CO3-1

Development of In-Reactor Observation System Using Cherenkov Light (IV)

N. Kimura, T. Takeuchi, A. Shibata, N. Takemoto,
A. Kimura, M. Naka, K. Nishikata, M. Tanimoto,
K. Tsuchiya, T. Sano1, H. Unesaki1, Y. Fujihara1,
K. Okumura1 and K. Nakajima1
Neutron Irradiation and Testing Reactor Center, JAEA
Research Reactor Institute, Kyoto University

1

INTRODUCTION: In research reactors, CCD cameras are used to observe reactor core for reactor operation
management, e.g. to prevent debris from falling. In order
to measure the reactor power and fuel burnup exactly bymeans of observation of Cherenkov light [1], the development of the on-line measurement device started in2009. In this study, the wavelength and the absolute irradiance of the Cherenkov light were measured by a spectrometer, and the Cherenkov light was observed by the
CCD camera.
EXPERIMENTS: The in-reactor surveillance system
was composed of the spectroscopy system and a visible
imaging system. The spectroscopy system (Maya2000,
OptoSirius Corp.) was able to measure the wavelength oflight in the range from 200 nm to 1100 nm. The visible
imaging system was composed of the CCD camera(AEC-100ZL, Q・I Inc.), a monitor recorder, and a controller. The latter system was used to observe the Cherenkov light with changing the aperture value. For
in-reactor surveillance, the spectrometer and the CCD
camera were inserted into core-observation pipe of KUR.
The Cherenkov light was measured when the reactor
power steadily increased from 0.2 MW to 1 MW, and was
at 5MW. The ND-filter and the aperture were changed
during the observation in order to confirm their performance for the Cherenkov light.

Fig. 1. The absolute irradiance of the Cherenkov light
as a function of wavelength.

RESULTS: The absolute irradiance of the Cherenkov
light as a function of wavelength is shown in fig. 1. The
result shows that the detected Cherenkov light, which has
a maximum value of absolute irradiance at 400 nm, distributes from 380 to 700 nm in wavelength independentof the reactor power. The relationship between the nominal transmittance value of ND-filter and the measurement
value is shown in fig.2. The result shows the measurement value by the spectroscopy system was well accorded with the nominal transmittance value of ND-filters at
the reactor power of 0.5, 1, and 5MW. As the result ofobservation of Cherenkov light by the visible image system, the images were fogged by halation during highpower operation. However, the amount of incident light
could be controlled by changing the ND-filters, and halation could be removed from those images.
CONCLUSION: The Cherenkov light from KUR core
was measured with the developed in-reactor surveillancesystem. The correlation between absolute irradiance of
the Cherenkov light and the nominal transmittance valueof ND-filters was evaluated by the spectroscope system.
As a result, the measurement value is good agreementwith the nominal value. On the other hand, the value by
the visible imaging system was obtained the same tendency of nominal transmittance value of ND-filters.
These results suggest that a possibility of a device to obtain the reactor power and fuel burnup information byobserving the Cherenkov light.
REFERENCES:[1] J.V. Jellry, Cherenkov Radiation and its Applications(Pergamon, New York, 1958).
[2] N. Takemoto, K. Tsuchiya, et al., KURRI Progress
Report, (2010) P. 204.

Fig. 2. Characteristics of the ratio of transmitted lightof the Cherenkov light by ND-filter.

採択課題番号 24038 チェレンコフ光を用いた試験研究炉の炉内監視手法の研究開発
共同通常（原子力機構）土谷邦彦、谷本政隆、那珂道裕、竹本紀之、柴田 晃、武内伴照、木村明博、
西方香緒里、木村伸明（京大・原子炉）中島 健、宇根崎博信、佐野忠史、藤原靖幸、奥村 清

CO3-2

Evaluation of Activation Products for Decomissioning

T. Kosako, T. Iimoto, K. Tani, Y. Fujimichi, K. Kumagai,
T. Kosaka, T. Watanabe and K. Yano
Graduate School of Engineering, The University of TokyoINTRODUCTION: In the case of decommissioning,radioactive inventory is used as a basic data for planningof decommissioning, occupational exposure estimation,
estimation of radioactive waste and public safety assessment. Then, evaluation of radioactive inventory is necessary information to carry out decommissioning safely andreasonably. According to previous studies [1], accuracy
improvement of BSW concrete (the largest amount ofmaterial in NPP) radioactive inventory evaluation has
been already considered. But, for the second largest
amount of material, metal, accuracy improvement is insufficient. This study is focused on improving accuracyof radioactive inventory evaluation of metals, considering
the difference of neutron flux [2].

EXPERIMENTS: Before experiment, activation calculation was conducted with two elemental composition
data; JAERI-M [3] and chemical analysis date of sample.

In neutron irradiation experiment, samples are same
standard metals used in practical commercial reactor. But
the neutron spectrum in slant is different from a typical
LWR. In order to reproduce the neutron spectrum in acommercial LWR, an optimal moderator should be selected. The transmission spectrum is calculated by using
MCNP-5 (Monte Carlo Simulation Code). Adjustment of
neutron spectrum is shown in Fig.1. For experiment, Cd
is selected as moderator. After irradiation experiment,
sample is measured by Ge semiconductor detector.

RESULTS: The result of SUS304 is shown in Table 1.
Measurements and calculated values showed a good
agreement, but there is a difference of 103 times between
calculation data and measured data in 60Co.
Table 1.

Experiment and calculation data of SUS304
Calculated data [Bq/g]
Measured data [Bq/g](RSD)

NuclideJAERI-M 6928 chemical analysis

51

Cr

9.93E+03

9.12E+03

5.49E+03(±2%)

Co

6.88E+02

7.87E+02

3.86E+03(±4%)

Co

2.04E-01

1.02E+02

1.04E+02(±9%)

Mn

9.06E+01

9.77E+01

5.17E+01(±12%)5.56E+06

5.51E+06

3.76E+06(±1%)

1.28E+02

1.42E+02

9.80E+01(±15%)

58

60

54

56

Mn

59

Fe

REFERENCES:[1] T. Ogawa, et al., Radiation Protection Dosimetry
146(1-3), p.356-9 (2011)
Fig. 1. Neutron energy spectrum of inner surface of
biological shield concrete.

[2] N. Kawata, et al., JNST, Vol. 9, No. 4, p. 405-418(2010)[3] K. Koyama, et al., JAERI-M 6928 (1977)

採択課題番24070
放射性核種生成量評価のための中性子による材料照射の研究
一般通常（東大院・工）小佐古敏荘、飯本武志、谷幸太郎、藤通有希、熊谷一城、小坂晃義、渡邉貴裕、矢埜孝三

Performance Evaluation of Radionuclide Monitoring Systems
during Reactor Operation

Hitachi, Ltd., Hitachi Research Laboratory
1
Research Reactor Institute, Kyoto University
INTRODUCTION:
Examinations of two kinds of radionuclide monitoring
systems using coincidence counting were conducted in
the Kyoto University Research Reactor (KUR). One uses
a reactor water monitor and the other uses a 60Co monitor
for the primary loop recirculation (PLR) piping.
It is important to detect fuel rod failure early in orderto improve safety of nuclear reactors. One of themeasures for early detection is changing from sampling
surveys to continuous monitoring, which is defined as
online monitoring. Some isotopes of radioiodine can beused as index radionuclides to ensure fuel integrity.
In addition, in order to reduce dose exposure caused by
60
Co which is the main radionuclide encountered byworkers during scheduled outages of boiling water reactors, water chemistry control during reactor operation iscarried out[1]. It is necessary to monitor the deposited60
Co on the inner surfaces of PLR piping to evaluate effects of water chemistry control.
However, many short-half-life radionuclides like 13N,
16
N and 19O present in the reactor water interfere with thedetection of gamma rays emitted by the iodine isotopes
and 60Co. 132I, 134I and 60Co have the characteristic ofemitting gamma rays in a cascade. By using a coincidence method, the gamma rays emitted by these radionuclides can be measured with a high signal-to-noise ratio
because most interfering radionuclides emit gamma rays
independently. Thus, monitoring systems based on the
coincidence counting method were developed.
EXPERIMENTS:(1) Reactor water monitor
Two LaBr3(Ce) scintillation detectors were set up near
the ion exchanger in the KUR. A standard mixed source
was also set up 2.5 to 3.0 cm from the detectors. The
source included two radionuclides emitting gamma rays
in a cascade, 60Co (1012 Bq) and 88Y (430 Bq). The air
dose rate around the detectors was estimated with apocket dosimeter and counting rates were as high as 40µSv/h at the time of the experiments. The gamma rays
were detected and their pulse-height and time information
was recorded with the measurement system. The information was used to measure the cascade radionuclides by
the coincidence method in an offline process.

containment vessel (PCV). The air dose rate during reactor operation was about 4.8mSv/h. Experimental measurement instruments using the coincidence counting
method were set up in the cold area.
3. RESULTS:(1) Reactor water monitor
As shown in Fig. 1, due to accidental coincidence, net
coincidence counting rates decreased with increasing air
dose rate. By converting 60Co measurement data to 132I
and 134I for the reactor water conditions, it was found that
more than 100 min were needed to detect fuel failure
with the device and a signal-to-noise ratio more than 10times larger was needed to satisfy the measurement limit
defined by regulation. Thus, the measurement efficiencyof the system must be enhanced by 10 times.
(2) 60Co monitor for PLR piping
The transition of coincidence counts rate for 60Co isshown in Fig. 2. The dotted line was calculated based on
the geometry of the two detectors and the 60Co source,
the detection efficiency of the gamma rays and the 60Co
source activity. It was confirmed that the coincidence
counts for 60Co could be detected for the dose rate of 4.8
mSv/h with a standard deviation of 19%.
0.12
Co-60

0.1

Y-88
0.08
0.06
0.04
0.02
0
0

10

20
30
40
Air dose rate [μSv/h]50

60

Fig. 1 Net coincidence counting rate.
Coincidence counts rate
[counts/80min.]K. Okada, K. Ueno, T. Tadokoro and K. Takamiya1

Coincedence counting rate[cps]

CO3-3

140
120
100
80
60
40
20
0

Measured
Calculated

0

2

4

6

Measurement time [hr]Fig. 2 The transition of coincidence counts for 60Co.

60(2) Co monitor for PLR piping
Two LaBr3(Ce) scintillation detectors and a 60Co
source, which was used to simulate the 60Co deposited on
the inner surfaces of PLR piping, were set up in the heatexchanger room of the KUR; this arrangement could
simulate the radiation environment inside of a primary

REFERENCE:[1] “Research Report Related to the Trend of the Water
Chemistry of the Domestic BWR Plants”, Japan Nuclear
Energy Safety Organization, Sep. 2012, pp.10-15 (inJapanese).

採択課題番号 24093 高バックグラウンド条件における Co-60 及び原子炉水中核種の検出 共同即時
（京大・原子炉）高宮 幸一（日立）田所 孝広、上野 克宜、岡田 耕一

CO3-4 Neutronic Characteristics of Lead in KUCA A core for Accelerator-Driven System
T. Sugawara, K. Nishihara, H. Iwamoto, T. Yagi1 and C.
Pyeon1
Japan Atomic Energy Agency
1
Research Reactor Institute, Kyoto University

EXPERIMENTS: The experiment was carried out inthe KUCA A core (EE1) shown in Fig. 1. A special assembly which included Al or Pb plates instead of polyethylene was placed at the center of the core, そ-15. For
the special assembly, five different loading patterns were
employed by changing the number of Al and Pb plates asshown in Fig. 2. Case Al60 was the reference case andthe difference of excess reactivities between Al60 andother cases were calculated as a replacement reactivity.
Case Pb40C was the case where 40 Pb cells were placedin the center of the fuel assembly and case Pb40E was thecase where 40 Pb cells were put in the edge of the fuelassembly 20-and-20.
Figure 3 presents the experimental results for each case.
As the number of the Pb plate increased, the replacement
reactivity from Al to Pb also increased. The replacement
reactivity was 20.8, 44.7 and 48.6 pcm in the case of
Pb40C, Pb40E and Pb60, respectively.
12

13

14

15

16

17

18

C3

F

F

F

F

F

S5

れ

F

F

F

F

F

そ

F

F

F'

F

F

F

F

F

F

F

F

F

F

F

10

F

F

よ
た

つ

S4

ね
な

F
C2

ら

Fig. 1.

F

Fuel

F'

Fuel with Pb or Al plate
Polyethylene reflector

C1

CControl rod

S

Safety rod

S6Core arrangement in the KUCA-A core (EE1)

Fig. 2.
Replacement reactivity (Al > Pb) [pcm]INTRODUCTION:
An Accelerator-Driven System(ADS) has been investigated in Japan Atomic Energy
Agency (JAEA) to transmute minor actinides discharged
from nuclear power plants. The ADS proposed by JAEA
is a lead bismuth eutectic (LBE) cooled-tank-type ADS.
It has been known that there was a major upgrade for the
cross section data of lead isotopes from JENDL-3.3 to
JENDL-4.0 and the upgrade affects to the neutronic design of the ADS [1, 2]. For instance, keff value calculated
by JENDL-3.3 was 0.97 and the value calculated by
JENDL-4.0 was 1.00 and the main cause of this difference was the cross section data of the lead isotopes[2].
This study aims to measure replacement reactivity
from aluminum plates to lead plates to validate the nuclear data of the lead isotopes.

Fuel cell for each case
150

Experiment
MCNPX-J40
MCNPX-J33 (All)SRAC-CITATION (J40)
SRAC-CITATION (J33)

100

50

0

Pb20

Pb40C

Pb40E

Pb60

Case

Fig. 3. Comparison of experimental results and calculation results
CALCULATIONS: MCNPX and SRAC-CITATION
codes were used for the calculation with JENDL-4.0 and
JENDL-3.3 libraries. The keff value was calculated foreach experimental case and the difference of the keff value
was calculated as the replacement reactivity.
Figure 3 also shows the calculation results for eachcase. For the MCNPX calculation, although all calculation results were larger than the experimental ones, theresults calculated with JENDL-4.0 were close to the experimental ones. The same tendency was observed in the
SRAC-CITATION calculation. It was guessed that thenuclear data of the lead isotopes in JENDL-4.0 was morereasonable than those in JENDL-3.3. On the other hand,
it is necessary to investigate the large discrepancy between the experiment and calculation results.
CONCLUSION: The replacement reactivity from thealuminum plates to the lead ones was measured and thecalculation results indicated that the lead nuclear data in
JENDL-4.0 might be more reasonable than those in
JENDL-3.3.
REFERENCES:[1] T. Sugawara et al., Atom Indonesia, 38, 2, (2012)
71-77.
[2] H. Iwamoto, et al., Proc. ND2013, New York, Mar.
4-8, (2013)

共同通常
採択課題番号 CA24101 KUCA-A 架台を用いた加速器駆動炉における鉛の核特性評価（日本原子力研究開発機構）菅原隆徳、西原健司、岩元大樹（京大・原子炉）卞哲浩、八木貴弘

Development of Activation Foil Neutron Detector for
Angular Distribution Measurements of Epi-Thrermal Neutrons in Reactors

K. Watanabe, S. Maruyama, A. Uritani, A. Yamazaki,
Y. Kawabata, C. Pyeon1 and T. Misawa1
Graduate School of Engineering, Nagoya University
1
Research Reactor Institute, Kyoto University
INTRODUCTION:
For basic design andperformance predictions of radiation shielding,numerical simulations are quite important. Although
the accuracy of the numerical simulations on theneutron transportation is reliable, basic experiments
still play an important role especially for neutron deep
penetration problems. The validity of the numericalsimulations has been confirmed through comparisonswith the integral experiments, because it is difficult toexperimentally obtain differential information, such as
neutron energy and directions. So far, we havedeveloped the measurement system of the angular
distribution of thermal neutrons. This system cansuccessfully observe the neutron streaming effectbetween the solid moderators of the Kyoto University
Critical Assembly (KUCA). As the next step, wedevelop a new directional detector for epi-thermal
neutrons. This system applies an activation foil withresonance neutron absorption in the epi-thermalenergy region as a neutron sensitive element. In thisreport, we demonstrate the direction distribution
measurements performed at the KUCA.
ACTIVATIONFOIL
DETECTORFORANGULAR DISTRIBUTION MEASUREMENTS:
As the activation foil for detection of the epi-thermal
neutrons, we adopt a silver foil. Silver-109 with theisotopic abundance of 48% has a large resonance
Plastic Scintillatorplastic（NE102A）
scintillator (NE102A)
IrradiationPart
part
Irradiation

capsuleCapsulecompression
Compression

slit
Slit

Pneumatic
pneumatic tube
tube
Cd sealed
Cd
Ag collimatorAg
collimator

aspiration
Aspiration
magnetic valve

Magnetic Valve

PMT (H6612)
PMT(H6612)

Ag foil

Ag foilabsorption peak at 5.2 eV. Figure 1 shows theschematic drawing of the proposed angular
distribution measurement system for the epi-thermal
neutrons.
We fabricated the prototype detector
system.
The fabricated system consists of an
irradiation capsule covered with a silver activation foil,
an irradiation part with a silver collimator for
confining the incident direction of epi-thermal
neutrons, a plastic scintillator for detection of particles emitted from the activated silver foil and apneumatic carrier system for the activation capsuletransportation. In order to measure the directional
distribution, the irradiation capsule is cyclicallytransported between inside and outside a reactor with apneumatic system and the irradiation part can rotate.
We demonstrated the directional distribution
measurements of epi-thermal neutrons at the solid
moderator core (A core) in the KUCA. The plastic
scintillator for -ray detection was placed outside thereactor and surrounded with a lead shielding tosuppress background radiations. Figure 2 shows the
directional distribution of the detected counts by the
fabricated detector system. The irradiation part wasplaced on the center line of the core and shifted toright side. The directional responses corresponding
to the detector positions were confirmed to beobserved. We concluded that the fabricated detector
system can acquire the information of the directional
distribution of epi-thermal neutrons.
0.032

Normarized count (a.u.)

CO3-5

Center pos.
Right shift pos.

0.03
0.028
0.026
0.024
0.022
-180

-120

-60

0

60

120

180

Angle of collimater opning direction (deg)Ag capsuleAg
capsuleFig. 1

Schematic drawing of the angular
distribution measurement system for
epi-thermal neutrons.

採択課題番号 CA24103

Fig. 2

Directional distribution of the detected
counts by the fabricated detector. The
irradiation part was placed on the centerline of the core and shifted to right side.

性子場特性評価を目的とした新型中性子検出器の開発に関する研究（名大・工）渡辺賢一、瓜谷章、山﨑淳、丸山秀典、川端勇矢、牛田雅人（京大・原子炉）三澤毅、卞哲浩

共同通常

CO3-6 Development of Subcriticality Measurement for Accelerator-Driven Reactor (VII)K. Hashimoto, W. Sugiyama, A. Sakon1, M. A. Binmaarof1, M. Kawasaki1, C. H. Pyeon2, T. Sano2, T. Yagi2
and T. Misawa2
Atomic Energy Research Institute, Kinki University
Interdisciplinary Graduate School of Science and Technology, Kinki University
2
Research Reactor Institute, Kyoto University
1

INTRODUCTION: A subcritical reactor system driven
by pulsed spallation neutrons generated from tungsten or
lead-bismuth target has been constructed in A-loading
facility of the KUCA and a series of cross-power spectral
analysis between time-sequence signal data from twoneutron detectors has been performed to develop the
methodology of on-line subcriticality monitoring for future accelerator-driven system. The preliminary results ofthe cross-power spectral analyses are showed in this report.
EXPERIMENTS: The present analyses were carried outin a reactor system referred to as A1/8”P60EU-EU. The
tungsten or lead-bismuth target was placed outside polyethylene reflector and the spallation neutrons were generated through the injection of 100MeV protons onto thetungsten or lead-bismuth target. As a pulse repetition
frequency (period), 20Hz (50ms) was employed. The
signals from three fission counters (FCs) for reactor
startup operation were fed to a fast Fourier transformer toanalyze the cross-power spectral density and to record thesignals as digital data. An analysis range in frequency
from 1.25 to 1000Hz was specified to obtain 800-point
spectral data from the transformer. From the spectral
density, the prompt-neutron decay constant was determined. As a subcritical rod pattern for this experiment,
we employed a shutdown pattern where all control andsafety rods were completely inserted.
RESULTS: Figure 1 shows a measured cross-power
spectral density between two fission counters FC1 andFC3. We can observe many delta-function-like peaks andthis feature is similar with that observed in a previous
analysis for 14MeV source [1]. However, no continuous
reactor-noise component can be seen from this figure.
This is because these fission counters are located far fromthe core and consequently the efficiency is extremely
low.
Assuming the one-point kinetics model, the top pointsof the uncorrelated peaks can be described as (   m )( )  A0,
(1)
 02   2
m 1



where

 m  2m /TR .


採択課題番号
CA21404(2)

In the above equations,  is angular frequency, 0 the
prompt-neutron decay constant of fundamental mode andTR pulse repetition period. A0 is a constant dependent on
detection efficiency, generation time, and accelerator parameters. The least-squares fits of equation (1) to thepeak point data for tungsten and lead-bismuth target leadto the prompt-neutron decay constants of 1076.8±33.9
and 1060.2±32.5[s-1], respectively. These fits underestimate the decay constant, compared with 1226.1±5.3[s-1]obtained by a pulsed neutron experiment. Figure 1 also
shows a significant deviation of the fitted curve from thepeak points in the higher frequency range over 500Hz.
Considering the contribution of a spatially higher
mode, the top points of the uncorrelated peaks can be
rewritten as (   m ) (   m )( )  A0

A
.
(3)
1
2
2



12   2
0
m 1
m 1





A further least-squares fit of the above equation (3) to thepoint data of delta-function-like peaks is shown in Fig.2,
where the deviation observed above disappears. The
least-squares fits of equation (3) for tungsten andlead-bismuth target lead to the consistent prompt-neutron
decay constants of 1208.9±50.2 and 1206.3±60.2[s-1],
respectively.
Spallation Source / Be-W Target

1E-3

Power Supply Noise

20Hz 40Hz



1E-5

FC1,FC3
2[V ]Fitted Curve = 1076.8  33.9


1E-7

1E-9

10

100

1000

Frequency f [Hz]Fig.1 Fit to peak points of equation (1) to uncorrelated
peak points
Spallation Source / Be-W Target

1E-3

20Hz 40Hz



1E-5

FC1,FC3
2[V ]Fitted Curve = 1208.9  50.2


1E-7

1E-9

10

100

1000

Frequency f [Hz]Fig.2 Fit to peak points of equation (3) to uncorrelated
peak points
REFERENCES:[1] A. Sakon et al., J. Nucl. Sci. Technol., 50 (2013)
481-492.

加速器駆動未臨界炉における共同通常
未臨界度測定高度化のための基礎実験（Ⅶ）（近大・原研）橋本 憲吾、杉山 亘（近大院・理工）左近 敦士、ムハマド・アイマン・ビンマー
ロフ、河崎 将大（京大・原子炉）卞 哲浩、佐野 忠史、八木 貴宏、三澤 毅

Active Gamma Ray Spectroscopy of Sub-Critical System Containing Stainless SteelCentral Research Institute of Electric Power Industry
1
Research Reactor Institute, Kyoto University

INTRODUCTION: In units 1 - 3 of the Fukushima
Daiichi power station, it is anticipated that molten fuelhas been mixed with reactor structural materials, which
act as a neutron absorber. In order to certify the criticality
margin brought by the mixture of the structural materials,
reaction rates of the neutron capture by them should bequantified in relation to fission reaction rates. For the
purpose, sub-critical experiments using uranium – stainless steel mixed cores have been launched for the firsttime.
EXPERIMENTS: SUS-304 (Fe, Ni (8-10.5%), Cr(18-20%)) is a typical structural material. SUS-304 plates(600mm x 62mm x 1.5mm) were prepared together with
the U-Al alloy fuel plates (93% 235U enrichment) and 3
types of core were mocked up as shown in Fig. 1. Massratios of SUS-304 to Uranium in the cores are 0, 15.4,
and 46.3, respectively. The subcritical multiplication factors of the cores are 0.86, 0.71, and 0.52, respectively,
when a 252Cf neutron source are loaded at the center ofthe cores. Thermal neutron flux distributions on the outersurface of the cores were measured with the 6Li scintillators. Gamma ray spectra were measured with a BGO
scintillator at 46 cm outside from the cores.
star: 252Cf, gray: assembly frame, green: U-Al fuel, black: SUS-304

increases, the count rate in the 6-10MeV region increases
and that in the 3-5MeV region decreases. By coupledneutron-photon transport calculations, the detection efficiencies of the gamma rays were deduced. Using the detection efficiency and the measured count rates, gamma
ray emission rates were deduced. The ratio of gamma ray
emitted in 6-10MeV to 3-5MeV energy regions are plotted against the calculated C/F in Fig. 4. The gamma ray
emission ratio monotonously increases with C/F. Accordingly, it was confirmed that the neutron capture rate by
the structural materials can be quantified in relation to thefission rate based on the active gamma ray spectroscopy.
dot: measurement

2.5

2.0 reactor
edge SUS-304: 0 plates
1.5
SUS-304: 45 plates

1.0
0.5

SUS-304: 90 plates
0.0

-14

SUS-304:90 plates
Fuel: 90 plates

Fig.1 SUS-304 loaded cores.
RESULTS: The measured neutron flux distributions
agree well with the calculations by the MCNP-5 codewith the nuclear data libraries based on the JENDL-4 asshown in Fig. 2. This means that the neutronics calculations have enough accuracy even in the sub-critical systems containing stainless steel (SUS-304).We calculated
the reaction rates of the neutron capture by SUS-304 andfission of 235U. Then the capture to fission ratio (C/F) was
deduced for each core.
The gamma ray pulse height spectra for thethree cores measured with the BGO scintillator are shownin Fig. 3. The gamma rays from capture reactions of 56Fe,
27
Al, and 58Ni are observed in 6-10MeV region and continuous spectra mainly caused by the fission reactions arefound in 3-5MeV region. As the SUS-304 loading ratio

-10

-8

-6

horizontal position (cm)

-4

-2

0

104

SUS: 0 plates
SUS: 45 plate
SUS: 90 plates

3

10

102

56Fe
58Ni

1

10

Mainly fission
27Al

100
2

4
6
8
pulse height (MeV)

10

12

Fig.3 gamma ray pulse height spectra for sub-criticalcores measured by BGO scintillator.
(6-10MeV)/(3-5MeV) γ ray emission ratio

SUS-304: 45 plates
Fuel: 135 plates-12

Fig.2 Thermal flux distribution outside cores

0

SUS-304: 0 plates
Fuel: 180 platesline: calculation
Assembly center

6Li(n,t) reactionrate (a.u)

Y. Nauchi, H. Ohta, H.Unesaki1, T. Sano1 and T. Yagi1

pulse height spectrum (/s/MeV)

CO3-7

1.4
1.2
1
0.8
0.6
0.4
0.2
00

0.2
0.4
0.6
0.8
1
calculated SUS304 capture / 235U fission

1.2

Fig.4 (6-10MeV) / (3-5MeV) gamma ray emission ratio
against calculated capture (SUS-304) / fission (235U) one.

共同通常
採択課題番号 CA24106 γ線計測を利用した未臨界度指標の計測に関する研究(2)（電中研） 名内泰志、太田宏一、（京大・原子炉）宇根崎博信、佐野忠史、八木貴宏

Subcriticality Measurement Experiment Using Inherent Neutron Source

INTRODUCTION: The Feynman-α method is a practical subcriticality measurement technique based on reactor
noise analysis, since it can be carried out simply with an
ordinary neutron detector system [1]. It is noted that thesufficient number of detected neutrons is required to accurately measure reactor noise. Thus, an external neutron
source, e.g. Am-Be or Cf, is used to increase detected
neutron counts. However, it is often unfavorable to perform measurement using an external neutron source in
actual nuclear fuel facilities. Therefore, the subcriticality
measurement using the inherent neutron source in nuclear
fuels, such as spontaneous fission and (α,n) reaction, wasdemonstrated in the previous study [2]. The purpose ofpresent study is to clarify the intensity of the inherent
neutron source and the measurement time required for thenoise analysis.
METHODOLOGY: In the Feynman-α method, the timeseries data of detected neutrons C(T) within a counting
gate width T is measured. Then, the neutron correlation
factor Y(T) is evaluated as follows:
Y (T ) ≡ σ 2 (T ) μ (T ) − 1 ,(1)2where μ (T) and σ (T) is the mean and the variance ofC(T), respectively. As shown in Eq. (1), Y(T) means adifference of “variance-to-mean ratio of C(T)” from unity,
thus Y(T) = 1 if the frequency distribution of C(T) is subjected to the Poisson distribution. On the basis of the one
point reactor kinetic equation, theoretical expressions ofY(T) and subcriticality (−ρ) can be described as follows:⎛ 1 − e −αT ⎞(− ρ ) = ⎛⎜ α − 1⎞⎟, (3)⎟⎟ , (2);
Y(T ) = Y∞ ⎜⎜1 −αTβ eff ⎜⎝ α 0 ⎟⎠⎝⎠where Y∞ is the saturation value of Y(T); α and α0 is theprompt neutron decay constant for target and critical systems, respectively; and βeff is the effective delayed neutron fraction. Using the Feynman-α method, α can beobtained by the least square fitting of Eq. (2) to measured
Y(T). Finally, subcriticality in dollar unit, (−ρ)/βeff, can beobtained by substituting α into Eq. (3), when the value of
α0 is estimated by another measurement or numericalanalysis.
EXPERIMENTS: The reactor noise experiments werecarried out in A3/8”P36EU-NU(3) core without external
neutron source. Four subcritical states, which areachieved by different patterns of six control rods andwithdrawing fuel and reflector assemblies in shutdown,
were measured. Four 3He detectors were placed in outerreflector region to collect time series data of detected
neutron counts. In the fuel plate used in KUCA, the inherent neutron sources mainly consist of spontaneous
fission of 238U and (α,n) reaction of 27Al. The inherent
neutron source intensity is ~1000 [neutrons/sec/core],

100

100

10

10
Y value

Graduate School of Engineering, Nagoya University
1
Research Reactor Institute, Kyoto University

which is estimated by the PHITS code [3] withTENDL-2011 alpha sub libraries. In order to verify theexperiments only with inherent source, we also carriedout the reactor noise experiments with external neutron
source (Cf source in the center of core). In addition, thereference values of (−ρ) were evaluated by the control
rod worth and the excess reactivity, which were obtainedby the rod drop and the period methods, respectively.
RESULTS: The values of Y(T) were analyzed with thebunching method [4] using the time series data (1 and 10min) measured only with the inherent source (Fig. 1).
Figure 1 shows the saturation value Y∞ is larger and thesaturation time is longer, as the value of (−ρ) close to
zero. Figure 2 shows α values obtained by the leastsquare fitting of Eq. (2) to measured Y(T). As shown in
Fig. 2, the 1- minute measurement only with inherent
source is not sufficient to accurately measure α value,
since the results of Y(T) have relatively large statisticalerrors as shown in Fig. 1. By taking longer time series
data, the α values obtained from 10-minutes time series
data is sufficient in this experiment. Hence, it is confirmed that the α value can be measured by a reactor
noise measurement only with the inherent neutron source,
if the measurement time is sufficiently long.

1

0.1

1 rod in3 rods in
6 rods in

1

Shutdown

0.1

0.01
1e-4

1e-3
1e-2
1e-1
Time gate width T [s]

0.01
1e-4

(a) 10 min

1e-3
1e-2
1e-1
Time gate width T [s](b) 1 minFig. 1 Variations of Y(T) for 4 subcritical states
1500
α value by fitting [1/s]

T. Endo, T. Buma, T. Shiozawa, C. H. Pyeon andT. YagiY valueCO3-8*

inherent source, 10 min
inherent source, 1minreference (Cf, 1 hour)

1000

500

0
0

1

2

3
4
(-ρ ) [%Δk/k]

5

6

Fig. 2 α values plotted for (−ρ)*(−ρ) in shutdown state was estimated by Eq.(3) with α0

REFERENCES:[1] D. G. Cacuci, Handbook of Nuclear Engineering:Vol.
3 Reactor Analysis, Springer, 1644-1647 (2010).
[2] K. Tonoike et al., J. Nucl. Sci. Technol., 41, 172-182(2004).
[3] K. Nitta et al., JAEA-Data/Code 2010-022 (2010).
[4] T. Misawa et al., Nucl. Sci. Eng., 104, 53-65 (1991).

採択課題番号 CA2417
固有中性子源を利用した未臨界度測定実験（名大・工）遠藤知弘、武馬孝仁、塩澤武司（京大炉）卞哲浩、八木貴宏

共同通常