benchmark test of jendl-4.0 based on integral experiments ... · benchmark test of jendl-4.0 ....

Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol. 2, pp.346-357 (2011)

c© 2011 Atomic Energy Society of Japan, All Rights Reserved.

346

ARTICLE

Benchmark Test of JENDL-4.0 Based on Integral Experiments at JAEA/FNS

Chikara KONNO 1,*, Kosuke TAKAKURA 1, Masayuki WADA 2, Keitaro KONDO 1,

Seiki OHNISHI 1, Kentaro OCHIAI 1 and Satoshi SATO 1

1 Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan 2 Japan Computer System, Mito-shi, Ibaraki-ken, 310-0805, Japan

The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0, was released in May, 2010. As the benchmark test of JENDL-4.0 in the shielding and fusion neutronics fields, we analyzed many integral benchmark experiments (in-situ and Time-of-Flight (TOF) experiments) with DT neutrons at JAEA/FNS with the MCNP code and JENDL-4.0. The experiments with assemblies including beryllium, carbon, silicon, vana-dium, copper, tungsten and lead, nuclear data of which were revised in JENDL-4.0, were selected for this benchmark test. As a result, it is demonstrated that JENDL-4.0 improves some problems pointed out in JENDL-3.3. JENDL-4.0 is comparable to ENDF/B-VII.0 and JEFF-3.1.

KEYWORDS: JENDL-3.3, JENDL-4.0, benchmark test, integral experiment, MCNP, FNS

I. Introduction1

The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0,1) was released in May, 2010. It is essential to validate JENDL-4.0 through analyses of integral benchmark experiments. Many integral bench-mark experiments2,3) for nuclear data verification have been carried out with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA) for 25 years. Thus we analyzed these experiments with JENDL-4.0 as one of benchmark tests of JENDL-4.0 in the shielding and fusion neutronics fields. Analyses with the recent other nuclear data, JENDL-3.3,4) JEFF-3.15) and ENDF/B-VII.0,6) were also carried out in order to uncover whether JENDL-4.0 was better or not. This paper describes these benchmark tests.

II. Overview of Integral Experiments at JAEA/FNS

Two types of integral experiments for nuclear data verifi-cation with DT neutrons have been performed for 25 years at JAEA/FNS. One is an in-situ experiment2) and the other is a Time-of-flight (TOF) experiment.3)

A typical experimental configuration of the in-situ expe-riment is shown in Fig. 1. The shape of most of the experimental assemblies is a quasi cylinder as shown in Fig. 1. The effective diameter of the assembly is 630 mm. The thickness of the experimental assemblies is different each experiment depending on material volume which we own. Neutron spectra, reaction rates of dosimetry reactions, gamma heating rates and so on are measured inside the ex-perimental assembly in this experiment. We have experimental data of lithium oxide, beryllium, graphite, sili- *Corresponding author, E-mail: [email protected]

con carbide (SiC), vanadium, iron, a type 316 stainless steel (SS316), copper, tungsten, etc.

A typical experimental configuration of the TOF experi-ment is shown in Fig. 2. Angular neutron spectra above 100 keV leaking from thinner assemblies than those as shown in Fig. 1 are measured for lithium oxide, beryllium, graphite, liquid-nitrogen, liquid-oxygen, iron, lead, etc. by

315mm

d+

200mm

T-Target

Experimental assembly

608mm

Aluminum frame

Cooling water

Fig. 1 Experimental configuration of in-situ experiment.

Benchmark Test of JENDL-4.0 Based on Integral Experiments at JAEA/FNS 347

VOL. 2, OCTOBER 2011

using the collimator system in this experiment. The size of experimental assemblies is different each experiment de-pending on material volume which we own. III. Analysis

The Monte Carlo code MCNP-4C7) and the A Compact ENDF (ACE) file FSXLIB-J408) of JENDL-4.0 were used for the analysis of the experiments. Calculations with the ACE files of the following nuclear data libraries were also carried out for comparison.

- JENDL-3.3 (ACE file : FSXLIB-J339)) - JEFF-3.1 (ACE file : MCJEFF3.1NEA10)) - ENDF/B-VII.0 (ACE file : endf70 in MCNP Data11))

JENDL Dosimetry file 9912) was adopted as the reaction cross section data for calculation of dosimetry reaction rates.

All the nuclear data in JENDL-3.3 are not modified in JENDL-4.0. The experiments with assemblies shown in Table 1 including beryllium, carbon, silicon, vanadium, copper, tungsten and lead, nuclear data of which were re-vised in JENDL-4.0, were selected for this benchmark test. Iron data are also modified in JENDL-4.0, but the bench-

mark results for the iron experiments with JENDL-4.0 have been reported13) already.

IV. Results and Discussion 1. Beryllium Experiments

The main modifications of the 9Be data from JENDL-3.3 to JENDL-4.0 are the followings; 1) the elastic scattering cross section data are re-calculated and slightly modified around the peak of ~ 600 keV as shown in Fig. 3, 2) the (n,2n) reaction cross section data are revised only near the threshold energy as shown in Fig. 3, 3) the (n,γ) reaction cross section data increase by ~10%.

Therefore the difference between the calculation results with JENDL-3.3 and -4.0 for the beryllium TOF and in-situ experiments is not so large. Figure 4 shows the measured and calculated angular neutron spectra leaking from the be-ryllium slab of 152 mm in thickness in the beryllium TOF experiment. All the calculated spectra agree with the meas-ured ones very well. Figure 5 plots ratios of calculation to experiment (C/E) for reaction rates of the 93Nb(n,2n)92mNb, 115In(n,n’)115mIn, 6Li(n,α)T and 235U(n, fission) reactions, which are mainly sensitive to neutrons above 10 MeV, neu-trons above 0.3 MeV, thermal neutrons and thermal neutrons, respectively. It is noted that the overestimation for the reac-tion rates of the 6Li(n,α)T and 235U(n, fission) reactions in the calculation result with JENDL-3.3 reduces by ~ 5% in that with JENDL-4.0.

Experiment Assembly Shape Size

Be in-situ Quasi

cylinder 630 mm in effective diameter 455 mm in thickness

TOF Quasi cylinder

630 mm in effective diameter 51, 152 mm in thickness

Gra- phite

in-situ Quasi cylinder

630 mm in effective diameter 608 mm in thickness

TOF Quasi cylinder

630 mm in effective diameter 51, 202, 405 mm in thickness

SiC in-situ Rectangle 457 mm x 457 mm x 711 mm in thickness

V in-situ Rectangle 254 mm x 254 mm x 254 mm in thickness covered with 50 mm thick graphite

Cu in-situ Quasi cylinder


W in-situ Quasi cylinder


Pb TOF Quasi cylinder

630 mm in effective diameter 51, 203, 406 mm in thickness

Fig. 2 Experimental configuration of TOF experiment.

Fig. 3 Cross section data of 9Be.

0

2

4

6

8

10

10-1 100 101

JENDL-3.3JENDL-4.0

Cro

ss s

ectio

n [b

]

Neutron Energy [MeV]

(a) elastic scattering

0.0

0.1

0.2

0.3

0.4

0.5

0.6

1 2 3 4 5

JENDL-3.3JENDL-4.0

Cro

ss s

ectio

n [b

]


(b) (n,2n) reaction

Table 1 Experimental assemblies

348 Chikara KONNO et al.

PROGRESS IN NUCLEAR SCIENCE AND TECHNOLOGY

10-4

10-3

10-2

10-1

100

101

10-1 100 101

Expt.ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0

Angu

lar F

lux

[1/s

r/m2 /le

thar

gy/s

ourc

e]


0 deg.

12.2 deg.(x1/ 10)

24.9 deg.(x1/ 100)

10-7

10-6

10-5

10-4

10-3

10-1 100 101


Angu

lar F

lux

[1/s

r/m

2 /leth

argy

/sou

rce]


41.8 deg.(x1/ 1000)

66.8 deg. (x1/10000)

Fig. 4 Angular neutron spectra leaking from beryllium slab of

152 mm in thickness in beryllium TOF experiment.

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 100 200 300 400

ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0

Cal

c. /

Exp

t.

Depth in assembly [mm]

(a) Reaction rate of93Nb(n,2n)92mNb

Expt. Error

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 100 200 300 400

ENDF/B-VII.0JEFF-3 .1JENDL-3.3JENDL-4.0

Calc

. / E

xpt.

Depth in the assembly [mm]

(b) Reaction rate of115In(n,n')115mIn

Expt. Er ror

0.8

0.9

1.0

1.1

1.2

1.3

1.4

0 100 200 300 400


Cal

c. /

Exp

t.


(c) Reaction rate of6Li(n,α)T

Expt. Error

0.9

1.0

1.1

1.2

1.3

1.4

1.5

0 100 200 300 400

ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0C

alc.

/ E

xpt.


(d) Reaction rate of235U(n,fission)

Expt. Error

Fig. 5 C/E in beryllium in-situ experiment.



2. Graphite Experiments The main modifications of the natC data from JENDL-3.3

to JENDL-4.0 are the followings; 1) the (n,γ) reaction cross section data increase by ~10% in the 1/v part, 2) the angular distribution data of the elastic scattering are taken from JENDL-3.2 as shown in Fig. 6. Thus the difference between the calculation results with JENDL-3.3 and -4.0 for the gra-phite TOF and in-situ experiments is not so large as shown in Figs. 7 and 8.

10-2

10-1

100

101

-1-0.500.51

JENDL-3.3JENDL-4.0

Ang

ula

r d

istr

ibu

tion

[1/s

r]

µ=cosθ (Center-of-Mass System) Fig. 6 Angular distribution data of elastic scattering of natC for

14 MeV neutron.

10-5

10-4

10-3

10-2

10-1

100

101

10-1 100 101


Angu

lar F

lux

[1/s

r/m2 /le

thar

gy/s

ourc

e]


0 deg .

12.2 d eg .(x1/10)

24.9 d eg .(x1/100)

Fig. 7 Angular neutron spectra leaking from graphite slab of 203 mm in thickness in graphite TOF experiment.

10-5

10-4

10-3

10-2

10-1

100

101

10-1 100 101


Angu

lar F

lux

[1/s

r/m2 /le

thar

gy/s

ourc

e]


0 deg .

12.2 d eg .(x1/10)

24.9 d eg .(x1/100)

Fig. 7 (Continued).

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600 800

ENDF/B-V II.0JEFF-3.1JENDL-3 .3JENDL-4 .0

Cal

c. /

Expt

.



Expt. Erro r

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600 800


Cal

c. /

Exp

t.


(b) Reaction rate of27Al(n,α)24Na

Expt. Error

Fig. 8 C/E in graphite in-situ experiment.



0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600 800


Cal

c. /

Expt

.


(c) Reaction rate of115In(n,n')115mIn

Expt. Error

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600 800


Calc

. / E

xpt.


(d) Reaction rate of235U(n,fission)

Expt. Error

Fig. 8 (Continued).

10-1

100

101

102

10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101

JENDL-3.3JENDL-4.0

Cro

ss s

ectio

n [b

]



0.0

0.2

0.4

0.6

0.8

1.0

0 5 10 15 20

JENDL-3.3JENDL-4.0

Cro

ss s

ectio

n [b

]


(b) non-elastic scattering

Fig. 9 Cross section data of 28Si.

3. SiC Experiment The main modifications of the natC data from JENDL-3.3

to JENDL-4.0 are described above. The Si isotope data in JENDL-4.0 were newly evaluated by using the TNG code.14) The resolved resonance parameters were taken from ENDF/B-VII.0. Thus there are many modifications in the Si isotope data of JENDL-4.0. As an example, the total and non-elastic scattering cross section data of 28Si, the abun-dance of which is over 90%, are shown in Fig. 9. However the difference between the calculation results with JENDL-3.3 and -4.0 for the SiC in-situ experiment is small as shown in Fig. 10.

0.6

0.8

1.0

1.2

1.4

0 200 400 600 800

ENDF/B-VII .0JEFF-3.1JENDL-3.3JENDL-4.0

Cal

c. /

Exp

t.



Expt. Error

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1.4

0 200 400 600 800

ENDF/B-VI I.0JEFF-3.1JENDL-3.3JENDL-4.0

Cal

c. /

Exp

t.



Expt. Error

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1.4

0 200 400 600 800


Cal

c. /

Exp

t.


(c) Reaction rate of235U(n,fission)

Expt. Error

Fig. 10 C/E in SiC in-situ experiment.



0.6

0.8

1.0

1.2

1.4

0 200 400 600 800


Cal

c. /

Expt

.


(d) γ-ray Heating Rate

Expt. Error

Fig. 10 (Continued). 4. Vanadium Experiment

The vanadium data in JENDL-4.0 were newly evaluated as 50V (abundance : 0.25%) and 51V (abundance : 99.75%) by using the CCONE code,15) while those in JENDL-3.3 were evaluated as natV. There are many modifications in the vanadium in JENDL-4.0 as shown in Fig. 11.

The calculation results for the vanadium in-situ experi-ment are compared with the measurement in Figs. 12 and 13. The underestimation for the reaction rate of the 115In(n,n’)115mIn reaction and for neutrons below a few keV in the calculation result with JENDL-3.3 is slightly im-proved in that with JENDL-4.0.

10-3

10-2

10-1

100

101

102

103

10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101

V-nat JENDL-3.3V-50 JENDL-4.0*0.0025V-51 JENDL-4.0*0.9975

Cro

ss s

ectio

n [b

]



0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 5 10 15 20

V-nat JENDL-3.3V-50 JENDL-4.0*0.0025V-51 JENDL-4.0*0.9975

Cro

ss s

ectio

n [b

]


(b) non-elastic scattering

Fig. 11 Cross section data of V.

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-6 10-5 10-4 10-3 10-2 10-1 100 101


Neu

tron

Flu

x [n

/cm

2 /leth

argy

/sou

rce]

Neutron Energy [MeV] Fig. 12 Neutron spectra at depth of 78 mm in vanadium in-situ experiment.



0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 100 200 300


Cal

c. /

Expt

.



Expt. Error

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 100 200 300


Calc

. / E

xpt.



Expt. Erro r

Fig. 13 C/E in vanadium in-situ experiment.

5. Copper Experiment

The modifications of the 63Cu and 65Cu data from JENDL-3.3 to JENDL-4.0 are only the cross section and angular distribution data of the elastic scattering as shown in Figs. 14 and 15.

The calculation results for the copper in-situ experiment are compared with the measured ones in Figs. 16 and 17. It is noted that the overestimation for neutrons from 10 keV to 100 keV in the calculation result with JENDL-3.3 reduces in that with JENDL-4.0, while the underestimation for neutrons below 1 keV in that with JENDL-3.3 is slightly enhanced in that with JENDL-4.0. 6. Tungsten Experiment

The tungsten isotope data in JENDL-4.0 were newly eva-luated with CCONE. There are many modifications in the tungsten isotope data in JENDL-4.0. Particularly the (n,2n) reaction cross section data of the tungsten isotopes in JENDL-4.0 are different from those in JENDL-3.3 as shown in Fig. 18.

The calculation results for the tungsten in-situ experiment are compared with the measured ones in Figs. 19 and 20. The underestimation for the reaction rate of the 115In(n,n’)115mIn reaction in the calculation result with JENDL-3.3 is improved in that with JENDL-4.0, while neu-trons above 10 MeV are slightly overestimated with the depth in the calculation result with JENDL-4.0.

10-2

10-1

100

101

102

103

10-2 10-1 100

JENDL-3.3JENDL-4.0

(a) 63Cu

Cro

ss s

ectio

n [b

]


10-1

100

101

102

103

10-2 10-1 100

JENDL-3.3JENDL-4.0

(b) 65Cu

Cro

ss s

ectio

n [b

]

Neutron Energy [MeV] Fig. 14 Elastic scattering cross section data of Cu isotopes.

10-3

10-2

10-1

100

101

102

-1-0.500.51

JENDL-3.3JENDL-4.0A

ngul

ar d

istr

ibut

ion

[1/s

r]

µ=cosθ (Center-of-Mass System)

(a) 63Cu

10-3

10-2

10-1

100

101

102

-1-0.500.51

JENDL-3.3JENDL-4.0A

ngul

ar d

istr

ibut

ion

[1/s

r]

µ=cosθ (Center-of-Mass System)

(b) 65Cu

Fig. 15 Angular distribution data of elastic scattering of Cu

isotopes for 15 MeV neutron.



10-8

10-7

10-6

10-5

10-4

10-6 10-5 10-4 10-3 10-2 10-1 100 101

Expt.ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0Ne

utro

n Fl

ux [n

/cm

2 /leth

argy

/sou

rce]

Neutron Energy [MeV] Fig. 16 Neutron spectra at depth of 228 mm in copper in-situ experiment.

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600


Calc

. / E

xpt.


(a) Integrated neutron flux En > 10 MeV

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400 600

ENDF/B-VI I.0JEFF-3 .1JENDL-3.3JENDL-4.0

Cal

c. /

Exp

t.


(b) Inte gra ted neutron flux 10 < En < 100 keV

0.2

0.4

0.6

0.8

1.0

1.2

0 200 400 600

ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0C

alc.

/ Ex

pt.


(c) Integrated neutron flux 10 < En < 100 eV

Expt. Error

0.4

0.6

0.8

1.0

1.2

1.4

0 200 400 600


Cal

c. /

Expt

.



Expt. Err or

Fig. 17 C/E in copper in-situ experiment.



0.0

0.5

1.0

1.5

2.0

2.5

5 10 15 20

JENDL-3.3JENDL-4.0

(a) 182WC

ross

sec

tion

[b]


0.0

0.5

1.0

1.5

2.0

2.5

5 10 15 20

JENDL-3.3JENDL-4.0

(b) 183W

Cro

ss s

ectio

n [b

]


0.0

0.5

1.0

1.5

2.0

2.5

5 10 15 20

JENDL-3.3JENDL-4.0

(c) 184W

Cro

ss s

ectio

n [b

]


0.0

0.5

1.0

1.5

2.0

2.5

5 10 15 20

JENDL-3.3JENDL-4.0

(d) 186W

Cro

ss s

ectio

n [b

]

Neutron Energy [MeV] Fig. 18 (n,2n) reaction cross section data of W isotopes.

10-7

10-6

10-5

10-4

10-3 10-2 10-1 100 101

Expt.ENDF/B-VII.0JEFF-3.1JENDL-3.3JENDL-4.0N

eutr

on F

lux

[n/c

m2 /leth

argy

/sou

rce]

Neutron Energy [MeV] Fig. 19 Neutron spectra at depth of 228 mm in tungsten in-situ experiment.



0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400

ENDF/B-VII .0JEFF-3.1JENDL-3.3JENDL-4.0

Calc

. / E

xpt.


(a) Reaction rate of93Nb(n,2n) 92mNb

Expt. Error

0.7

0.8

0.9

1.0

1.1

1.2

1.3

0 200 400

ENDF/B-V II.0JEFF-3.1JENDL-3.3JENDL-4.0

Calc

. / E

xpt.



Exp t. Erro r

0.4

0.6

0.8

1.0

1.2

1.4

0 200 400


Cal

c. /

Exp

t.


(c) Reaction rate of197Au(n,γ)198Au

Expt. Error

0.0

0.5

1.0

1.5

2.0

0 200 400


Cal

c. /

Expt

.



Expt. Error

Fig. 20 C/E in tungsten in-situ experiment.

0.0

0.5

1.0

1.5

2.0

2.5

5 10 15 20

JENDL-3.3JENDL-4.0

(a) 204Pb

Cro

ss s

ectio

n [b

]


0.0

0.5

1.0

1.5

2.0

2.5

3.0

5 10 15 20

JENDL-3.3JENDL-4.0

(b) 206Pb

Cro

ss s

ectio

n [b

]


0.0

0.5

1.0

1.5

2.0

2.5

3.0

5 10 15 20

JENDL-3.3JENDL-4.0

(c) 207Pb

Cro

ss s

ectio

n [b

]


0.0

0.5

1.0

1.5

2.0

2.5

3.0

5 10 15 20

JENDL-3.3JENDL-4.0

(d) 208Pb

Cro

ss s

ectio

n [b

]

Neutron Energy [MeV] Fig. 21 (n,2n) reaction cross section data of Pb isotopes.



10-4

10-3

10-2

10-1

100

101

10-1 100 101


Angu

lar F

lux

[1/s

r/m2 /le

thar

gy/s

ourc

e]


0 deg.

12.2 deg.(x1/10)

24.9 deg.(x1/100)

10-7

10-6

10-5

10-4

10-3

10-1 100 101


Angu

lar F

lux

[1/s

r/m

2 /leth

argy

/sou

rce]


41.8 deg.(x1/ 100)

66.8 deg. (x1/1000)

Fig. 22 Angular neutron spectra leaking from lead slab of

203 mm in thickness in lead TOF experiment.

7. Lead Experiment The lead isotope data in JENDL-4.0 were also newly

evaluated with CCONE. There are many modifications in the lead isotope data in JENDL-4.0. Particularly the (n,2n) reaction cross section data of the lead isotopes in JENDL-4.0 are different from those in JENDL-3.3 as shown in Fig. 21.

The calculation results for the lead TOF experiment are compared with the measured ones in Fig. 22. The underes-timation for neutrons above 1 MeV and the overestimation for neutrons below 1 MeV in the calculation result with JENDL-3.3 are resolved in that with JENDL-4.0.

V. Conclusion

We analyzed the fusion neutronics integral experiments at JAEA/FNS with JENDL-4.0 as one of benchmark tests of JENDL-4.0 released in May, 2010. Through the comparison with the calculation results with JENDL-3.3, the followings are found out. 1) Beryllium experiments : The overestimation for the reac-

tion rates of the 6Li(n,α)T and 235U(n, fission) reactions, which are mainly sensitive to thermal neutrons, in the calculation result with JENDL-3.3 by ~ 5% reduces in that with JENDL-4.0.

2) Graphite and SiC experiments : The calculation results with JENDL-4.0 are almost the same as those with JENDL-3.3.

3) Vanadium experiment : The underestimation for the reac-tion rate of the 115In(n,n’)115mIn reaction and for neutrons below a few keV in the calculation result with JENDL-3.3 is slightly improved in that with JENDL-4.0.

4) Copper experiment : The overestimation for neutrons from 10 keV to 100 keV in the calculation result with JENDL-3.3 reduces in that with JENDL-4.0, while the underestimation for neutrons below 1 keV in that with JENDL-3.3 is slightly enhanced in that with JENDL-4.0.

5) Tungsten experiment : The underestimation for the reac-tion rate of the 115In(n,n’)115mIn reaction in the calculation result with JENDL-3.3 is improved in that with JENDL-4.0, while neutrons above 10 MeV are slightly overestimated with the depth in the calculation result with JENDL-4.0.

6) Lead experiment : The underestimation for neutrons above 1 MeV and the overestimation for neutrons below 1 MeV in the calculation result with JENDL-3.3 are drastically improved in that with JENDL-4.0.

It is concluded that some problems in JENDL-3.3 are re-solved in JENDL-4.0. Moreover it is demonstrated that JENDL-4.0 is comparable to ENDF/B-VII.0 and JEFF-3.1. References

1) K. Shibata, O. Iwamoto, T. Nakagawa, N. Iwamoto, A. Ichihara, S. Kunieda, S. Chiba, K. Furutaka, N. Otuka, T. Ohsawa, T. Murata, H. Matsunobu, A. Zukeran, S. Kamada, J. Katakura, “JENDL-4.0: a new library for nuclear science and engineer-ing,” J. Nucl. Sci. Technol., 48, 1-30 (2011).

2) F. Maekawa, C. Konno, Y. Kasugai, Y. Oyama, Y. Ikeda, Data Collection of fusion neutronics benchmark experiment con-ducted at FNS/JAEA, JAERI-Data/Code 98-021, Japan Atomic



Energy Research Institute (JAERI) (1998). 3) Y. Oyama, S. Yamaguchi, H. Maekawa, Experimental results

of angular neutron flux spectra leaking from slabs of fusion reactor candidate materials, I, JAERI-M 90-092, Japan Atomic Energy Research Institute (JAERI) (1990).

4) K. Shibata, T. Kawano, T. Nakagawa, O. Iwamoto, J. Katakura, T. Fukahori, S. Chiba, A. Hasegawa, T. Murata, H. Matsunobu, T. Ohsawa, Y. Nakajima, T. Yoshida, A. Zukeran, M. Kawai, M. Baba, M. Ishikawa, T. Asami, T. Watanabe, Y. Watanabe, M. Igashira, N. Yamamuro, H. Kitazawa, N. Yamano, H. Takano, “Japanese evaluated nuclear data library version 3 revision-3: JENDL-3.3,” J. Nucl. Sci. Technol., 39, 1125-1136 (2002).

5) A. Koning, R. Forrest, M. Kellett, R. Mills, H. Henriksson, Y. Rugama (Ed.), The JEFF-3.1 nuclear data library, JEFF Re-port 21, OECD Nuclear Energy Agency (2006).

6) M.B. Chadwick, P. Oblozinsky, M. Herman, N.M. Greene, R.D. McKnight, D.L. Smith, P.G. Young, R.E. MacFarlane, G.M. Hale, S.C. Frankle, A.C. Kahler, T. Kawano, R.C. Little, D.G. Madland, P. Moller, R.D. Mosteller, P.R. Page, P. Talou, H. Trellue, M.C. White, W.B. Wilson, R. Arcilla, C.L. Dunford, S.F. Mughabghab, B. Pritychenko, D. Rochman, A.A. Son-zogni, C.R. Lubitz, T.H. Trumbull, J.P. Weinman, D.A. Brown, D.E. Cullen, D.P. Heinrichs, D.P. McNabb, H. Derrien, M.E. Dunn, N.M. Larson, L.C. Leal, A.D. Carlson, R.C. Block, J.B. Briggs, E.T. Cheng, H.C. Huria, M.L. Zerkle, K.S. Kozier, A. Courcelle, V. Pronyaev, S.C. van der Marck, “ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology,” Nucl. Data Sheets, 107, 2931-3059 (2006).

7) J.F. Briesmeister (Ed.), MCNP - A general Monte Carlo

n-particle transport code, version 4C, LA-13709-M, Los Alamos National Laboratory (LANL) (2000).

8) K. Okumura, K. Kojima, C. Konno, M. Nagao, T. Imaizumi, K. Kosako, The libraries FSXLIB and MATXS based on JENDL-4.0, to be published to JAEA-Data/Code, Japan Atomic Energy Agency (JAEA), [in Japanese].

9) K. Kosako, N. Yamano, T. Fukahori, K. Shibata, A. Hasegawa, The libraries FSXLIB and MATXSLIB based on JENDL-3.3, JAERI-Data/Code 2003-011, Japan Atomic Energy Research Institute (JAERI) (2003).

10) O. Cabellos, Processing of the JEFF-3.1 cross section library into a continuous energy Monte Carlo radiation transport and criticality data library, NEA/NSC/DOC 18, OECD NEA Data Bank (2006).

11) MCNP5/MCNPX-EXE, RSICC CODE PACKAGE CCC-740 (2009).

12) K. Kobayashi, T. Iguchi, S. Iwasaki, T. Aoyama, S. Shimakawa, Y. Ikeda, N. Odano, K. Sakurai, K. Shibata, T. Nakagawa, A. Zukeran, JENDL Dosimetry File 99 (JENDL/D-99), JAERI 1344, Japan Atomic Energy Research Institute (JAERI) (2002).

13) C. Konno, M. Wada, K. Kondo, S. Ohnishi, K. Takakura, K. Ochiai, S. Sato, “Detailed benchmark test of JENDL-4.0 iron data,” to be published to Fusion Eng. Design.

14) K. Shibata, C. Fu, Recent improvements of the TNG statistical model code, ORNL/TM-10093, Oak Ridge National Labora-tory (ORNL) (1986).

15) O. Iwamoto, “Development of a comprehensive code for nuc-lear data evaluation, CCONE, and validation using neutron-induced cross sections for uranium isotopes,” J. Nucl. Sci. Technol., 44, 687-697 (2007).

benchmark test of jendl-4.0 based on integral experiments ... · benchmark test of jendl-4.0 ....

Documents