CO2-1

Research and Development for Accracy Improvement of Neutron Nuclear Data on
Long-lived Radioactive Nuclei at KURRI-Linac
spectra with and without 237Np sample by the 6Li-glass
detector is shown in Fig. 2. We could observe the significant differences between those spectra with a high signal
to noise ratio..
The results show that the developed capture -ray andtransmission neutron detection system has a high performance to determine the resonance parameters.

J. Hori, T. Sano, Y. Takahashi and H.Yashima
Research Reactor Institute, Kyoto University

INTRODUCTION: There are strong requests for reducing the uncertainty of neutron capture cross section
data of minor actinides (MAs) to estimate the transmutation rate of those long-lived radioactive nuclei in the innovative reactor system. In recent years, intense pulsed
spallation neutron sources such as J-PARC[1], n-TOF[2]and LANSCE[3] facilities became available to remarkably improve the precision of neutron TOF data. However,there are discrepancies out of a range of tolerance between current experimental results. It is understood thatthe unrecognized systematic errors make a difference. Inorder to recognize and reduce the systematic errors, theproject entitled as “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)” has been started. In this project, weaim at obtaining the resonance parameters precisely ofMAs by combining the neutron capture -ray measurement to transmission neutron measurement atKURRI-Linac. In this year, new detection system of neutron capture  rays and transmission neutron was constructed and its performance was confirmed.

Fig.1 Preliminary results of neutron capture cross sectionof 237Np

EXPERIMENTS: We constructed 4 bismuth germinate (BGO) scintillation detectors composed of 12 BGO
cylindrical crystals having 2 inch. in diameter and 2 inch.
in length for measurement of neutron capture  rays. On
the other hands, a 6 mm thick GS20 6Li-glass scintillator
was used as a transmission neutron detector.
In the performance test, we used a sealed MA sample of237
Np. Neptunium oxide powder of 1.13 g packed in analuminum disk container of 30 mm in diameter and 0.4
mm thick wall. The sample was placed in the geometricalcenter of the BGO detectors using an aluminum sample
folder. To decrease the constant background due to decay
 rays from radioactivity (26 MBq) of 237Np, a 3 mm
thick lead shield was inserted in the sample folder and
BGO detectors. A distance between the sample and theKURRI-LINAC neutron source was 10 m. To obtain theincident neutron spectrum, a 10B sample was also used.
Transmission neutron spectra with and without the 237Np
sample were measured by the 6Li-glass detector.
RESULTS: Preliminary result of neutron capture crosssection of 237Np is shown in Fig. 1. The relative crosssections were normalized to the evaluated value ofJENDL-4.0[4] at 0.0253 eV. A correction of neutron multiple-scattering and self-shielding in the sample has not
be made yet. Although the amount of measuring time isabout one hour, enough net counts were obtained in theresonance region. Comparison of transmission neutron

Fig.2 Comparison of transmission neutron spectra withand without 237Np sample by the 6Li-glass detector
Present study includes the result of “Research and Development for accuracy improvement of neutron nuclear
data on minor actinides” entrusted to the Japan Atomic
Energy Agency by the Ministry of Education, Culture,Sports, Science and Technology of Japan (MEXT).
REFERENCES:[1]M. Igashira et al., Nucl. Instrum. Meth., A 600, 332 (2009)..
[2]U. Abbondanno et al., Phys. Rev. Lett. 93, 161103 (2004).
[3]M. Jandel et al., Phys. Rev. C 78, 034609 (2008).
[4]K. Shibata et al., J. Nucl.Sci.Technol.,48, 1 (2011).

27017- 66 -CO2-2

Activation Measurements of Neptunium-237 at KURRI-Linac

S. Nakamura, K. Terada, Y. Shibahara1, A. Uehara1,T. Fujii1, T. Sanao1, Y. Takahashi1 and J. Hori11index was 0.053. The reaction rate of the 237Np sample
was calculated with 312-keV and 984-keV -ray yields,
detection efficiencies, decay data and -ray emission
probabilities. The reaction rate was (7.12±0.25) ×10-14(1/s). The experimental reaction rate was compared withthe calculated one with evaluated nuclear data of 237Np.
The neutron flux distribution at the irradiation position
was calculated by MVP2.0[2] with JENDL-4.0[3], andplotted in Fig.2. The calculated flux was normalizedwith the reaction rate of Au monitor. With the normalized
neutron flux and the cross-section data 237Np fromJENDL-4.0[3], the calculated reaction rate resulted in(7.41±0.14)×10-14 (1/s). The calculated reaction rate
was good agreement with the experimental one within thelimits of error.

Japan Atomic Energy Agency
Research Reactor Institute, Kyoto University

INTRODUCTION: A series of experiments with anactivation method has been performed to measure thermal-neutron capture cross-sections of Minor Actinides(MAs) under the project entitled by “Research and development for Accuracy Improvement of neutron nuclear
data on Minor ACtinides (AIMAC)”. Neptunium-237 is
one of important MAs because it contributes to thelong-term radiotoxicity of nuclear wastes. Then, the present work performed cross-section measurements of237
Np by an activation method with a neutron source
generated by the KURRI-Linac.
EXPERIMENTS: A standardized solution of 237Np(4 kBq) was pipetted on a quartz plate (10mm×15mm×2mmt), and dried by an infrared lamp. Then, the Np sample was put into an aluminum target holder together withCo and Au foils as neutron monitors. The target holder
was set near the moderator tank at the target room ofKURRI-Linac. The irradiation was performed for 14hours. The KURRI-Linac was operated under the condition: repetition rate 50Hz, pulse width 4s, 108A beam
current and 3-kW power.
After the irradiation, induced activities of the samples
were measured with a high purity Ge detector. Figure 1
shows an example of -ray spectrum of the irradiated237
Np sample.

Fig.2 Neutron flux distribution calculated by MVP Codes
CONCLUSION: Since the thermal-neutron flux was2
found to be 108 (n/cm s) order, activation measurements
would be possible under adequate experimental conditions: irradiation time, repetition rate, beam current andso on. The activation measurement of 237Np was performed. It was found that this measurement supported theevaluated data of JENDL-4.0.
The authors would like to thank staffs of Kyoto University Research Reactor Institute for their support.
Present study includes the result of “Research and Development for accuracy improvement of neutron nuclear
data on minor actinides” entrusted to the Japan Atomic
Energy Agency by the Ministry of Education, Culture,Sports, Science and Technology of Japan (MEXT).

Fig.1 -ray spectrum of the irradiated 237Np sample
RESULTS and DISCUSSION: Neutron flux components were derived from the induced activities of themonitors on the basis of the Westcott’s convention[1].
82
The thermal-neutron flux was 2.97×10 (n/cm s) at theirradiation position of 3-kW operation. The epi-thermal

REFERENCES:[1]C.H. Westcott et al., “Proc.2nd Int.Conf.Peaceful Useof Atomic Energy, Geneva”, United Nations, NewYork,16, p.70 (1958).
[2]Y. Nagaya et al., “MVP/GMVP II: General Purpose
Monte Carlo Codes for Neutron and Photon Transport
Calculations based on Continuous Energy and Mutigroup Methods”, JAERI 1348 (2005).
[3]K. Shibata et al., J.Nucl.Sci.Technol.,48, p.1 (2011).

27050- 67 -Development of the Epi-Thermal Neutron Measurement Method
Using a Thick Boron Sample

National Metrology Insitute of Japan, National Instituteof Advanced Industrial Science and Technology
1
Department of Engineering, Nagoya University
2
Research Reactor Institute, Kyoto University

INTRODUCTION: Evaluation of neutron fluence andneutron dose equivalent for the epi-thermal neutron region is very important in work places with neutron
sources or nuclear fuels as well as irradiation fields in aboron neutron capture therapy. The epi-thermal neutrons
are usually measured using a gold activation detector, agas proportional counter or a scintillator. However, it is
not easy to determine precisely the neutron fluence forthe epi-thermal neutrons in an irradiation field because oflarge uncertainty of reaction cross sections in theepi-thermal region. In the present study, we have developed an epi-thermal measurement method that is notaffected by nuclear reaction cross sections. We also developed an epi-thermal neutron camera consisting ofGEMs and resonance filters for neutrons up to 10 keV.
EXPERIMENTS: A collimated neutron beam wasobtained by the photo neutron reaction using a water-cooled tantalum target at the KURRI Linac [1].
Characteristics of a neutron detection system composedof a 6Li6natGd10B3O9:Ce+ (LGB) scintillator and anNaI(Tl) scintillator were experimentally evaluated bymeans of the time-of-flight (TOF) method. The 50mm-diameter and 5-mm thick LGB scintillator was set atthe center of the beam line. The 76.2 mm-diameter and76.2 mm thick NaI(Tl) was located out of neutron beamat an angle of 135 or 90 degrees with respect to the neutron beam direction. When the LGB scintillator detects
neutrons by the 10B(n,γα) reaction, 478 keV monoenergetic gamma rays are produced and subsequently detected with the NaI(Tl) scintillator. Moreover, the absolute
neutron fluence is determined by measuring gamma raysfrom the 10B(n,γα) reaction with the NaI(Tl) scintillator
in setting a 5-cm thick natB total absorption sample infront of the LGB scintillator. Number of alpha particles
and gamma rays produced by the 10B(n,γα) reaction in
the LGB scintillator are absolutely determined by thecoincidence measurements. Uncertainty caused by thecross sections of the 10B(n,) reaction is reduced by using the boron total absorption sample.
RESULTS: Fig.1(a) and (b) show ratio of the pulse
height spectra and the TOF spectra of the NaI(Tl) detector with and without the thick natB total absorption sample.
In Fig.1(a), 478 keV gamma rays due to the10B(n,) reaction is detected around 600 channel.
These figures indicate the relative ratio of countrate of the NaI(Tl) scintillator with and without
the boron total absorption sample. In the measurements, it was confirmed that the present detection system derived the neutron fluence in thethermal and epi-thermal region.The experimental
results will be basic data in order to develop a novel
portable epi-thermal neutron detector.
103

Counts/Channel

T. Matsumoto, A. Masuda, H. Tomita1, K. Ito1, Y.
Ichinose1, T. Iguchi1, K. Watanabe1, A. Uritani1, J. Hori2and Y. Sakurai2

without B sample
with B sample(a)1021011000200103(b)Counts/Channel

CO2-3102400600
PH Channel

800

1000

without B sample
with B sample
2s/channel

1011000500
1000
TOF Channel

1500

Fig.1. Ratio of (a) the pulse height spectra and (b) theTOF spectra of the NaI(Tl) detector with and without thethick natB total absorption sample.
REFERENCES:[1] K. Kobayashi et al., Annu. Rep. Res. Reactorinst. Kyoto Univ. 22, 142 (1989).

27052- 68 -CO2-4

Measurements of Americium Isotopes by Activation Method at KURRI-Linac
trum of the 243Am sample was measured for 20 hours,
and is shown in Fig.3.

S. Nakamura, K. Terada,
Y. Takahashi1, T. Sanao1and J. Hori1
1

Japan Atomic Energy Agency
Research Reactor Institute, Kyoto University

INTRODUCTION: Neutron capture cross section
measurements have been conducted for Minor Actinides(MAs) under the research project entitled by “Researchand development for Accuracy Improvement of neutron
nuclear data on Minor ACtinides (AIMAC)”.
Since americium-241 and -243 have relatively long
half-life as 432yr and 7370yr respectively, these isotopes
are important nuclides among MAs in terms of environment load reduction. This is why these two americium
isotopes were selected in the present measurements.

Fig.2 An example of -ray spectrum of the irradiated241
Am sample

EXPERIMENTS: Americium-241, 243 samples weresealed on Al plates with 22mm in diameter and 0.5mm inthickness, respectively. Their active areas were 20mm indiameter. The amounts of the samples were 583kBq for
241
Am, and 2MBq for 243Am, respectively. Each Am
sample was packed into an Al target holder. Then, setsof monitors of Au and Co foils were sealed at adequate
positions on the Al holders. Then, the target holders wereset near the moderator tank at the target room of KURRI-Linac.
Figure 1 shows the irradiation set-up of the 243Am target near the moderator tank at the target room of KURRI-Linac. The 241Am target was irradiated for 35 hours
under beam conditions: repetition rate 100pps, beamwidth 4s, beam current 220A, and 6-kW power operation, while the 243Am target was irradiated for 40 hours
under another beam conditions: repetition rate 50pps,
beam width 2.5s, current 67A, and 2-kW power operation.

Fig. 1

Set-up of the 243Am target at the target room ofKURRI-Linac

After the irradiations, induced activities of the samples
were measured with an -ray spectrometer, and a highpurity Ge detector. Figure 2 shows an example of -rayspectrum obtained by measurements of the irradiated241
Am sample for 11 days. On the other hand, the 243Am
sample was measured by the Ge detector. The -ray spec-Fig.3 An example of -ray spectrum of the irradiated243
Am sample
ANALYSIS and RESULTS: As for the 241Am experiment, the neutron flux was approximately 1.4×108(n/cm2s) on the basis of the Westcott’s convention. Fromthe peak counts of 241Am and 242Cm in Fig.2, the rate of241
Am(n,)242gAm reaction was obtained as (8.63±0.80)
×10-14(/s). As for the 243Am experiment, the neutron
flux was approximately 9.0×107(n/cm2s) at the target
position. The decay -rays of 744 and 898 keV of 244gAm(10.1h) were observed as shown in Fig.3. The target
amount of the 243Am sample was obtained with the yieldsof 228 keV -ray emitted from 239Np in equilibrium with243
Am. From the ratio of -ray yields of 744 keV of244g
Am and 228 keV of 239Np, the 243Am(n,)244gAm reaction rate was obtained as (1.14±0.03)×10-15(/s), which
only included statistical errors.
Furthermore, detailed analyses are underway for the241
Am and 243Am experiments carried out at KURRI-Linac.
The authors would like to thank staffs of Kyoto University Research Reactor Institute for their support.
Present study includes the result of “Research and Development for accuracy improvement of neutron nuclear
data on minor actinides” entrusted to the Japan Atomic
Energy Agency by the Ministry of Education, Culture,
Sports, Science and Technology of Japan (MEXT).

27081- 69 -