CO2-1

High-Precision Mass Analysis of RI Sample for Cross-Section Measurements

S. Nakamura, Y. Shibahara1, A. Uehara1, T. Fujii2,A. Kimura, B. Hales and O. Iwamoto
Japan Atomic Energy Agency
Kyoto University Research Reactor Institute2
Osaka University
1

INTRODUCTION: The ImPACT project[1] aims to realizea "large reduction of and exploitation of resources in high-level
nuclear waste by nuclear transmutation", by using an accelerator
to accomplish nuclear transmutation of long-lived fission products (LLFPs). To perform accelerator-based nuclear transmutation effectively, it is necessary to acquire reaction cross-section
data for radionuclides across a wide spectrum of incident particle
energies. This project targets the major LLFPs: 107Pd, 93Zr, 135Cs,
126
Sn, and 79Se, as well as incidentally the medium-lived fission
product 137Cs. We targeted the 135Cs because of its long half-life(2.3×106 yr[2]). However, it is impossible to obtain a pure 135Cs
sample because it is not supplied by available vendors. This isalso why the quantity of data is so scarce [3, 4]. Isotope separation of 135Cs and 137Cs is difficult, and therefore 135Cs is included
as an impurity in available 137Cs samples. The isotope abundance of 135Cs and 137Cs in a standard 137Cs source has beenmeasured by mass spectrometry. Then, the amount of 135Cs canbe obtained from yields of 662 keV decay gamma-rays of 137Cs
and the isotope abundance. It is also necessary to suppress contamination of the mass spectrometer from radioactive 137Cs and135Cs. It is necessary to perform highly precise analysis for anextremely small amount of sample. Therefore, this research aims
to apply mass spectrometry to a very small amount of radioisotope sample, and to confirm its effectiveness.
EXPERIMENTS: A standard solution of 137Cs (370kBq)
was obtained through the Japan Radioisotope Association. An
appropriate amount of the 137Cs solution was dispensed, and thenconverted to nitrate by a chemical processing. The mass spectrometer TRITON was used for this analysis, which was madeby Thermo Fisher Scientific. A small amount of the chemicallyprocessed solution, about 10Bq, was pipetted onto a Re filament
with a TaO activator, and then dried. Three filaments with applied Cs solution were prepared. The filaments were attached
onto the ion source of the mass spectrometer.
RESULTS AND DISCUSION: Figure 1 shows the resulting mass spectrum of the standard 137Cs sample. From the yieldsof the mass spectrum, the obtained 135Cs/137Cs isotope ratio was
0.868±0.004 (±2σ). Before the chemical processing, false peaksof about 1% were observed as the part of red line, which would

28054

come from sources other than Cs or molecular ion. The falsepeaks appear at Cs masses plus 0.1u. On the other hand, it isknown that peaks from Ba appear on the almost same mass.
Since the false peaks decayed faster than Cs peaks with theelapse of time, it is supposed that the false peaks were due tosome kind of chemical impurities. In the case of the chemicallyprocessed sample, false peaks did not appear as shown with theblack line in Fig.1. From this, it is thought that false peaks were
due to some kind of chemical impurities in the Cs sample because the false peaks became extinct after chemical processing.
The results of the isotope ratios in both cases were in agreementwith each other within the limits of errors.

Fig. 1 Measured mass spectrum of 137Cs sample
CONCLUSION: This work successfully measured the isotope ratio of 135Cs and 137Cs with an accuracy of 0.5% using astandard solution of 137Cs. This technique makes it possible toanalyze a very small amount (picogram order) of radioactive
sample. When the Kyoto University Reactor is operated again innear future, neutron capture cross-section measurements of 135Cs
shall be performed by activation method with the 137Cs solution
analyzed in this work.
AcknowledgementThe authors would like to thank the staff of Kyoto University
Research Reactor Institute for their support.
This research was funded by ImPACT Program of Council
for Science, Technology and Innovation (Cabinet Office, Government of Japan).
REFERENCES:[1] http://www.jst.go.jp/impact/index.html[2]R.B. Firestone et al., Table of Isotopes, 8th ed. Update, John
Wiley and Sons, New York, (1998).
[3]V.A. Anufriev et al., Soviet Atomic Energy, 63, 851(1988).
[4]T. Katoh et al., J.Nucl.Sci.Technol., 34, 431(1997).

CO2-2

Measurement of Doppler Effect by Small Accelerator Neutron Source (1)T. Sano, J. Hori, Y. Takahashi, K. Nakajima, H. Unesaki
Research Reactor Institute, Kyoto University

INTRODUCTION: In order to reduce TRU, the research and technology development entitled as “TRUburning fast reactor cycle using uranium-free TRU metalfuel” have been started Japan at October 2014 [1]. The
feature of the fast reactor is high content TRU and Zr
without uranium in the fuel alloy so that additional TRU
is not produced. On the other hand, uranium-free TRU
metallic fuel leads to the reduction of the Doppler reactivity. Thus, the utilization of fuel alloy such as Mo and
Nb instead of Zr is considered as one of the countermeasures [2]. As the Doppler effects depend on the magnitude of self-shielding at the resonances, it is importantto verify the Doppler effects at each resonance of the fuelalloy (Mo or Nb) materials to evaluate the feasibility of
the uranium-free TRU metallic fuel. However, the differential experiment of the Doppler effects for Mo andNb has not been carried out so far. Therefore, we haveinitiated the measurement of the Doppler effects for Mo
sample by Time-of-flight (TOF) method with the KURRI-LINAC pulsed neutron source.
EXPERIMENTS: We measured the Doppler effects ofMo sample with the KURRI-LINAC pulsed-neutron
source. In the experiment, neutron capture rates in Mo
sample at 300 K and 600 K were obtained by prompt
gamma-ray measurement with the TOF method. The Mo
sample was placed in the center of a heating device at a
distance of 10 m from the Ta target. The surface temperature of the sample was observed by thermoelectric couple and controlled to be constant during the irradiation bya glass-heater. Two kinds of natural molybdenum samples
with different thickness of 0.5 mmt and 3.0 mmt wereprepared to identify the neutron self-shielding effects foreach resonance. Those were metallic plates of 2.0×2.0
cm2. The measurements with thick and thin Mo samples,the graphite sample with 3.0 mm in thickness, and nosample (blank run) at 300 K and 600 K were carried out.
RESULTS: The measured TOF spectra with 0.5 mmt
and 3.0 mmt Mo samples at 300K are shown in Fig. 1.
The main resonances of Mo-95, Mo-96, Mo-97 andMo-98 were clearly observed in the energy range from 10eV to 1 keV. For further analysis, counting statistics observed in TOF spectrum were improved by rebinning thedata into the JSF70 group structure [5] corresponding tothe lethargy width of 0.25. Comparison of the rebinned
TOF spectra for the 3.0 mmt sample at 300 K and 600 K
is shown in Fig. 1. The Doppler capture rate ratio definedas the capture rate at 600 K divided by it at 300 K wasobtained for each energy group. The Doppler capture rate
ratio were 1.3 ± 0.7 % at 44.9 eV, 8.0 ± 0.7 % at 70.9 eV,
1.0 ± 0.7% at 131.4 eV and 6.5 ± 3.0% at 358.6 eV resonances.
44.9 eV : 1.3±0.7%

70.9 eV : 8.0±0.7%131.4 eV : 1.0±0.7%
358.6 eV: 6.5±3.0%

Fig. 2 the experimental TOF spectra for 3.0 mmt Mo sample at 300 K
and 600 K

This study is being performed under the contract with
the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT) in the framework ofMEXT’s Nuclear System Research and Development
Program.
REFERENCES:[1] K. Arie, et. al., Proc. of Global 2015, Sep. 20-24, Paris, France, Paper 5096, (2015).
[2] K. Arie, et. al., Proc. of ICAPP 2014, Charlotte, USA,April 6-9, (2014).
[3] Y. Nagaya et. al., JAERI1348, Japan Atomic Energy
Research Institute, (2005).
[4] K. Shibata, et. al., J. Nucl. Sci. Technol. 48(1), 1-30
(2011).
[5] M.Nakagawa, K. Tsuchihashi, JAERI 1294 (1984)

Fig. 1 The measured TOF spectra with Mo samples, graphite sample and no sample

28061

CO2-3 Development of Neutron Detectors for Precise Measurement of Epi-thermal Neutrons
T. Matsumoto, A. Masuda, H. Harano, H. Tomita1, Y.
Ichinose1, K. Uema1, T. Iguchi1, K. Watanabe1, A. Uritani1 and J. Hori2
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 fluenceand neutron 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 a boron neutron capture
therapy (BNCT). It is not easy to determine precisely the neutron fluence for the epi-thermal
neutrons in an irradiation field because of large
uncertainty of reaction cross sections in theepi-thermal region. In the present study, we
have developed neutron detectors for absolutemeasurement of epi-thermal neutrons, high intensity epi-thermal neutrons in a BNCT field andan imaging detector for a hidden material.
EXPERIMENTS:
A collimated neutron beam
was obtained by the photo-neutron reaction using
a water-cooled tantalum target at the KURRI
Linac [1]. We experimentally evaluated characteristics of the epi-thermal neutron detector for
absolute measurement, the high intensity
epi-thermal neutron detector and the imaging
detector.
(1) Epi-thermal neutron detector for absolutemeasurement
The epi-thermal neutron detector for absolute measurement is composed of a 6Li6natGd10B3O9:Ce+ (LGB)
scintillation detector and two BGO scintillation
detectors. In our previous experiments, we used
an NaI(Tl) scintillation detector instead of theBGO scintillation detectors. However, the NaI(Tl)scintillator has large sensitivity to neutrons as acomparison with the BGO scintillator. The 50mm-diameter and 5-mm thick LGB scintillator
was set at the center of the beam line. The 50.8mm-diameter and 50.8 mm thick BGO scintillators were located on both sides of the LGB scintillator and out of neutron beam. When the LGB
scintillator detects neutrons by the 10B(n,γα) reaction, 478 keV monoenergetic gamma rays areproduced and subsequently detected with theBGO scintillators. Moreover, the absolute neutron
fluence is determined by measuring gamma raysfrom the 10B(n,γα) reaction with the BGO scintillators in setting a 1-cm thick 10B4C total absorption sample in front of the LGB scintillator.

28062

Number of alpha particles and gamma rays produced by the 10B(n,) reaction in the LGB scintillator are absolutely determined by the coincidence measurements. Pulse height spectra of theLGB and BGO scintillators from the 10B(n,) reaction were obtained in the coincidence measurements. However, the evaluation of the background due to scattered neutrons was not sufficient. In future, we will improve the measurement process based on the present experimentalresults.
(2) High intensity epi-thermal neutron detector
The high intensity epi-thermal neutron detector
is important in the BNCT. The detector is simply composed of a 6Li-glass scintillator and a
7Li-glass scintillator and photomultipliers (PMTs).
Current from an anode output of the PMT is detected by a current integrator. The 7Li-glass
scintillator is used to subtract the gamma raycontribution detected by the 6Li-glass scintillator.
We experimentally evaluated the difference between the neutron detection efficiencies of the6Li-glass and 7Li-glass scintillators using theLinac neutron source. Finally, we successfully
obtained the relation between the thermal neutron flux and output current using the neutron
detection efficiencies in the thermal neutron
standard field of the National Institute of Advanced Industrial Science and Technology. In
future, we will perform the high flux neutron irradiation using the KUR.
(3) Imaging detector
The imaging detector is used to search hidden
materials such as nuclear materials and radioactive sources. The imaging detector is composedof multi-pixel type CdTe detectors. The imaging
detector detects prompt gamma rays produced bythe neutron capture reaction in the hidden materials or gamma rays from a hidden radioactive
source. After that, the imaging detector determines a position of the hidden material or thehidden radioactive source. Characteristics of thedetector were evaluated using gamma rays fromthe neutron capture reactions of 10B and 197Au.
We will continue the experiments in order to verify potential of the practical realization using theTOF method.
REFERENCE:[1] K. Kobayashi et al., Annu. Rep. Res. Reactorinst. Kyoto Univ. 22, 142 (1989).
This work was supported by JSPSKENHI(JP16K21679,JP16K09030).

KA-CO2-4

Research and Development for Accracy Improvement of Neutron Nuclear Data on
Long-lived Radioactive Nnuclei at KURRI-Linac

J. Hori, T. Sano, Y. Takahashi and H.Yashima1
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 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 [1]. In this project,
we aim at obtaining the resonance parameters preciselyof MAs by combining the neutron capture -ray measurement to transmission neutron measurement atKURRI-Linac. Neptunium-237 is one of the most important MAs with a long half-life. In this year, neutron
total cross section and capture cross section of 237Np
were measured and resonance analysis was also performed.
EXPERIMENTS: The total and capture cross section
measurements have been performed by the neutron
time-of-flight (TOF) method using the 46-MeV electronlinear accelerator at the Research Reactor Institute, Kyoto
University (KURRI-LINAC). Transmitted neutrons weredetected by a 6 mm thick GS20 6Li-glass scintillator.
Neptunium oxide powder of 1.13 g packed in an aluminum disk container of 30mm in diameter and 0.4 mm
thick wall, which was placed at a distance of 10.15 mfrom the photo-neutron source. An aluminum disk container without the neptunium oxide powder was also used
as a dummy case. The linac was operated with an electron energy of about 30 MeV, an averaged beam currentof 17 A, a repetition rate of 50 Hz, a pulse width of 0.1
s. Neutron capture gamma rays from the sample were
measured with a 4 bismuth germinate (BGO) scintillation detectors composed of 12 BGO cylindrical crystals
having 2 inch. in diameter and 2 inch. in length. A capture sample was set in the center of the detector. Threesamples of 237Np with different thickness were used; theactivities of the samples were 26, 5.2 and 1 MBq, respectively. The incident neutron flux shape was measured
with a 10B sample.

SAMMY code [ 2] in the energy range from 10-2 to 80 eV.
The resonance parameters of the 0.49-eV resonance weremodified to reproduce the present results. The neutron
capture yields for the samples with different thickness
were also obtained as shown in Fig. 2. The neutron
self-shielding and multiple-scattering effect in the sample
depends on the resonance parameter. Therefore, the data
were used for the cross-check of resonance parameters.

Fig. 1 Preliminary results of neutron total cross section of237
Np

Fig. 2 Capture yields for three 237Np samples measured by
the BGO spectrometer
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:RESULTS: Preliminary result of total cross section of237
Np is shown in Fig. 1. The resonance analysis of thededuced cross section values were performed with the28076[1]H. Harada et al., EPJ Web of Conferences 93, 06001 (2015).
[2]N. W. Larson, ORNL/TM-9179/R6, Oak Ridge National
Laboratory, 1998.