CO2-1 Development of Measurement Method for Epi-Thermal Neutrons Using the 10B(n,)Reaction
T. Matsumoto, A. Masuda, H. Harano, H. Tomita1, K. Ito1,
Y. Ichinose1, J. Kawarabayashi1, T. Iguchi1, K. Watanabe1,
A. Uritani1, J. Hori2, Y. Sakurai2

The detection system were experimentally tested using
the neutron beam from a neutron source with the photo
neutron reaction by means of the TOF method. Fig.1
shows the pulse height spectra of the NaI(Tl) scintillator
with and without the coincidence measurements. In the
NaI(Tl) scintillator, 478 keV gamma rays due to the10
B(n,) reaction is detected around 600 channel. The
counts due to other gamma rays by the coincidence
measurements. In the experimental room, many boron isused in shielding materials and collimators. Background measurements with only NaI(Tl) detector werealso performed.
Characteristics of a prototype of epithermal neutron imaging camera were also measured using the TOF method.
The prototype detector consists of a silver plate as a resonance filter, a B4C thermal neutron absorber, and a
GEM with a neutron converter of 10B.
RESULTS: From the coincidence measurement, countrates for the LGB detector (N)) and the NaI(Tl) detector(N) and coincidence count rate (N) are derived. Neutron capture reaction rate due to the 10B(n,) reaction isderived from NN/N. Neutron fluence was finally determined using ratio of count rates in measurements withand without the thick natB total absorption sample. The
experimental results are checked by Monte Carlo simulations in future.

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 dose equivalent for the epi-thermal neutron region is very importantin work places with neutron sources or nuclear fuels aswell as irradiation fields in a boron neutron capture therapy. A new calibration method for the response of neutron dosimeters has been developed using a pulse white
neutron beam[1]. On the other hand, it is not easy tomeasure precisely the neutron fluence for epi-thermal
neutrons in an irradiation field, although a gold activation
method is usually used in a thermal neutron region, andan elastic neutron scattering reaction with hydrogen atoms is adopted in a fast neutron region to measure precise neutron fluence. In the present study, we have developed a measurement method for epi-thermal neutrons
using a 6Li6natGd10B3O9:Ce+ (LGB) and an NaI(Tl) scintillators. We also developed an epi-thermal neutron camera consisting of GEMs and resonance filters for neutrons
up to 10 keV.
EXPERIMENTS: A collimated neutron beam throughcollimators was obtained by the photo neutron reaction
using a water-cooled tantalum target at the KURRI Linac[2]. The 50 mm-diameter and 5-mm thick LGB scintillator was set at the center of the neutron beam. The
76.2 mm-diameter and 76.2 mm thick NaI(Tl) was located out of neutron beam at an angle of 135 degrees withrespect to the neutron beam direction. When the LGB
scintillator detects neutrons by the 10B(n,) reaction,
478 keV monoenergetic gamma rays are produced andsubsequently detected with the NaI(Tl) scintillator. In
the coincidence measurements, the neutron capture reaction rate in the LGB scintillator is obtained without detection efficiencies of the LGB and the NaI(Tl) scintillators. Moreover, the absolute neutron fluence is determined by measuring gamma rays from the 10B(n,) reaction with the NaI(Tl) scintillator in setting a 5-cm thicknat
B total absorption sample in front of the LGB scintillator.

Counts/Channel

104
478 keV  ray

103

Net PH (N)

102
101
100

Coincidence PH
0

1000
2000
NaI(Tl) Pulse Height Channel

Fig.1. Pulse height spectra measured with the NaI(Tl)detector.
REFERENCES:[1] T. Masumoto et al., KEK proc. 2011 (8), 218-225(2012).
[2] K. Kobayashi et al., Annu. Rep. Res. Reactorinst. Kyoto Univ. 22, 142 (1989).

採択課題番号 26055 熱中性子フルエンス率の測定の高度化とその国際標準化に関する研究共同通常
通常採択（産総研）原野英樹、松本哲郎、増田明彦、（名大）井口哲夫、瓜谷章、河原林順、渡辺賢一、富田英生、伊藤海、一之瀬裕一郎、杉本大（京大・原子炉）堀順一、櫻井良憲
103CO2-2

Measurement of Gamma Ray and Neutron Spectrum of Curium Isotope (3)Y. Nauchi, J. Hori1, T. Sano1

rates are compared in Table 2. Considering the differenceof neutron spectra shown in Fig 3, the count rates in thetwo detectors agree fairly well (18% difference inCM244). Accordingly, absolute measurement of neutrons
from the sample would be capable for CM samples taking
the neutron spectrum into account.
REFERENCES:[1] T. Fujii et al., Conf. Record 2010 Annual Meeting of
AESJ pp58-66 (on CD-ROM).
[2] A. G. Seamster et al., Phys. Rev. C 29 (1984) 394
Table 1: Isotopic composition of Cm samples [1]Central Research Institute of Electric Power Industry
Research Reactor Institute, Kyoto University

1

244

Sample

245

Cm

246

Cm

Cm

CM244

88.53

3.13

8.34

CM246

24.18

1.11

62.18

247

Cm

248

3.00

Cm

9.53
CM244
CM246

gamma ray spectra (cps/MeV)

208Tl
100

4.82MeV102.235MeV
2.61MeV
2.98MeV 3.04MeV
1

3.04MeV
3.79MeV
2

2.5

3

3.5
Energy(MeV)

4

4.5

5

Fig. 1  ray spectrum measured for CM244 and CM246 .
gamma ray spectra (cps/MeV)

1000

CM244 behind Poly.
CM246 behind Poly.
CM244 No Poly

H(n,) Al(,)
100

10

1
2.15

2.2

2.25
Energy (MeV)

2.3

2.35

Fig. 2 H(n,) spectrum overlapped by Al(,) spectrum
Pulse height spectrum (cps/MeV)INTRODUCTION: Yields and spectra data of neutrons and higher energy (>3MeV)  rays from thespontaneous fission of 244Cm has been focused on since
they are major source in spent nuclear fuel. The radiations would be significant indicators of existence of nuclear materials. However those measured data are scarce.
KURRI have possessed Cm samples (Table 1 [1]) in
which Cm2O3 are diluted in aluminum (Al) powder encapsulated in Al cladding. In this work, preliminary
characterizations of CM244 and CM246 are done.
GAMMA RAY MEASUREMENTS:  rays from samples were measured with a Ge detector of 20%
effeiciency.  rays radiated subsequent to  decays were
measured. For CM244 and CM246 samples, multiple rays of different energies from 243, 244, 245Cm were wellmeasured. The activity ratio of CM244/CM246 are
0.935± 0.0191 and 0.714±0.004 for 244,245Cm, respectively. The ratios agree with the data in Table 1.
Then, higher energy  rays from 2 to 5MeV are focusedon to clarify whether noise spectra exist or not for measurements of 2.223MeV and fission  rays. As shown in
Fig. 1 and broader peaks from 2.5 to 5MeV are found forCM244. The bump spectra exist only slightly for CM246
so it is considered the fission  ray spectrum dominates
the energy region above 2.3MeV in CM246. The resolutions of the peaks are poorer than 2.6145MeV  ray from208
Tl. That indicates the  rays are radiated from compound nucleus moving with a certain kinetic energy morethan keV. Based on literatures, the 2.235MeV  rays
might be attributed to 27Al(,) reactions [2]. The originof the other higher energy  rays are unknown althoughneutron direct interactions with Ge detertor materials are
not the candidates since such bump spectra were notfound for a case when 252Cf was put alternatively to CM
samples.
The higher energy bumps should be eliminated to quantify fission  ray spectrum. For the purpose with the samples, coincidence measurements with multiple detectors
would be required as well as studies on nuclear physics to
explore the reactions radiating such  rays.
NEUTRON MEASUREMENTS: By putting CM samples and Californium (252Cf) neutron source behind polyethylene block of 10cm thickness, the H(n,) 2.223MeV 
rays were measured with the Ge detector. As shown in
Fig. 2, 2.235MeV  ray spectrum overlaps the 2.223MeV
one for CM samples. The former was subtracted from thelatter by normalizing the count rate of the former over a
region from 2.235 to 2.250MeV to that of latter. Then the
count rate of the 2.223MeV  rays was estimated. On theother hand, fast neutrons from the CM samples were
measured with a NE213 scintillator. The measured count

101

CM244
CM246
Cf252

100
10-1
10-2
10-3
0

1

2
3
4
5
6
Pulse height (MeV-electron equivalent)

78Fig. 3 Neutron spectrum measured with NE213.
Table 2: Count rate ratio of neutrons for CM / 252Cf
MethodCM244/252Cf

rel.err

CM246/252Cf

H(n,γ)

0.212

0.030

0.860

0.019

NE213

0.180

0.063

0.843

0.012

採択課題番号 26057 キュリウム核種のガンマ線と中性子発生スペクトルの測定(3)（電中研）名内泰志（京大・原子炉）堀順一、佐野忠史
104

rel.err

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CO2-3 Development of Hybrid Ce:LiCAF Scintillator for Nuclear Data Measurement
T. Fujiwara1,4, D. Matsuyama2, K. Tagi2, Y. Kusumawati2,
J. Hori3, T. Sano3, and M. Uesaka1

excellent performance for high energy neutron measurement. 8 m TOF line at KURRI.
In addition, low gamma-ray sensitivity great gamma/neutron discrimination with pulse height and showedvery quick recovery from gamma-burst which enableshigh energy neutron measurement.

1

Nuclear Professional School, The University of Tokyo
Department of Nuclear Engineering and Management,
The University of Tokyo
3
Research Reactor Institute, Kyoto University
4
National Metrology Institute of Japan, National Instituteof Advanced Industrial Science and Technology
2

INTRODUCTION: Precise nuclear data measurement
is needed for new generation reactor design and development. In this field, neutron Time-of-Flight (TOF)
measurement with pulsed neutron source is playing important role.
In this work, we report on the first nuclear data measurement at KURRI-Linac neutron source with hybrid
Ce:LiCAF scintillator which has high neutron detection
efficiency and better gamma-ray discrimination[1][2].
Ce:LiCAF covered with plastic scintillator[3] showed excellent characteristics in the detection efficiency andgamma-ray discrimination compared with other conventional scintillators, such as Li-Glass and Li:ZnS scintillators. Especially, excellent pulse height discrimination ofneutrons from gamma rays and very short decay time arevery promising for a new field of fast neutron detection
and great recovery from the gamma-flash.

Fig. 1. Pulsed height spectrum of Ce:LiCAF based
hybrid scintillation neutron detector. Showed clear neutron peak and low sensitivity to gamma-ray.

EXPERIMENTS: Neutron transmission experiments(NRTA: Neutron Resonance Transmission) were carriedout at 8m TOF line at KURRI-Linac. 237Np sample wasplaced at the neutron beam line, and the detector wereplace in 50 cm behind the sample.
Accelerator was operated in 50Hz and 100ns pulse
width. We have fabricated hybrid scintillator based on 10
×10 × 2mm size Ce:LiCAF scintillator crystal. The crystal were covered with plastic scintillator (PPO and
POPOP) and coupled to photo multiplier tube (Hamamtsu
R329-05). 5mm thick Pb filter was placed to reduce
gamma-rays. Fast Comtec MPA3 were used for DAQ
system.

Fig. 2. First result of neutron transmission experiment(NRTA: Neutron Resonance Transmission) of 237Np.
The measurement were carried out at 8m TOF line at
KURRI-Linac

RESULTS: Fig. 1 shows pulse height spectrum ofneutron detector. The line in green shows energy spectrum of neutron and gamma-ray mixed field, and the lingin pink represents a energy spectrum when neutron wereshielded with Boron. As shown in Fig. 1, clear neutron
peak and very low sensitivity to gamma-ray were observed, which stands for great gamma-ray discrimination.
This is due to coupling Ce:LiCAF scintillator with plastic
scintillator which enables to improve alpha/beta ratio.
With our detector, we successfully measured high energy
neutrons up to 2.4 keV. Fig. 2 shows the absorption neutron cross-section of 237Np.
We conclude that TOF spectrum of Ce:LiCAF showed

REFERENCES:[1] A. Yamazaki et al., Nucl. Instr. and Meth. A (2011),
doi:10.1016/j.nima.2011.02.064[2] K.Watanabe et al., Conference record of NSS IEEE,
201[3] T. Fujiwara et al., Neutron News, vol. 23, no. 4, pp.
31–34, Nov. 2012.

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V=5#7DJIWL/ * V=5#7W>6 5@F+ 2B
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105

CO2-4

Investigation of Variable Neutron Field for Accurate Determination of
Thermal Neutron Capture Cross Section at KURRI-LINAC

J. Hori, T. Yagi, T. Sano, Y. Takahashi, H. Yashima

ured TOF spectra and the evaluated neutron capture crosssections of JENDL-4.0 [2] as listed in Table 2. The calculated reaction rates were in agreement with the experimental ones within 2 % accuracy. As a result, it was confirmed that the neutron spectrum information obtained by
the TOF method can be used for activation analysis.
In nearly future, we will apply the spectrum determination method to the activation experiments for MA.

Research Reactor Institute, Kyoto University

In order to reduce radioactive
toxicity, feasibility for nuclear transmutation of minor
actinides (MAs) and long-lived fission products(LLFPs) by utilizing innovative nuclear system has
been investigated. The report on OECD/NEA No.6410[1] indicates that there are large gaps on capture crosssections of minor actinides (MAs) between current
uncertainty and required accuracy.
Since high-intensity-pulsed neutron sources withspallation reaction such as CERN, LANL and J-PARC
become available, statistical error of relative neutron
capture cross sections of MAs with time-of-flight(TOF) method was decreasing drastically. However,accurate normalization of TOF data is still an important issue, and unrecognized bias effect needs to beeliminated as much as possible. In this study, we proposed the thermal capture cross section measurement
methods using variable neutron fields and developed
variable neutron fields in the KURRI-LINAC.
INTRODUCTION:

Table 1 Boric-acid concentration of the light water moderator

EXPERIMENTS: In order to cross checked the neutron capture cross sections by integral measurements
using variable neutron flux field, KURRI-LINAC
based neuron source was utilized. Neutrons producedby photonuclear reactions were moderated by lightwater. Four kinds of variable neutron flux fields wereconstructed by merging boric-acid solution into thelight water moderator, as listed in Table. 1. Neutron
spectra of these fields were obtained by TOF measurement of the capture gamma-rays from the 10B(n,)reaction with twelve Bi4Ge3O12 (BGO) scintillators,
which located at 12.7±0.02 m from the photoneutron
source. The linac was operated with a repetition rate of
50 Hz, a pulse width of 100 ns, an averaged current
was 18 A and an electron energy of about 30 MeV.
Gold foils with and without Cd cover were also irradiated for each moderator. Reaction rates were obtainedwith an activation foil method.

Fig. 1 Neutron spectra for moderators with different
boric-acid concentration measured by TOF
methodTable 2 Comparison of normalized reaction rates of
197
Au(n,)198Au

Present study includes the results 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).

RESULTS: Figure 1 shows the neutron spectra formoderators with different boric-acid concentration measured by the TOF method. The thermal bumps were fittedwith the Maxwell distribution and the neutron temperatures were determined from 319.5 to 430.8 K. On theother hands, the reaction rates of 197Au(n,)198Au derived
from the activation method experimentally were compared with the calculated reaction rates using the meas-

REFERENCES:[1] M. Salvatores et al., WPEC-26, OECD NEA (2008).
[2] K. Shibata et al., J. Nucl. Sci. Technol. 48(1) (2011) 1-30.

採択課題番号 26069
長寿命放射性核種の中性子核データ精度向上に係る研究開発（京大・原子炉）堀 順一、佐野 忠史、高橋 佳之、八島 浩
106

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CO2-5Experimental Study on Non-destructive Assay with a Pulsed Neutron Source
RESULTS: The TOF spectra for the mixture composedof natU, 237Np and 243Am are shown in Figs. 1 and 2. Although many resonance dips from the impurities of 237Np
and 243Am were observed (Fig. 1), there are no differences between TOF spectra with only natU (blue line) andthe mixture (red line) (Fig. 2). This result indicates thatthe self-indication TOF spectrum was not greatly influenced by nuclide impurity. It was experimentally shownthat the contribution from the other nuclide can be remarkably suppressed by applying the self-indication
method.

J. Hori, T. Sano, Y. Takahashi, H. Unesaki
and K. Nakajima
Research Reactor Institute, Kyoto University
INTRODUCTION: Non-destructive assay of nuclear
fuel materials in spent nuclear fuel is a key technology
for nuclear material accountancy and critical safety. Neutron Resonance Densitometry (NRD) with a pulsed neutron source has been developed for a non-destructive assay of material [1]. In this method, we can identify andquantify the target nuclide by measuring the neutron resonance absorption with the Time-of-Flight (TOF) technique. There are two ways to measure the neutron resonance absorption. One is Neutron Resonance Transmission Analysis (NRTA) [2] and another is Neutron Resonance Capture Analysis (NRCA) [3, 4].
In this work, we have proposed a new concept of“self-indication method” combined with NRTA and
NRCA. In the self-indication method, we set an indicatorconsisted of target nuclide with a high purity at thebeam-downstream from a sample. By detecting the reaction products such as neutron capture  rays or fission
products from the indicator with the TOF method, thetransmission neutron can be measured indirectly. The
self-indicator is a transmission neutron detector which
has a high efficiency around the objective neutron resonance energies of target nuclide, so that it enables us toquantify effectively the amount of resonance absorptionof the target nuclide. Moreover, it is hard to be affectedby the decayed -rays from the fuel.

Fig. 1 TOF spectra obtained with NRTA for natU and mixtureof natU, 237Np and 243Am

EXPERIMENTS: We demonstrated a nondestructive
assay for nuclear material using a 46-MeV electron linear
accelerator at the Kyoto University Research Reactor
Institute. A mixture sample of a natural uranium plate andsealed minor actinides such as 237Np and 243Am was used
as a sample. The natural uranium plate was 10×20 mm2
and weighted 5.8 g. The samples of 237Np and 243Am
were oxide powder, which was pressed into a pellet 20mm in diameter and encapsulated in an aluminum
disk-shaped container 30 mm in diameter with
0.5-mm-thick walls. The activities of 237Np and 243Am
were 26 and 868 MBq, respectively. The 10B plug 8 mm
thick or a natural uranium plate of 10×20 mm2 andweighted 5.8 g was used as an indicator. Prompt-capture
-rays from the indicator were measured with a pair of
Bi4Ge3O12(BGO) detectors in the TOF experiment. Since
the 10B indicator can absorb most neutrons with energies
below the epithermal-region, it was equivalent to theconventional NRTA. In the latter case, it was theself-indication measurement.

Fig. 2 TOF spectra obtained with the self-indication method
for natU and mixture of natU, 237Np and 243Am

This work was supported by JSPS KAKENHI Grant
Number 24760714.
REFERENCES:[1] H. Postma et al., Radioanalytical and Nucl. Chemistry,
248, (2001) 115-120.
[2] J. W. Behrens et al., Nucl. Technol., 67, (1984)
162-168.
[3] Y. Kiyanagi et al., J. Nucl. Sci. Technol., 42, (2005)
263-266.
[4] A. Pietropaolo et al., Applied Spectroscopy, 64,(2010) 1068-1071.

採択課題番号 26070

パルス中性子源を用いた核燃料物質の非破壊定量及び可視化技術の開発（京大・原子炉）堀 順一、高橋 佳之、佐野 忠史、中島 健、宇根崎 博信
107

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