CO3-1

Development of In-reactor Observation System Using Cherenkov Light (Ⅲ)

N. Kimura, N. Takemoto, H. Nagata, 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
1
Research Reactor Institute, Kyoto University
INTRODUCTION: In research reactors, a CCD camera is used to observe reactor core for reactor operation
management, e.g. to prevent debris falling. In order tomeasure the reactor power and fuel burnup exactly bymeans of observation of Cherenkov light [1] and gamma
ray information, the development of the on-line measurement device has been started since 2009. In this study,
as a part of development of in-reactor surveillance,
wavelength and absolute irradiance of Cherenkov light
were measured by spectroscopy, and Cherenkov light wasobserved by the CCD camera. The neutral density filters(ND-filters) [2] were used to investigate the transmittance of Cherenkov light in these measurements.
EXPERIMENTS: The diagrammatical drawing of
in-reactor surveillance system is shown in Fig.1. The
system is composed of the spectroscopy system and thevisible image system. The spectroscopy system (Maya2000, OptoSirius) is able to measure the wavelength oflight in the range from 200nm to 1100nm. The visible
image system is composed of the CCD camera(AEC-100ZL, Q・I), monitor recorder and controller. The
system is used to observe Cherenkov light with changing
the aperture. For in-reactor surveillance, the spectroscope
detector and the CCD camera were inserted in core observation pipe of KUR. Cherenkov light was measured
while the reactor power was increased from 20kW to
1MW steadily, and when the reactor power was at 1MWand 5MW. The ND-filters were installed into the detector
and the CCD camera to investigate transparency of Cherenkov light. The change of the ND-filter and the aperture
were carried out during the observation of Cherenkov

Fig. 1. Cherenkov light measurement apparatus.

light. Six kinds of ND-filters were used such as, 50.0,
25.0, 12.5, 6.3, 3.1 and 1.6% of nominal values.
RESULTS: The range of the wavelengths of Cherenkov light was from 380 to 700nm, and maximum value ofabsolute irradiance was recorded at 400nm. Therefore,
the wavelength of 400nm of Cherenkov light was focusedon, and the effect of ND-filters in absolute irradiance wasinvestigated. The relationship of transmittance between
the nominal value and measurement value of ND-filters
is shown in Fig. 2. This means the measurement value by
the spectroscopy system was well accorded with thenominal value of ND-filters at the reactor power of 0.5, 1,
and 5MW. As the result of observation of Cherenkov
light by the visible image system, similar transparency
tendency was also derived under low reactor power.
However, the image was fogged by halation when thereactor power was 1MW and 5MW. Halation could be
removed from those images by changing the aperture of
the CCD camera.
CONCLUSION: The measurement of Cherenkov light
was carried out with the simple in-reactor surveillancesystem. As a result, correlation between absolute irradiance of Cherenkov light and the nominal value of
ND-filters was evaluated by the spectroscope system, andthe measurement value was well accorded with the nominal value. As for the observation result by the visible
image system, similar transparency tendency to the spectroscopy system was derived. Therefore, it was found that
there is a possibility to develop a device to obtain thereactor power and fuel burnup information exactly. From
now on, correlation between absolute irradiance of Cherenkov light and gamma ray information will be analyzed
in detail.
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. Relationship of transmittance between the nominal
value and the measured value of ND-filters.

採択課題番号 23070 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、木村明博、永田 寛、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

Measurements of Angular Distribution of Epi-thermal Neutrons in Reactors Using theCompact Directional Neutron Detector

K. Watanabe, S. Maruyama, A. Uritani, A. Yamazaki,
Y. Kawabata, C.H. Pyon1 and T. Misawa1
Graduate School of Engineering, Nagoya University
1
Research Reactor Institute, Kyoto University
INTRODUCTION:
For basic design andperformance predictions of nuclear reactors, numericalsimulations are quite important. The accuracy of thenumerical simulations on the neutron transportation
has been quite high and reliable, but basic experiments
still play an important role. The validity of thenumerical simulations has been confirmed throughcomparisons with the integral experiments, because it
is difficult to experimentally obtain differential
information, such as neutron energy and directions.
We, therefore, have developed the measurement
technique of the neutron angular distribution. So far,
we have developed the measurement system of theangular distribution of thermal neutrons. This system
can successfully observe the neutron streaming effectbetween the solid moderators of the Kyoto University
Critical Assembly (KUCA). As the next step, we
have attempted to measure the angular distribution of
epi-thermal neutrons. We have proposed a new
directional detector for epi-thermal neutrons. We
adopt an activation foil with the neutron resonanceabsorption in the epi-thermal energy region as the
epi-thermal neutron detection element.
DIRECTIONAL EPI-THERMAL NEUTRONDETECTOR: As the activation foil for detection of
the epi-thermal neutrons, we adopt a Ag foil.
Silver-109 with the isotopic abundance of 48% has alarge resonance absorption peak at 5.2 eV. Since thePlastic Scintillator（NE102A）
plastic scintillator(NE102A)

Irradiation
part
IrradiationPart

CapsulecapsuleAgcollimator
Ag collimator

aspiration
magnetic valve

PMT (H6612)

PMT(H6612)

Ag
Agfoilfoil

102
101
100
1500

AgAgcapsule
capsuleFig. 1The schematic drawing of the angular
distribution measurement system for the
epi-thermal neutrons.

No sealed
Cd sealed
Cd+Ag sealed

103

compression

slit
Slit

Pneumatic
pneumatic tube
tube
CdCd sealed

maximum energy of β-rays of radioactive nuclide
110
Ag, which is produced by neutron activation from109
Ag, is relatively high (2.89 MeV), the emitted
β-rays can be detected using a plastic scintillator easily
and effectively. The half life of 110Ag is relatively
short (24.6 sec). In order to measure the foil
cyclically, the irradiation capsule covered with a Ag
foil is transported between inside and outside a reactor
with a pneumatic carrier system. Figure 1 shows theschematic drawing of the angular distribution
measurement system for the epi-thermal neutrons.
We fabricated a prototype detector system without therotating collimator.
The fabricated system wasoperated at the KUCA (Kyoto University Critical
Assembly) to check the fundamental operation of thesystem. The plastic scintillator for -ray detection
was placed outside the reactor.
The activation
capsule was transported between inside and outside
the reactor by the pneumatic carrier system. As thesignal output from the -ray detector, we counted thenumber of events with higher signal amplitude than agiven discrimination level. The temporal change ofthe  ray counts obtained by the plastic scintillator isshown in Fig. 2. When the irradiation part wascovered by a Cd sheet, the signal intensity
dramatically decreased. The foil activation without
the Cd cover was confirmed to mainly arise fromthermal neutrons. When the irradiation part was
surrounded by both the Cd and Ag sheets, the signal
intensity additionally decreased.
This reduction
component corresponds to the activation by
epi-thermal neutrons.

count

CO3-2

Fig. 2

2000
time[second]

The temporal change of the  ray
counts obtained by the plastic
scintillator.

採択課題番号 CA23101 原子炉内の中性子特性評価を目的とした
共同通常
光ファイバ型中性子検出器の開発に関する研究（名大・工）瓜谷章、渡辺賢一、山﨑淳、丸山秀典、（京大・原子炉）三澤毅、卞哲浩CO3-3

Measurement of Neutron Generation Time by Pulsed Neutron Source

Japan Atomic Energy Agency
1
Research Reactor Institute, Kyoto University
INTRODUCTION: The maximum likelihood analysis
method for pulse neutron source experiment (PNS-ML)
has been developed in experiments using KUCA [1] [2]to monitor sub-criticality of an accelerator-driven system(ADS). In this technique, the most likely sub-criticality
can be deduced from a series of counted data from a detector. In the present study, both sub-criticality andneutron generation time are deduced from the same
counted data.
REGRESSION METHOD: The next formula is derived from one-point dynamic equation and a periodic
boundary condition.
1
1
1 𝑓𝑒 −𝛼𝑡 +
�,� 1 − 𝑒 −𝛼⁄𝑓𝜌� − 1 𝜌�Λ𝑛(𝑡) = 𝑆̂ �𝛼=−𝜌� − 1�ΛHere, 𝑛(𝑡) is a time evolution of counting rate from a� = Λ/βeff a
detector, 𝑆̂ a variable for normalization, Λneutron generation time divided by effective delayed
neutron, f a pulse repetition frequency of DT neutron
source, ρ� = 𝜌/βeff a reactivity, and βeff effective a ratio
of delayed neutrons.
We fitted this formula to experimental results and ob� and ρ�tained the most likely values and errors of 𝑆̂, Λconsidering statistic error of the counters.

the PNS-ML method remains consistency. Another observation is longer neutron generation time for the deeper
sub-critical core.
CONCLUSION: The measurement of neutron generation time by pulsed neutron source was demonstrated byusing different frequent pulses. The PNS-ML method
showed excellent consistency comparing to the conventional method. The comparison of neutron generation
time with the calculation is the future work.
REFERENCE:[1] KURRI Prog. Rep. 2007, CO3-4 (2007).
[2] KURRI Prog. Rep. 2009, CO3-6 (2009).

Table 1

CR positionkeff (rodworth)

1

(1 +

𝑁𝑝

𝑁𝑑

) , where 𝑁𝑝 and 𝑁𝑑 is

area of prompt and delayed neutron, respectively.

EXPERIMENTS: A BF3 counter was set beside the
KUCA-A core with a DT neutron source. Three subcritical cores with different positions of the control andsafety rods were employed (Table 1). Period of the DT
neutron source was varied from 10 to 100 msec.
� of each core with
Theoretically, the obtained ρ� and Λ
different periods should be equal, although the shape ofcounting ratio is different. The purpose of experiment is
to check this equality in the case of conventional method
and PNS-ML method.
RESULTS:
Figure 1 shows results of sub-criticality
deduced by conventional and PNS-ML method from thesame experimental data. The three sub-criticalities ofeach core are equal in the PNS-ML method, while slightdependency on the pulse frequency is observed in theconventional method. This tendency is emphasized in
the neutron generation time in Fig. 2. The result of theconventional method is not usable, although the result of

SC3
All CRs&
SRs full in

0.985

0.959

0.996

SC1

SC2

-2
Conventional
PNS-ML

-3
-4
-5

SC3

-6
10

� is obtained by combining
Therefore, Λ
𝛼10, 40, 100
BF3SC2
CR1 to 3
full in-1

Fig.1.
5.55

Λ / β (msec)

�=two methods as, ΛCR1 full in

0

CONVENTIONAL METHOD: There are two analysis methods applicable for the PNS experiments: alpha-fitting method and area method. The alpha value
� is deduced by the first one, andthat is a function of ρ� and Λρ� by the second one.

Conditions of the PNS experimentDT pulse period (msec)
Counters
Core Index
SC1

ρ / β ($)

K. Nishihara, T. Sugawara, H. Iwamoto, C.H. Pyeon1 andT. Yagi140
DT-pulse period (msec)

100

Experimental result (sub-criticality)
Conventional
PNS-ML

SC3

4.5
SC2
4
SC1
3.5
3
10

Fig.2.

40
DT-pulse period (msec)

100

Experimental result (neutron generation time)

採択課題番号 CA23103
パルス中性子源を用いた ADS のための基礎実験（日本原子力研究開発機構）西原健司 菅原隆徳 岩元大樹（京大・原子炉）卞哲浩

共同通常

CO3-4 Measurements of Thorium-Fueled Core Characteristics and Replacement Worth of Thorium
Plates to Aluminum Plates (1)T. Kitada, T. Fujii, K. Wada, Vu Thanh Mai, N. Takaki1,
A. Yamaguchi1, H. Watanabe1 and H. Unesaki2
Graduate School of Engineering, Osaka University
1
Department of Nuclear Engineering, Tokai University,
2
Kyoto University Research Reactor Institute
INTRODUCTION: Thorium is widely known as agood candidate of Uranium, and is considered as a next
generation fuel. However the accuracy of the cross section is quietly lower than that of Uranium because thenumber of experiments with Thorium is not so many andalso validation study is not enough. This low accuracy
brings large uncertainty in calculation results of Thorium-fueled core performance such as conversion ratio.
The object of this experiment is to validate the Thorium
cross section through the comparison of calculation andmeasured results for Thorium-fueled core characteristics
and replacement worth of Thorium plates to aluminum
plates.
EXPERIMENTS: The assembled core is the first core
to be critical with unit fuel cell named “7/8PETEETEE”
which is shown in Fig. 1.
E
poly
U

Th

poly

E
E
poly
poly
U
U

Th

poly

E
E
poly
poly
U
U

232

Th/ 235U=15.2
H/235U=約140
poly = 1/8inch , EU plate = 1/16inch , Th plate = 1/8inch

Fig.1. Unit Fuel Cell(7/8”PETEETEE).
In the experiment, critical approach was performed. The
number of fuel assembly and control rod arrangements at
critical were analyzed before the experiment so as tomatch the constraints (maximum excess reactivity andcontrol rod worth), and each step from initial core to critical core was also analyzed with the comparison of Uranium-fueled core and the Thorium-fueled core performed
with other unit fuel cells. The core arrangements at eachstep are illustrated in Fig. 2.
8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23
よ
FC#2
FC#1
た
れ
C3
S5
そ
N
つ
②
ね
S4
C1
な
1 1 1 1 1 1 1ら①
3 3 2 2 2 3 3むC2 5 4 4 4 5 S6
う
FC#3
UIC#6
7 6 8
ゐ
の
お
UIC#4
く

B7/8"PETEETEE fuelReflector (polyethylene)
N

Neutron source

UIC

UIC detector

FC

Fission chamber

C1～C3

Control rod

S4～S6

Safety rod

Fig. 2. Core arrangements at each step from initial core to
critical core (B7/8”P17ETEETEE).
The number marked in fuel assemblies in Fig.2 shows thefuel assemblies added at each step of critical approach.
Initial core was assembled by 29 fuel assemblies, and at
next step fuel assemblies marked as “1” in Fig. 2 wereadded to measure count rate by fission chambers and 3He
detector marked as “①” and “②”. The number of fuelassemblies was slightly increased, and the critical core
was realized at last step:8 with 51 fuel assemblies. The
inverse count rate ratio curve obtained through the critical approach is shown in Fig. 3 for the case of all control
rods withdrawn(j=3).

j=3

1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0

BF3#1
BF3#2

FC#1
FC#2
FC#3
29

31

33

35

37

39

41

43

45

47

49

51

Fig.3. Inverse count rate ratio curve obtained forB7/8”P17ETEETEE core.
The excess reactivity was measured by the period method
and the value is about 0.15%dk/k, and the control rod
worth for C1, C2 and C3 measured by rod drop method
are 0.306, 0.323 and 0.321%dk/k, respectively. These
values satisfy their constraints to operate the core.
The core shown in Fig.2 was slightly changed to low
excess reactivity state so as to measure the replacement
worth. Thorium plates were replaced to aluminum plates
to measure the replacement worth by the change in excess reactivity. The position of the replacement was thecenter fuel assembly named な 15 and the number of thereplacement was two or four at around the center in coreheight. This measurement was done to check the C1 rod
position at critical with different replacement situation,
because the calibration curve of C1 rod obtained by fitting many experimental results (period method) was confirmed to have enough accuracy to check the excess reactivity. The results are summarized in Table 1, and theresults are roughly good agreement in the analysis with
continuous energy Monte-Carlo analysis with JENDL4.0.
Table1. Results of excess reactivity
at each replacement situation.
core configuration
partial+Al4
partial+Al2
partialexcess reactivity[Δ k/k]
2.44E-03
1.43E-03
4.35E-04

error
1.11E-05
3.59E-06
3.06E-06

RESULTS: Thorium-fueled core which has never assembled at KUCA was assembled. Criticality approachand related measurements were successfully performed.
The detailed analysis of the results is in progress.

採択課題番号 CA23104トリウム燃料炉心の炉心特性およびトリウム置換反応度測定(その 1)共同通常（阪大院工）北田孝典、藤井貴志、和田憲拓、Vu Thanh Mai (東海大工)高木直行、渡辺大貴、
山口晃範（京大・原子炉）宇根崎博信

CO3-5Experiments on Subcriticality Measurement Using
Neutron Source Multiplication Method Based on Deteced-Neutron Multiplication Factor

INTRODUCTION: The neutron source multiplication
method (NSM) is one of the practical subcriticality
measurement techniques, since NSM does not require
any special equipment other than a stationary external
neutron source and an ordinary neutron detector system.
And the correction factors were considered to play important roles in order to accurately estimate subcriticality(−ρ) from measured neutron count rates[1]. The aim of
the experiments was to clarify how to reduce the impact
of correction factors by setting a neutron detector at an
appropriate detector position. For this purpose, the detected-neutron multiplication factor kdet was utilized[2].
The factor kdet is defined by a ratio of total number ofdetected fission-neutrons to total number of detected allneutrons. By the aid of kdet, if effective neutron multiplication factor keff at a reference state is known beforehand,a value of keff at target state is estimated as follows:Σψ
f c, target keff, target = 1 − f s (1 − f c, ref keff, ref ) d ref , (1)Σ d ψ target
where the subscripts “ref” and “target” mean values atreference and target subcritical states, respectively; fs is asource-flux correction factor to fix the difference of
non-fission component of neutron count rate between
reference and target states; fc is conversion factor fromkdet to keff, defined by the ratio of kdet/keff. These factors fs
and fc can be calculated by forward neutron flux calculations only without adjoint calculations. When the neutron
detector is set at an appropriate position where kdet=keff, it
is expected that correction factor fc could be negligible.
On the basis of this idea, the appropriate detector position
can be predicted with the use of numerical analyses of keffand spatial dependency of neutron count rates.
EXPERIMENTS: The NSM experiments were carriedout in A1/8”P60EU-EU(3) core with three types of external neutron sources: Am-Be source in outer reflector
region; Cf source in the center of core; DT source. FourHe-3 detectors and three fission chambers were put inouter reflector region, and neutron count rates were
measured for six subcritical states by changing six control rods. Note that count rates were corrected to eliminate dead-time effect of detector. In order to verify (−ρ)
measured by NSM, the reference values of (−ρ) for six
states were evaluated by control rod worth obtained by
the rod drop method.
NUMERICAL ANALYSIS: Numerical analyses wereconducted with the use of three-dimensional and multienergy-group SN neutron transport code THREEDANT[3]. The macroscopic crosssections used in DANTSYS

採択課題番号 CA23106

Fig. 1. kdet=keff map for all rod-in, Am-Be case
6%

6%
He3#1

5%

5%

He3#2
He3#3

4%

(-ρ) by NSM [Δk/k]

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

1

were attained by SRAC2006/PIJ[4] with JENDL-4.0[5].
By the aid of numerical analyses, kdet and correction factors, fs and fc, can be evaluated for each spatial mesh.
RESULTS: In this paper, the results of Am-Be source
were presented. Figure 1 shows one of the numerical results of relative difference between kdet and keff at the axial center of core, for all control rod-in case. In Fig. 1,
green colored regions can be regard as the appropriate
detector positions of kdet=keff, thus, it is predicted that
positions of He3#3 and FC#3 are appropriate detector
positions in the NSM experiments. Figure 2 summarizes
the experimental results of NSM. As shown in Fig. 2, themeasured (−ρ) by NSM was confirmed to be improved
when fs and fc corrections are applied, and (−ρ) of He3#3
and FC#3 are relatively well in agreement with reference
values even without fs and fc corrections because of theirappropriate detector positions.

(-ρ) by NSM [Δk/k]

T. Endo, T. Fujita, S. Kato, K. Tsujita, K. Nakano
C. H. Pyeon1 and T. Yagi1

He3#4
FC#1
FC#2

3%

FC#3

2%
1%

4%
3%
2%
1%

0%

0%

0%

1%

2%

3%

0%

1%

2%

3%

(-ρ) by control rod worth [Δk/k](-ρ) by control rod worth [Δk/k](a) without fs and fc correction(b) with fs and fc correction

Fig. 2. Measured subcriticality by NSM for Am-Be case
REFERENCES:[1] D. G. Cacuci, Handbook of Nuclear Engineering:
Vol.3 Reactor Analysis, Springer, (2010) 2147-2152 .
[2] T. Endo, et al., Ann. Nucl. Energy, 38(2011)
2417-2427.
[3] R. E. Alcouffe, et al., LA-12969-M, (1995).
[4] K. Okumura, et al., JAEA-Data/Code 2007-004(2007).
[5] K. Shibata, et al., J. Nucl. Sci. Technol., 48(2011)
1-30 .

出中性子増倍率に基づいた中性子源増倍率法による
共同通常
未臨界度測定実験（名大・工）遠藤知弘、藤田達也、加藤慎也、辻田浩介、中野幸太郎（京大炉）卞哲浩、八木貴宏

Measurement of Gamma Ray Spectrum for Estimation of Subcriticality Index

Y. Nauchi, T. Kameyama*, H. Unesaki1, T. Misawa1 T.
Sano1, T. Yagi1
Central Research Institute of Electric Power Industry
1
Research Reactor Institute, Kyoto University* Present address: Tokai University
INTRODUCTION: The neutron multiplication factor isdetermined by fission and radiative capture reactions.
Some of those reactions emit  rays of spectrum intrinsic
to each reaction. We have conducted development oftechniques to quantify the reaction rates by counting the rays in order to estimate the multiplication factor of subcritical systems.
EXPERIMENTS: Subcritcal cores of 2x2 3x2, and 4x2
assemblies driven by a 252Cf source are mocked up in
KUCA-C light water tank and  rays from the cores were
measured by a BGO scintillator placed 43cm outside thecore (Fig.1).  rays were measured twice for each core.
Each measurement time was 0.5~1h. Count rate of  rays
(3~5MeV) in the second measurement is slightly larger(<1%) than that in the first one for every cores. It indicates accumulation of fission products (FP) even in the
low power sub-critical cores.  rays were also measured
for background cases where the 252Cf source was discharged. Comparison of the pulse height spectra is shownin Fig. 2. In the background spectrum, 1.592MeV and
2.521MeV  rays from 140La are identified and no  rays
were measured in higher energy region (> 3MeV). Background subtracted (net) pulse height spectrum consists of
2.223MeV peak corresponding to H(n,) reactions, continuum part (3-6.5MeV) and peak structure in 7~8MeV
energy region. Major component of the peak structure is
identified Al(n,) reaction by additional measurements of rays for cases loading either aluminum or SUS 304
plates between the cores and the BGO scintillator.
DATA ANALYSIS: We assumed that the net pulse heightspectrum consists of 4 components that are from 1)
spontaneous fission of 252Cf, 2) fission of 235U, 3) Al(n,),4) SUS(n,). At first, ray emission spectra were surveyed. For the 1) and 2) components, an evaluation on
prompt  ray by Verbeke [1] is adopted. In addition, spectra of  rays from FP accumulated in the 252Cf source andthe 235U fuel were calculated with the FPGS90 code [2].
For 3) and 4) components, thermal neutron capture  raydata [3] were adopted. Pulse height responses of the BGO
scintillator for  rays of the emission spectra are simulated by coupled photon-electron transport calculations with
the MCNP-5 code. By the responses and the measured
pulse height spectra,  ray emission and reaction rates are
deduced (see plotted lines in Fig.2.). The estimated reaction rates are compared to the neutronics calculations by
the MCNP-5 code in Table 1. The reaction rates ofAl(n,) by the present method agree well with those by
the neutoronics calculations. Accordingly quantificationof capture reaction rates by  ray measurements is promising. On the other hand, the present method gives larger
fission rates of 235U than those by the neutronics calculations, although ratios of the two fission rates do not vary
for the 3 sub-critical cores. The facts indicate that otherreactions occur emitting  rays in energy region 3~6MeV.
We shall continue numerical analysis of  ray yield bycoupled neutron-photon transport analysis to establish thequantification method of fission rate.

Fig.1 Horizontal geometry of  ray measurement.
Measured
Measured background
net spectrum
Clac: Cf S.F.
Calc: Cf FP decay
Calc: 235U fission
Calc. 235U FP decay
Calc: Al(n,)Calc: SUS(n,)Calc:total

1000

Pulse height spectrum (cps/MeV)

CO3-6

100

10

1

0

2

4
6
Pulse height (MeV)

8

10

Fig. 2  ray pulse height spectra and calculated responses.
Table1. Comparison of reaction rates by  ray measurements with those by neutronics calculation by MCNP-5
Cores

Al(n,) reaction rates

235

U fission rates

Present

Neutronics Calc.

Present

Neutronics Calc.

2x2

1.48E+05

1.46E+05

1.57E+06

1.05E+06

3x2

2.71E+05

2.72E+05

3.01E+06

1.99E+06

4x2

4.39E+05

4.67E+05

5.39E+06

3.42E+06

REFERENCES:[1] J.M.Verbeke et al., UCRL-AR-228518, 2009.
[2] H.Ihara et al., JAERI-Data/Code 95-104, 2995.
[3] CapGam http://www.nndc.bnl.gov/capgam/

採択課題番号 CA23107
γ線計測を利用した未臨界度指標の計測に関する研究
共同通常(電中研)名内泰志、亀山高範*、笹原昭博、（京大・原子炉）宇根崎博信、三澤毅、佐野忠史、八木貴宏 * 現: 東海大

CO3-7

Reactor Noise Experiments with Time Series Data by High Time Resolution (2)A. Kitano1, Y. Tamagawa2, I. Ogawa2, N. Fujihara2,
A. Kawamura2, T. Misawa3, H. Taninaka1 and T. Hazama1
1

Japn Atomic Energy Agency
Fukui University
3
Research Reactor Institute, Kyoto University

2

INTRODUCTION: Many reactor noise measurements
have been conducted and reported so far. Kitamura performed the noise experiments taking advantage of thetime series data of pulse train from the neutron detector. [1]In this measurement, the time series data of pulse train
was used for the noise analysis of Feynman-to evaluate
the prompt neutron decay constant p. The data acquisition system with the time resolution of 2 nsec was constructed with the 8ch 500MHz FADC board driven with
FPGA. This FADC board was used for nuclear rare decay
experiment to take the waveforms of the radiation detectors originally, developed to save the waveform andtimestamp of trigger timing of the B-10 neutron proportional counter as shown in Fig. 1.

RESULTS: As shown in Fig. 2, the data acquisition
system modified in this study gives the suitable Y value
for Feynman-method. The results of Feynman-methodshown in Table 1, were in good agreement with the reference value within a reasonable range
for Case1-4.
CONCLUSION: It was demonstrated that the applicability of the modified data acquisition system with FADC
to the measurement of sub-criticality was confirmed bythis experiment. This system shows that the discriminatedtrigger timing is not the rising point of pulse exactly, andthe time correction of every signal pulse to determine thereal rising time using waveforms will be needed to describe the faster events.

採択課題番号 CA23109

Relation between pulse height and rising point

Y value

EXPERIMENTS: This experiment was carried out inthe KUCA-A core using the polyethylene moderator.
Three Control rods (C1-C3) and three safetyrods(S4-S6) are arranged in this experiment. The noise
measurement was conducted for the following conditions,-103.4₵, -38.65₵, -13.4₵ (Case1-3). The time series dataobtained as time stamp data by pulse signal from the detectors located around the core was processed to evaluate
p. The reference values were determined by the control
rod worth based on the rod drop experiment and period
method.
The delayed criticality was achieved by C1 at the position of 668.53mm from the fully withdrawn position withother control rods withdrawn fully.

Fig. 1-103.4¢Gate time(sec)Fig. 2 The results of Feynman-methodin subcritical state by -103.4₵

Table 1

Case
1
2
3
4

Measured sub-criticality
by Feynman-method
Detector(b)

B-10
B-10 (b)B-10 (b)He-3 (c)Sub-criticality (₵)
Measurementa-103.4-38.65-13.40-38.65

Reference (a)
-90.3
-34.3
-7.40
-41.2

(a) Evaluated by the control position and ctirical point(b) B-10 lined poroportional counter(c) He-3 proportional counter

RFERENCE:[1] Y. Kitamura, et al.,”Reactor Noise Experiments byUsing Acquisition System for Time Series Data of Pulse
Train”, J. Nucl. Sci. Technol., 36[8], 653-600(1999).

高時間分解能データ処理を活用した炉雑音測定(2)（原子力機構）北野 彰洋、谷中 裕、羽様 平（福井大）玉川 洋一、小川 泉、藤原 直生、川村（京大・原子炉）三澤 毅

篤史

共同通常