. (2010)

-800

-600

-400

-200

0

200

400

600

All

Np-

237

Pu-2

38

Pu-2

39

Pu-2

40

Pu-2

41

Pu-2

42

Am

-241

Am

-243

Cm-2

44

N-1

5

Pb-2

04

Pb-2

06

Pb-2

07

Pb-2

08

Bi-2

09

Δke

ff d

ue t

o lib

rary

exc

hang

e (p

cm) J4 to EB7

J4 to JF31

Statistics error=20pcm

-800

-600

-400

-200

0

200

400

600

Cr-5

0Cr

-52

Cr-5

3Cr

-54

Mn-

55Fe

-54

Fe-5

6Fe

-57

Fe-5

8N

i-58

Ni-6

0N

i-61

Ni-6

2N

i-64

Mo-

92M

o-94

Mo-

95M

o-96

Mo-

97M

o-98

Mo-

100

Zr-9

0Zr

-91

Zr-9

2Zr

-94

Zr-9

6

Δke

ff d

ue t

o lib

rary

exc

hang

e (p

cm) J4 to EB7

J4 to JF31

Statistics error=20pcm

Δkeff due to library exchange of single nuclide from JENDL-4 to ENDF/B-VII or JEFF-3.1

estimated by MCNPX.

Identification of typical contributors for the differences among the calculated k-

eff values with different nuclear data library, JENDL-4.0 ENDF/B-VII, and

JEFF-3.1

Np-237, Pu-240, Am-241, Am-243, N-15, Fe-56

Cause of difference for calculation results

9

The cause of the difference between the calculation results with JENDL-4.0

and JENDL-3.3 are the cross section of

Am-241, Pb-206, and Pb-207.

H. Iwamoto, K. Nishihara, T. Sugawara and K. Tsujimoto, “Sensitivity and Uncertainty Analysis for a

Minor-Actinide Transmuter with JENDL-4.0”, Proc. ND2013, March 4 – 8, 2013, New York, U.S.A. (2013).

241Am

206Pb

207Pb

Fig. Nuclide-wise contribution for the difference

between calculated k-eff with JENDL-4.0 and 3.3

fis n cap inl el m

241Am 76 208 255 293 1 1

206Pb ― ― -34 674 3 -5

207Pb ― ― -14 560 14 -4

Table Reaction-wise contributions for the

difference between calculated k-eff with

JENDL-4.0 and 3.3

Summary of IAEA-CRP benchmark

10

Small difference between participants (codes) with same library.

k-effective disperse from 0.98 to 1.0 at BOC and 0.93 to 0.96 at

EOC.

To design an ADS is very difficult, if estimated

k-effective changes 2 to 3 %dk.

Uncertainty should be evaluated.

Uncertainty evaluation results for k-eff

11

Total uncertainty: 1.04%k

Fig. Nuclear- and reaction-wise Breakdown of criticality uncertainty

Pb, Fe-56 - Inelastic scattering - Elastic scatterring

MA, Pu - Capture cross section - Fission spectrum - Fission cross section

Major contributions to uncertainty LBE

Iron

Nitrogen N-15

Covariance data in JENDL-4.0

The uncertainty for k-eff of 1% is smaller than the result of IAEA

benchmark activity (2-3%).

241Am capture cross section

12

Fig. 241Am capture cross section

Un

cert

ain

ty (

%)

S.D

. (%

) Se

nsi

tivi

ty (

x10

-3)

Cro

ss s

ecti

on

(b

arn

)

107 106 105 104 103 102 101 100

107 106 105 104 103 102 101 100

107 106 105 104 103 102 101 100

Neutron energy (eV)

Sensitive region

Covariance data (1s S.D.)

Sensitivity coefficients

Energy breakdown of uncertainties

High uncertainty region

Neutron energy (eV) 106 105 103 104

Dominant region

Nuclear-data evaluation in this region (From ~1 keV to 1 MeV) affects the uncertainties in the integral parameters!

206Pb inelastic cross section

13 Neutron energy (eV)

Fig. 206Pb inelastic scattering cross section

Un

cert

ain

ty (

%)

S.D

. (%

) Se

nsi

tivi

ty (

x10

-3)

Cro

ss s

ecti

on

(b

arn

)

Neutron energy (eV)

107 106 105 104 103

107 106 105 104 103

107 106 105 104 103

Sensitivity coefficients

Covariance data (1s S.D.)

Energy breakdown of uncertainties

107 106 105

Sensitive region

High uncertainty region

Dominant region

Nuclear-data evaluation in this region (From threshold to ~3MeV) affects the uncertainties in the integral parameters of the ADS.

14

Summary

The current status of nuclear data is not so satisfactory for

the neutronics design of ADS as MA transmutation system.

The benchmark results for the neutronics parameters of ADS showed that there

were large differences among the calculated parameters with different nuclear

data libraries. For example, the differences for the calculated k-eff values were

approximately 3%.

Improvement of nuclear data, not only for MA and Pu but

also other nuclide (Pb, N-15), is significantly necessary for

transmutation system.

In order to meet the requirements for the nuclear data to

improve the design accuracy, continuous effort including

not only differential experiments but also the integral

experiments are important issue.