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ASNC upgrade for nuclear material accountancy of ACPF
Accepted Manuscript ASNC upgrade for nuclear material accountancy of ACPF Hee Seo, Seong-Kyu Ahn, Chaehun Lee, Jong-Myeong Oh, Seonkwang Yoon
PII: DOI: Reference:
S0168-9002(17)31127-0 https://doi.org/10.1016/j.nima.2017.10.045 NIMA 60191
To appear in:
Nuclear Inst. and Methods in Physics Research, A
Received date : 13 September 2017 Revised date : 16 October 2017 Accepted date : 17 October 2017 Please cite this article as: H. Seo, S. Ahn, C. Lee, J. Oh, S. Yoon, ASNC upgrade for nuclear material accountancy of ACPF, Nuclear Inst. and Methods in Physics Research, A (2017), https://doi.org/10.1016/j.nima.2017.10.045 This is a PDF file of an unedited manuscript that has been accepted for publication. As a service to our customers we are providing this early version of the manuscript. The manuscript will undergo copyediting, typesetting, and review of the resulting proof before it is published in its final form. Please note that during the production process errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.
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NIMA–D–17–00892.Rev1
1
ASNC Upgrade for Nuclear Material Accountancy of ACPF
2
Hee Seo a,*, Seong-Kyu Ahn a, Chaehun Lee a, Jong-Myeong Oh a, and Seonkwang Yoon a,b
3 4 a
5
Korea Atomic Energy Research Institute, 989-111 Daedeok, Yuseong, Daejeon 34057, Korea b
6
University of Science & Technology, 217 Gajeong, Yuseong, Daejeon 34113, Korea
7 8
Highlights
9 10
The ACP safeguards neutron counter (ASNC) for nuclear material accountancy of the ACPF was upgraded.
11
The remote-handling and maintenance capabilities for hot-cell operation were improved.
12
Various parameters were measured and determined for detector characterization.
13 14 15
Abstract
16
A safeguards neutron coincidence counter for nuclear material accountancy of the Advanced
17
spent-fuel Conditioning Process Facility (ACPF), known as the ACP Safeguards Neutron
18
Counter (ASNC), was upgraded to improve its remote-handling and maintenance capabilities.
19
Based on the results of the previous design study, the neutron counter was completely rebuilt,
20
and various detector parameters for neutron coincidence counting (i.e., high-voltage plateau,
21
efficiency profile, dead time, die-away time, gate length, doubles gate fraction, and stability)
22
were experimentally determined. The measurement data showed good agreement with the
23
MCNP simulation results. To the best of the authors’ knowledge, the ASNC is the only
24
safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell.
25
The final goals to be achieved were (1) to evaluate the uncertainty level of the ASNC in nuclear
26
material accountancy of the process materials of the oxide-reduction process for spent fuels and
27
(2) to evaluate the applicability of the neutron coincidence counting technique within a strong *
Corresponding author Email address: [email protected] (H. Seo) 1
NIMA–D–17–00892.Rev1 28
radiation field (e.g., in a hot-cell environment).
29 30
Keywords: Safeguards, Nuclear material accountancy, Non-destructive assay, Neutron
31
detector
32 33 34 35
1. Introduction
36
An advanced spent-fuel conditioning process facility (ACPF) has been built in KAERI and
37
recently reconstructed for demonstration of the oxide-reduction process using UO2 spent fuel
38
discharged from a commercial PWR nuclear power plant [1–4]. The ACPF is a kg-scale hot-cell
39
facility consisting of two cells: a maintenance cell and a process cell. The cells are 2.2 m × 2.0
40
m × 4.3 m and 8.1 m × 2.0 m × 4.3 m in size, respectively. The process cell, though operated in
41
an air environment, contains, for the oxide-reduction process, an Ar compartment that uses
42
highly corrosive molten salt (LiCl) as an electrochemical reaction medium. The batch size is
43
about 600 gHM. There are five shielded windows: one for the maintenance cell, three for the
44
process cell with air environment, and one for the Ar compartment. Indeed, the ACPF is a test-
45
bed facility not only for pyroprocessing itself but also for various safeguards technologies such
46
as nuclear material accountancy, process monitoring, and containment and surveillance (C/S).
47
In this study, a well-type passive neutron coincidence counter for the ACPF, known as the
48
ACP safeguards neutron counter (ASNC) [5–7], was rebuilt to enhance its remote-handling and
49
maintenance capabilities. The ASNC is based on coincidence neutron counting, rather than
50
multiplicity counting, with the Cm balance technique [8]. First, the amount of 244Cm in a sample
51
will be determined from the measured doubles rate with a predetermined calibration curve. Then,
52
the amount of nuclear materials of interest can be determined from the measured
53
and the ratio of Pu/244Cm or
235
244
Cm amount
U/244Cm, which can be determined from burn-up code 2
NIMA–D–17–00892.Rev1
54
calculations, gamma-ray spectroscopy, or chemical analysis. In a previous study [9], the original
55
version of the ASNC was built and tested successfully using spent-fuel rod cuts; however, it
56
showed limited remote-handling and maintenance capabilities. To overcome this limitation, the
57
ASNC was redesigned based on MCNP6 [10] simulations and a gamma-ray irradiation test [11].
58
The objectives of this redesign work were (1) to optimize the inner structure of the ASNC in
59
consideration of the batch size and gamma-ray dose rate, (2) to reduce its weight and footprint,
60
(3) to improve the remote-handling and maintenance capabilities, and (4) to reduce the
61
measurement uncertainty by improving the detection efficiency and flattening the efficiency
62
profile. Based on this design work, we built an upgraded version of the ASNC and installed it in
63
the air cell of the ACPF. The final goals to be achieved were (1) to evaluate the uncertainty level
64
of the ASNC, which is based on the Cm balance technique, in nuclear material accountancy of
65
the process materials of the oxide-reduction process (i.e., input/output materials as well as LiCl
66
salt) and (2) to evaluate the applicability of the ASNC in a hot-cell environment. This paper
67
describes the details of the inner structure of the upgraded ASNC and discusses the experimental
68
results for detector characteristics including the high-voltage plateau, efficiency profile, dead
69
time, die-away time, gate length, doubles gate fraction, and stability.
70 71 72
2. Components of ASNC
73 74
The ASNC consists of internal and external gamma-ray shields, a neutron moderator, an 3
75
external neutron shield, neutron reflectors,
76
processing circuits, a sample container, a cable enclosure, and an LED-based monitoring box.
77
Each component can be extracted individually from the body whenever there is any problem or
78
replacement requirement, which feature provides an enhanced remote maintenance capability.
3
He neutron detectors and associated signal
NIMA–D–17–00892.Rev1 79
Some components are shown in Fig. 1.
80
81 82
Fig. 1. Components of ASNC: (a) internal gamma-ray shield, (b) external gamma-ray shield, (c)
83
neutron moderator, (d) external neutron shield, (e) cable enclosure and LED-based monitoring
84
box, and (f) sample container.
85 86
2.1. Internal gamma-ray shield
87 88
The role of the internal gamma-ray shield is to reduce the pile-up effect [12] due to gamma-
89
rays, emitted from the sample to be measured. It is made of 5-cm-thick lead (Pb), as determined
90
by MCNP simulations and an irradiation test that considered the batch size and isotopic
4
NIMA–D–17–00892.Rev1
91
compositions of spent fuels [11]. For increased Pb strength and improved processibility, the
92
shield actually consists of a melted mixture of 96.8% Pb and 3.2% tin (Sn). The housing, in
93
consideration of the weight of the shield, was fabricated of 4-mm-thick stainless steel (STS304)
94
to dimensions of 14.2 cm inner diameter (ID), 25.6 cm outer diameter (OD), and 82.2 cm height
95
(H), for a total weight of 302 kg.
96 97
2.2. External gamma-ray shield
98 99
In light of the frequent malfunctioning of electronics as the result of the long-term
100
irradiation of gamma-rays in a hot-cell [13,14], the role of the external gamma-ray shield is to
101
absorb gamma-rays emitted from other sources in a hot-cell. Additionally, the external gamma-
102
ray shield can further reduce the pile-up effect in the neutron detector. It was fabricated of the
103
same material as used for the internal shield to dimensions of 49.8 cm (ID), 56.4 cm (OD), and
104
82.2 cm (H), for a total weight of 448 kg.
105 106
2.3. Neutron moderator
107 108
The role of the neutron moderator is to thermalize neutrons emitted from sources to be
109
measured, thereby enhancing detection efficiency for a 3He neutron detector. The moderator is
110
made of high-density polyethylene (HDPE, ρ=0.96 g/cm3) for superior moderation power and
111
processibility. The thickness was selected in order to maximize detection efficiency through
112
MCNP simulations. A 1-mm-thick cadmium (Cd) sheet (99.95% purity, Advent Research
113
Materials Ltd., UK) neutron absorber was installed at the middle level of the neutron detectors,
114
covering 10 cm and 120 degrees of those detectors, in order to flatten the axial detection
115
efficiency. An additional Cd sheet (1-mm-thick, 30-cm-high) was wrapped around the outside of
5
NIMA–D–17–00892.Rev1
116
the moderator to absorb thermal neutrons coming from the outside, which were thermalized in
117
the external neutron shield. The moderator was fabricated to overall dimensions of 25.7 cm (ID),
118
49.7 cm (OD), and 64.0 cm (H), for a total weight of 103 kg.
119 120
2.4. External neutron shield
121 122
The role of the external neutron shield, made of HDPE, is to reduce the background of
123
neutrons emitted from other sources in a hot-cell. Incident neutrons are thermalized and
124
absorbed by the external neutron shield and/or Cd sheet installed around the outside of the
125
moderator, as described above. The neutron shield was fabricated to dimensions of 56.4 cm (ID),
126
64.5 cm (OD), and 82.2 cm (H), for a total weight of 57.1 kg.
127 128
2.5. Cable enclosure and LED-based monitoring box
129 130
The role of the cable enclosure is to cover the cables that transfer signals generated from
131
detectors and supply high-voltage (HV) and low-voltage (LV) power to the detectors and
132
electronics, respectively. LEMO connectors were used in consideration of their remote-handling
133
capability. Via the LED monitoring box, the condition of each detector could be visually
134
monitored. The materials used in the fabrication of the cable enclosure and LED-based
135
monitoring box were aluminum (A6061) and stainless steel (STS304), respectively. The
136
combined weight is 9.1 kg.
137 138
2.6. Sample container
139 140
The role of the sample container is to facilitate sample handling in a hot-cell environment
6
NIMA–D–17–00892.Rev1
141
via a manipulator. It has a cavity for sample placement (cavity size: 13.4 cm (D) × 26.0 cm (H)).
142
The container was fabricated of aluminum (A6061) in consideration of that material’s low
143
neutron-absorption cross-section. The overall weight, including the neutron reflectors (see
144
below), is 7.8 kg.
145 146
2.7. Neutron reflectors
147 148
The role of the two neutron reflectors located at the top and bottom of the cavity,
149
respectively, is to increase detection efficiency by reducing neutron leakage and to flatten the
150
axial efficiency profile. The material used for the bottom reflector was nickel (99.5% purity,
151
Carpenter Technology Corporation, USA), and its housing material was STS304. Nickel has
152
positive properties as a neutron reflector, because it has a high-elastic-scattering cross-section
153
with a low-absorption cross-section; however, its heavy weight is a disadvantage in terms of
154
remote-handling and mechanical stability. Thus, nickel was used only for the bottom reflector,
155
whereas the top reflector was fabricated of graphite. The sizes of the top and bottom reflector
156
are 12.8 cm (D) × 17.0 cm (H) and 13.5 cm (D) × 10.0 cm (H), and their weights, 3.78 and 12.7
157
kg, respectively.
158 159
2.8. 3He neutron detectors and associated electronics
160 161
The 3He neutron detectors (RS-P4-0820-116, RE Reuter Stokes, USA) previously used in
162
the ASNC were reused in order to save costs. The 3He-filling pressure and the size of an active
163
volume are 4 atm and 1 in (D) × 20 in (H), respectively. The quenching gas, inserted into the
164
detector to prevent unnecessary avalanches by UV light, is nitrogen gas (N2). The detector body
165
is 0.8-mm-thick aluminum.
7
NIMA–D–17–00892.Rev1
166
In the former ASNC, there was a problem in extracting the 3He detectors from the HDPE
167
moderator using a manipulator, because the inner structure had been deformed over the course
168
of many years of use. As a solution, a guide tube has been installed around the 3He detector.
169
Aluminum (A6061) was used as the guide tube material owing to its neutron-absorption cross-
170
section and mechanical strength. PDT 110A electronics (Precision Data Technology, USA),
171
designed for high-radiation environments (>1 Mrad), an integrated circuit suitable for charge-
172
sensitive pre-amplification, signal shaping, leading-edge discrimination, LED driver, HV
173
filtering, and OR logic applications (summation of the logic pulses) was utilized for signal
174
processing. Overall a total of 24 neutron detectors were installed in a circular-shape pattern.
175 176
Figure 2 shows the inner structure of the upgraded ASNC, and Table 1 summarizes the components’ properties.
177
178 179
Fig. 2. Inner structure of ASNC: (a) schematic side-view and (b) photograph.
180
8
NIMA–D–17–00892.Rev1
181 182
Table 1. Properties of components. Component
Materials
Size
Weight (kg)
Housing
ID: 14.2 cm OD: 25.6 cm H: 82.2 cm ID: 49.8 cm OD: 56.4 cm H: 82.2 cm
302
Stainless steel (STS304) 4-mm-thick
448
Stainless steel (STS304) 4-mm-thick
HDPE
ID: 25.7 cm OD: 49.7 cm H: 64.0 cm
103
None
neutron HDPE
ID: 56.4 cm OD: 64.5 cm H: 82.2 cm ID: 41.4 cm OD: 49.7 cm H: 9.6 cm
57.1
None
9.1
Aluminum (A6061) 3-mm-thick
Internal gamma-ray Pb 96.8% Sn 3.2% shield External gamma-ray Pb 96.8% Sn 3.2% shield Neutron moderator
External shield
Cable enclosure
Al (A6061)
LED-based monitoring box
Stainless steel W: 16.0 cm (STS304) D: 10.0 cm H: 30.0 cm Al (A6061) D: 14.0 cm H: 48.1 cm
Sample container Neutron reflectors
Top
Graphite
Bottom
Ni
3
He detectors
4 atm of 3He (+ N2 gas)
D: 12.8 cm H: 17.0 cm D: 13.5 cm H: 10.0 cm D: 2.54 cm H: 50.8 cm
Stainless steel (STS304) 1.5-mm-thick 7.8
Aluminum (A6061) 3-mm-thick
3.78
Aluminum (A6061) 3-mm-thick 12.7 Stainless steel (STS304) 3-mm-thick Model: RS-P4-0820-116 Body: 0.8-mm-thick aluminum
183 184
2.9. Configuration
185 186
Figure 3 shows the configuration for signals, LV and HV transmission between the inside of
187
the hot-cell and the operation area. The shift register (JSR-14, Canberra, USA), DC power
188
supply (E3648A, Agilent, USA), 4-channel derandomizer (LANL, USA) [15] and PC are
189
located in the operation area. The HV is supplied to the detector from the shift register, while the
190
+12 V DC power is supplied to the electronics from the DC power supply. Although the
191
electronics requires +5 V for operation, we applied +12 V and used the DC voltage regulator
9
NIMA–D–17–00892.Rev1
192
(Aquila Technologies, USA), located immediately in front of the electronics, to convert the +12
193
V into +5 V in consideration of the voltage drop as the current travels through the long cable
194
(~20 m) between the inside and outside of the hot-cell. The signals produced in each detector are
195
summed using the integrated OR circuit in the PDT 110A electronics. In order to reduce the
196
dead time, a total of 24 detector signals were configured into four groups (instead of one). Six
197
detector signals are summed for each group; hence, the four summed signals are transmitted to
198
the derandomizer for reduction of input synchronization loss. The four group signals are
199
summed once again in the derandomizer to produce one summed signal, which is finally fed into
200
the shift register for analysis. The LED signal produced in each detector is directly connected to
201
the LED installed in the monitoring box to enable monitoring of the status of each detector in
202
the operation area through the shielded window. The JSR-14 shift register is controlled by INCC
203
(IAEA Neutron Coincidence Counting) software [16].
204
Figure 4 shows the upgraded ASNC installed in the hot-cell of the ACPF. There are only
205
three operations that need to be handled remotely by a manipulator for measurement: (1)
206
opening/closing of the cover of the sample container, (2) moving of the sample container
207
upwards and downwards, and (3) moving of the sample container leftwards and rightwards.
208
These operations were all tested successfully with the manipulators. By these remote operations,
209
the sample to be measured can be safely inserted into the sample container and moved to the
210
measurement location (i.e., inside the cavity).
10
NIMA–D–17–00892.Rev1
211 212
Fig. 3. Configuration for signals and high- and low-voltage transmission.
213
214 215
Fig. 4. ASNC installed in hot-cell with indications of remote-handling parts.
216 217
11
NIMA–D–17–00892.Rev1 218 219
3. Detector characterization
220
Various detector parameters need to be determined in order to characterize the measurement
221
system and to correct the measurement data. The parameters determined in this study were the
222
HV plateau, dead time, efficiency profile, die-away time, gate length, doubles gate fraction, and
223
stability.
224 225
3.1. High-voltage (HV) plateau
226 227
The HV plateau for each detector as well as each group was measured after applying a gain-
228
matching procedure for each detector [17]. As shown in Fig. 5, the gains of the 24 neutron
229
detectors were well matched to each other in order to have the same shape for each detector as
230
well as each group. From these HV plateaus, the operating voltage was determined to be 1720 V.
231
Additionally, the HV plateau of each group was compared with that of the summed signal finally
232
fed into the shift register from the derandomizer. It was found that the averaged count loss at the
233
operating voltage of 1720 V was ~0.21% at the count rate of ~4.33×104 cps, due to the process
234
of summing the four group signals into one in the derandomizer. The HV plateaus were
235
measured for 30 s at 20 V intervals from 1400 to 1900 V. A 252Cf calibration source (L3-693,
236
Eckert & Ziegler Isotope Products, USA) with a neutron emission rate of ~1.93×105 n/s was
237
utilized.
12
NIMA–D–17–00892.Rev1
238 239
Fig. 5. High-voltage plateaus for each detector and group after normalization (top) and for each
240
group and summed signal (bottom).
241 242
3.2. Dead time
243 244
As the count rate increases, the signal loss increases due to the detection system (i.e.,
245
detector and electronics)’s dead time. Various factors can affect the degree of dead time of a
246
3
247
diameter, shaping time, and configuration of preamps. In the ASNC, each 3He proportional
248
counter has a preamp for reduction of the dead-time effect. The signal loss due to the dead time
249
can be corrected empirically by (1) the doubles-to-singles-ratio method, (2) the paired-source
He-based neutron detector, including the 3He-filling pressure, type of quenching gas, detector
13
NIMA–D–17–00892.Rev1
250
method, or (3) the source-intensity-ratio method [18]. In this study, the dead-time correction
251
factors were determined using the source-intensity-ratio method, due to the availability of
252
calibration sources. By this method, a strong and a weak 252Cf source is measured separately; the
253
ratio of the dead-time-corrected count rates should be equal to the ratio of the source intensities.
254
The relationship between the ratios can be expressed by
255 256
(1)
257
(2)
258 259
where S and D are the dead-time-corrected singles and doubles count rates, respectively, Y is the
260
source intensity, and the subscripts s and w represent the weak and strong sources, respectively.
261
Two
252
Cf sources with neutron emission rates of ~1.92×105 n/s (L3-693) and 642.4 n/s
262
(9594NC, AEA Technology, UK) for the strong and weak sources, respectively, were used;
263
accordingly, the source intensity ratio was 299.2. In order to sufficiently reduce the random error,
264
the measurements were performed over 7952 cycles × 30 s/cycle (=66.3 h) and 240 cycles × 30
265
s/cycle (=2 h) for the weak and strong sources, respectively. The dead-time-corrected singles (Sc)
266
and doubles (Dc) count rates [19] can be calculated using the equations
267 268
(3)
269
(4)
270 271
where Sm and Dm are the measured singles and doubles rates, respectively, and δ is given by δ =
272
(A + B·10-6×Sm) μs; where A and B are constants with the relationship B=A2/4 [20]. As a result,
273
the dead-time coefficient A for singles can be calculated from Eqs. (1) and (3), and for doubles,
14
NIMA–D–17–00892.Rev1 274
from Eqs. (2) and (4). The solutions for coefficient A for singles (AS) and doubles (AD) are
275
276
(5)
277
(6)
278 279
where S and D are the singles and doubles rates, respectively, the subscripts s and w represent
280
the weak and strong sources, respectively, and k is the source intensity ratio (i.e., Ys/Yw). From
281
the measurement data and Eqs. (5) and (6), the constants A and B were determined to be 7.44 μs
282
and 13.84 μs2 for the singles rate, respectively, and 2.13 μs and 1.13 μs2 for the doubles rate,
283
respectively. These constants A and B were applied to the INCC software in order to
284
automatically correct the measurement data from the dead-time loss. The fractional count loss
285
for singles was determined to be 7.71% at the count rate of ~4.7×104 cps while 8.78% at the
286
doubles rate of ~9.2×103 cps. Table 2 summarizes the measurement data before and after dead-
287
time correction as well as the fractional count loss.
288 289
Table 2. Measured singles and doubles count rates before and after dead-time correction. Source intensity (n/s)
Singles ± 1σ
Doubles ± 1σ
Before correction (cps)
After correction (cps)
Fractional Before count loss correction (%) (cps)
After correction (cps)
Fractional count loss (%)
Weak source
642.4
156.23 ±0.03
156.28 ±0.03
0.03 ±0.03
30.87 ±0.02
30.88 ±0.02
0.03 ±0.09
Strong source
1.92×105
43149.8 ±3.09
46756.2 ±3.09
7.71 ±0.01
8427.1 ±8.35
9237.8 ±8.35
8.78 ±0.13
Ratio
299.2
276.2 ±0.06
299.2 ±0.06
273.0 ±0.32
299.2 ±0.33
290
15
NIMA–D–17–00892.Rev1 291
3.3. Efficiency profile
292 293
In order to reduce the systematic error due to the sample location within the cavity, the
294
detection efficiency profile should be as flat as possible in both the axial and radial directions.
295
To this end, we designed the ASNC for a flat efficiency profile by MCNP simulations and
296
compared the measured efficiency with the calculated data. The efficiency profiles were
297
measured using the same
298
cm (2-cm intervals) in the axial direction and from 0 to 4.5 cm (1.5-cm intervals) in the radial
299
direction. The source location (0,0) was considered to be the center of the cavity. The
300
measurements for each source location were performed over 10 cycles × 10 s/cycle. The
301
measured efficiency was determined from the dead-time corrected singles rate as divided by the
302
source intensity. The average detection efficiency for the axial and radial source positions was
303
determined to be 24.4±0.07 and 24.3±0.05%, respectively (Fig. 6). The measurement precision
304
errors were smaller than the plot symbols. Indeed, in terms of the efficiency profiles in both the
305
axial and radial directions, the ASNC showed a flat response regardless of source location;
306
therefore, the entire cavity volume can be considered to be a flat efficiency zone. Additionally,
307
the measured efficiency profiles showed excellent agreement with the calculated data, which
308
represents the accuracy of the MCNP model.
252
Cf calibration source (L3-693) at locations ranging from -10 to 10
16
NIMA–D–17–00892.Rev1
309 310
Fig. 6. Measured and calculated efficiency profiles in axial (top) and radial (bottom) directions.
311 312
3.4. Die-away time
313 314
The die-away time is the mean lifetime of a neutron within the detection system [19]. At any
315
moment, a neutron emitted from the sample can disappear from the detection system, either by
316
being absorbed in the moderator or in the neutron detector, or by escaping. It depends on a
317
variety of factors including the structure (size, shape, and presence of cadmium) and efficiency
318
of the detection system as well as the physical/chemical properties of the nuclear material to be
319
measured (scattering, moderation, and production of additional neutrons by induced fissions).
17
NIMA–D–17–00892.Rev1
320
Let us assume that the neutron population in the detection system as a function of time could be
321
expressed by the single exponential
322 323
(7)
324 325
where N(t) and N(0) are the neutron population at time t and 0, respectively, and τ is the die-
326
away time. First, the die-away time of the ASNC was evaluated by MCNP simulations for which
327
the detection probability as a function of time was calculated. After fitting to the exponential
328
decay function with a single time constant, the calculated die-away time was determined to be
329
59.0±0.28 μs (Fig. 7).
330
331 332
Fig. 7. Calculated detection probability as a function of time with fitted curve by exponential
333
decay function.
334 335
The die-away time also can be determined from the measured doubles rates with different
336
gate lengths, followed by exponential growth function curve fitting [22] or by calculation using
337
the equation [19]
18
NIMA–D–17–00892.Rev1
338
(8)
339 340
where G1 and G2 are the gate lengths with the correlation of G2=G1×2, and D1 and D 2 are the
341
doubles count rate with the gate lengths of G1 and G2, respectively. The measured doubles rates
342
with different gate lengths for a fixed pre-delay of 4.5 μs are summarized in Table 3. The
343
measurements were performed over 40 cycles × 30 s/cycle using the
344
~1.92×105 n/s). The die-away time, with curve fitting with the exponential growth function, was
345
determined to be 65.2±1.1 μs (Fig. 8). The die-away times determined from Eq. (8) are
346
summarized in Table 4. In general, the die-away time was increased as the gate interval
347
increased. It is worthwhile to note that a neutron coincidence counter might have more than one
348
die-away time according to the gate interval [18]. Considering the die-away times determined
349
from the simulation (τsimulation=59.0 μs) and from the curve fitting of the measurement data (τcurve
350
fitting=65.2
351
which validates the model. For practical purposes, we will use the value of 62.4 μs determined
352
from the doubles gate method using Eq. (8). This was calculated from the data on the gate
353
interval 45–90 μs, which fully covers the die-away times of 59.0 and 65.2 μs.
252
Cf source (L3-693,
μs), the die-away times of the ASNC from simulation and measurement are similar,
354
355 356 357
Fig. 8. Measured doubles count rates as function of gate length with curve fitting by exponential growth function. 19
NIMA–D–17–00892.Rev1 358
Table 3. Measured doubles rate and its relative error with different gate lengths. Gate length (μs)
Doubles rate ± 1σ (cps)
Relative error (%)
16
3293.1±9.1
0.276
32
5850.1±14.0
0.240
45
7432.4±17.5
0.236
64
9270.5±22.2
0.239
90
11045.8±27.9
0.252
128
12653.0±35.7
0.282
180
13888.3±43.9
0.316
256
14638.3±54.0
0.369
359 360
Table 4. Calculated die-away times from measured doubles count rates with different gate
361
intervals. Gate interval (μs)
Die-away time (μs)
16–32
63.2
32–64
59.6
45–90
62.4
64–128
63.5
90–180
66.3
128–256
69.1
362 363
3.5. Gate length
364 365
The gate length in neutron coincidence counting is the time window applied to the shift
366
register in order to record the doubles rate if two neutron pulses are received within the
367
predefined time window. The accidentals (random coincidences) are increased as the gate length
20
NIMA–D–17–00892.Rev1
368
increases, while some of the reals (true coincidences) can be lost if the gate length defined is too
369
short. The optimal gate length for coincidence counting can be determined from the measured
370
relative error on the doubles rate at different gate lengths. Figure 9 shows the relative doubles
371
errors ranging from 16 to 256 μs, as listed in Table 3. The error of the doubles rates is relatively
372
flat in the region 32–64 μs. We have decided to use the gate length of 64 μs because it is close to
373
the theoretical optimum of 1.27×die-away time (i.e., Goptimum ≈ 1.27×τ) [19].
374
375 376
Fig. 9. Measured doubles errors at different gate lengths.
377 378
3.6. Doubles gate fraction
379 380
The doubles gate fraction is the ratio between the doubles rate with the finite gate length and
381
that with the infinite gate length. It can be calculated by the equation [19], under the assumption
382
that the neutron population can be expressed by a single exponential decay,
383 384
(9)
21
NIMA–D–17–00892.Rev1
385 386
where fd is the doubles gate fraction, P is the pre-delay, τ is the die-away time, and G is the gate
387
length. Using the values of P, τ, and G determined from the previous sections, the fd was
388
determined to be 0.60. However, as mentioned above, there are more than one die-away time for
389
the coincidence counter; therefore, in order to remove the dependency on the die-away time in
390
calculation, the fd was calculated using the equation [21]
391 392
(10)
393 252
394
where νs1 and νs2 are the first and second factorial moments for the
395
(νs1 = 3.757 and νs2 = 11.962 [19]); S and D are the dead-time-corrected singles and doubles
396
rates, respectively; and ε is the detection efficiency for the 252Cf source at the center location. In
397
order to use the Eq. (10), there are requirements for the source condition: no (α,n) neutrons and
398
no multiplication, which are usually satisfied for a
399
measurement data (S = 46756.2 cps, D = 9237.8 cps, and ε=24.3%), the double gate fraction fd
400
of the ASNC was determined to be 0.51. It is worthwhile to note that the decay of the neutron
401
population in the ASNC may have multiple components to their die-away curve, instead of
402
following a single exponential decay, due to the 5-cm-thick inner gamma-ray shield and the Cd
403
sheet installed around the 3He detector for flattening the axial detector efficiency. This could be
404
one of the reasons to cause the difference between the values determined from Eq. (9) and (10).
252
Cf source, respectively,
Cf source. From this equation and the
405 406
3.7. Stability
407 408
Because the ASNC will be operated in the hot-cell environment, the stability is one of the
409
major factors in considering the cost and time required for maintenance. In order to evaluate the
22
NIMA–D–17–00892.Rev1
410
stability of the ASNC, it was tested by measuring the background for 20,000 cycles × 30 s/cycle
411
(~7 days). Figure 10 shows the measured singles and doubles backgrounds for 16,000 cycles.
412
Although more measurements than these were performed, the INCC recorded the raw data of
413
each cycle for only 16,000 cycles. The measured singles and doubles backgrounds were 0.704
414
and 0.037 cps, respectively. The relative standard deviations for singles and doubles were 0.14
415
and 0.27%, respectively. There were some spikes due to spallation events caused by cosmic rays;
416
however, these were removed from the data by the INCC software’s quality control tests.
417
The detector parameters determined in this study are summarized in Table 5.
418
419 420
Fig. 10. Background count rate of ASNC for singles (top) and doubles (bottom).
421
23
NIMA–D–17–00892.Rev1 422
Table 5. Summary of ASNC detector parameters. Parameter
Value
Cavity size (Flat efficiency zone)
13.4 cm (D)×26.0 cm (H)
Operating voltage
1720 V for Singles
A = 7.44 μs B = 13.84 μs2
for Doubles
A = 2.13 μs B = 1.13 μs2
Dead-time coefficient
Detection efficiency (for 252Cf at center of cavity)
24.3%
Die-away time
62.4 μs
Gate length
64 μs
Pre-delay
4.5 μs
Doubles gate fraction
0.51
423 424 425
4. Summary
426 427
The ACPF, a hot-cell facility in KAERI, has been refurbished for oxide-reduction-process
428
testing using spent fuels. Several process- and safeguards-related instruments were installed and
429
are being tested. In the present study, the ASNC, a safeguards neutron coincidence counter, was
430
upgraded to improve its performance in terms of remote-handling, maintenance, and neutron-
431
detection capabilities. Specifically in the latter case, detection efficiency was improved with
432
better efficiency profiles in both the axial and radial directions. Various parameters including the
433
high-voltage plateau, efficiency profile, dead time, die-away time, gate length, doubles gate
434
fraction, and stability were measured for detector characterization. Some measured parameters
435
were compared with MCNP simulation results. In general, the measurement data showed good
436
agreement with the simulation data. To the best of the authors’ knowledge, the ASNC is the only 24
NIMA–D–17–00892.Rev1
437
safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell.
438
In the near future, the upgraded ASNC will be further tested for calibration (with spent-fuel rod
439
cuts) as well as for overall performance (with input and output materials from the oxide-
440
reduction process).
441 442
Acknowledgments
443
This work was supported by a National Research Foundation of Korea (NRF) grant funded by
444
the Korean government (MSIP) (No. NRF-2017M2A8A5015084).
445 446 447
References
448
1. I.J. Cho, D.H. Kook, K.C. Kwon, E.P. Lee, W.M. Choung, G.S. You, Design and
449
Verification of Shielding for the Advanced Spent Fuel Conditioning Process Facility,
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2. G.S. You, W.M. Choung, J.H. Ku, I.J. Cho, D.H. Kook, K.C. Kwon, E.P. Lee, W.K. Lee,
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Design and construction of an advanced spent fuel conditioning process facility (ACPF),
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3. S.N. Yu, J.K. Lee, B.S. Park, I.J. Cho, K. Kim, Hot cell renovation in the spent fuel
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6. T.H. Lee, H.D. Kim, J.S. Yoon, S.Y. Lee, M. Swinhoe, H.O. Menlove, Preliminary
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1423–1427.
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8. P.M. Rinard, H.O. Menlove, Application of Curium Measurements for Safeguarding at
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Reprocessing Plants, Los Alamos National Laboratory Report LA-13134-MS, Los Alamos,
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USA, 1996.
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conditioning process safeguards neutron counter for PWR spent fuel rods, Nucl. Technol.
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176 (2011) 147–154.
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10. T. Goorley et al, Initial MCNP6 release overview, Nucl. Technol., 180 (2012) 298–315.
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11. H. Seo, C. Lee, J.M. Oh, S.J. An, S.K. Ahn, S.H. Park, J.H. Ku, Monte Carlo Simulations of
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Safeguards Neutron Counter for Oxide Reduction Process Feed Material, J. Korean Phys.
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Soc. 69 (2016) 1175–1181.
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12. D. Mazed, S. Mameri, R. Ciolini, Design parameters and technology optimization of 3He-
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filled proportional counters for thermal neutron detection and spectrometry applications,
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Radiat. Meas. 47 (2012) 577–587.
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13. H.O. Menlove, P.M. Rinard, S.F. Klosterbuer, K. Czock, G. Tributsch, Y.G. Lee, S.H. Cha,
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H.D. Kim, The calibration of the DSNC for the measurement of Cm-244 and plutonium,
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Los Alamos National Laboratory Report LA-UR-99-6217, Los Alamos, USA, 1999.
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14. H.J. Kim, W.I. Ko, S.Y. Lee, H.D. Kim, The Effect of the Neutron Multiplication Factor on
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a Nuclear Material Measurement in Dry-Processed Spent Fuel Material, J. Nucl. Sci.
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Technol. S4 (2004) 384–387.
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counters with AMPTEK preamplifiers, Los Alamos National Laboratory Report LA-UR-
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16. M.S. Krick, B. Harker, INCC Software User's Manual, Los Alamos National Laboratory Report LA-UR-10-6227, Los Alamos, USA, 2009. 17. H.O. Menlove, Amplifier Setup Procedure for He-3 Tubes, Los Alamos National Laboratory Report LA-UR-06-2644, Los Alamos, USA, 2005.
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18. A.M. LaFleur, S.K. Ahn, H.O. Menlove, M.C. Browne, H.D. Kim, H.D., Characterization
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safeguards measurements, Radiat. Meas. 61 (2014) 83–93.
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19. N. Ensslin, W.C. Harker, M.S. Krick, D.G. Langner, M.M. Pickrell, J.E. Stewart,
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Manual, Los Alamos National Laboratory Report LA-13422-M, Los Alamos, USA, 1998.
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21. L.G. Evans, Pulse Train Analysis Applied to the Re-Evaluation of Deadtime Correction
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Factors for Correlated Neutron Counting. University of Birmingham, Ph.D. Dissertation,
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PII: DOI: Reference:
S0168-9002(17)31127-0 https://doi.org/10.1016/j.nima.2017.10.045 NIMA 60191
To appear in:
Nuclear Inst. and Methods in Physics Research, A
Received date : 13 September 2017 Revised date : 16 October 2017 Accepted date : 17 October 2017 Please cite this article as: H. Seo, S. Ahn, C. Lee, J. Oh, S. Yoon, ASNC upgrade for nuclear material accountancy of ACPF, Nuclear Inst. and Methods in Physics Research, A (2017), https://doi.org/10.1016/j.nima.2017.10.045 This is a PDF file of an unedited manuscript that has been accepted for publication. As a service to our customers we are providing this early version of the manuscript. The manuscript will undergo copyediting, typesetting, and review of the resulting proof before it is published in its final form. Please note that during the production process errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.
*Manuscript Click here to view linked References
NIMA–D–17–00892.Rev1
1
ASNC Upgrade for Nuclear Material Accountancy of ACPF
2
Hee Seo a,*, Seong-Kyu Ahn a, Chaehun Lee a, Jong-Myeong Oh a, and Seonkwang Yoon a,b
3 4 a
5
Korea Atomic Energy Research Institute, 989-111 Daedeok, Yuseong, Daejeon 34057, Korea b
6
University of Science & Technology, 217 Gajeong, Yuseong, Daejeon 34113, Korea
7 8
Highlights
9 10
The ACP safeguards neutron counter (ASNC) for nuclear material accountancy of the ACPF was upgraded.
11
The remote-handling and maintenance capabilities for hot-cell operation were improved.
12
Various parameters were measured and determined for detector characterization.
13 14 15
Abstract
16
A safeguards neutron coincidence counter for nuclear material accountancy of the Advanced
17
spent-fuel Conditioning Process Facility (ACPF), known as the ACP Safeguards Neutron
18
Counter (ASNC), was upgraded to improve its remote-handling and maintenance capabilities.
19
Based on the results of the previous design study, the neutron counter was completely rebuilt,
20
and various detector parameters for neutron coincidence counting (i.e., high-voltage plateau,
21
efficiency profile, dead time, die-away time, gate length, doubles gate fraction, and stability)
22
were experimentally determined. The measurement data showed good agreement with the
23
MCNP simulation results. To the best of the authors’ knowledge, the ASNC is the only
24
safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell.
25
The final goals to be achieved were (1) to evaluate the uncertainty level of the ASNC in nuclear
26
material accountancy of the process materials of the oxide-reduction process for spent fuels and
27
(2) to evaluate the applicability of the neutron coincidence counting technique within a strong *
Corresponding author Email address: [email protected] (H. Seo) 1
NIMA–D–17–00892.Rev1 28
radiation field (e.g., in a hot-cell environment).
29 30
Keywords: Safeguards, Nuclear material accountancy, Non-destructive assay, Neutron
31
detector
32 33 34 35
1. Introduction
36
An advanced spent-fuel conditioning process facility (ACPF) has been built in KAERI and
37
recently reconstructed for demonstration of the oxide-reduction process using UO2 spent fuel
38
discharged from a commercial PWR nuclear power plant [1–4]. The ACPF is a kg-scale hot-cell
39
facility consisting of two cells: a maintenance cell and a process cell. The cells are 2.2 m × 2.0
40
m × 4.3 m and 8.1 m × 2.0 m × 4.3 m in size, respectively. The process cell, though operated in
41
an air environment, contains, for the oxide-reduction process, an Ar compartment that uses
42
highly corrosive molten salt (LiCl) as an electrochemical reaction medium. The batch size is
43
about 600 gHM. There are five shielded windows: one for the maintenance cell, three for the
44
process cell with air environment, and one for the Ar compartment. Indeed, the ACPF is a test-
45
bed facility not only for pyroprocessing itself but also for various safeguards technologies such
46
as nuclear material accountancy, process monitoring, and containment and surveillance (C/S).
47
In this study, a well-type passive neutron coincidence counter for the ACPF, known as the
48
ACP safeguards neutron counter (ASNC) [5–7], was rebuilt to enhance its remote-handling and
49
maintenance capabilities. The ASNC is based on coincidence neutron counting, rather than
50
multiplicity counting, with the Cm balance technique [8]. First, the amount of 244Cm in a sample
51
will be determined from the measured doubles rate with a predetermined calibration curve. Then,
52
the amount of nuclear materials of interest can be determined from the measured
53
and the ratio of Pu/244Cm or
235
244
Cm amount
U/244Cm, which can be determined from burn-up code 2
NIMA–D–17–00892.Rev1
54
calculations, gamma-ray spectroscopy, or chemical analysis. In a previous study [9], the original
55
version of the ASNC was built and tested successfully using spent-fuel rod cuts; however, it
56
showed limited remote-handling and maintenance capabilities. To overcome this limitation, the
57
ASNC was redesigned based on MCNP6 [10] simulations and a gamma-ray irradiation test [11].
58
The objectives of this redesign work were (1) to optimize the inner structure of the ASNC in
59
consideration of the batch size and gamma-ray dose rate, (2) to reduce its weight and footprint,
60
(3) to improve the remote-handling and maintenance capabilities, and (4) to reduce the
61
measurement uncertainty by improving the detection efficiency and flattening the efficiency
62
profile. Based on this design work, we built an upgraded version of the ASNC and installed it in
63
the air cell of the ACPF. The final goals to be achieved were (1) to evaluate the uncertainty level
64
of the ASNC, which is based on the Cm balance technique, in nuclear material accountancy of
65
the process materials of the oxide-reduction process (i.e., input/output materials as well as LiCl
66
salt) and (2) to evaluate the applicability of the ASNC in a hot-cell environment. This paper
67
describes the details of the inner structure of the upgraded ASNC and discusses the experimental
68
results for detector characteristics including the high-voltage plateau, efficiency profile, dead
69
time, die-away time, gate length, doubles gate fraction, and stability.
70 71 72
2. Components of ASNC
73 74
The ASNC consists of internal and external gamma-ray shields, a neutron moderator, an 3
75
external neutron shield, neutron reflectors,
76
processing circuits, a sample container, a cable enclosure, and an LED-based monitoring box.
77
Each component can be extracted individually from the body whenever there is any problem or
78
replacement requirement, which feature provides an enhanced remote maintenance capability.
3
He neutron detectors and associated signal
NIMA–D–17–00892.Rev1 79
Some components are shown in Fig. 1.
80
81 82
Fig. 1. Components of ASNC: (a) internal gamma-ray shield, (b) external gamma-ray shield, (c)
83
neutron moderator, (d) external neutron shield, (e) cable enclosure and LED-based monitoring
84
box, and (f) sample container.
85 86
2.1. Internal gamma-ray shield
87 88
The role of the internal gamma-ray shield is to reduce the pile-up effect [12] due to gamma-
89
rays, emitted from the sample to be measured. It is made of 5-cm-thick lead (Pb), as determined
90
by MCNP simulations and an irradiation test that considered the batch size and isotopic
4
NIMA–D–17–00892.Rev1
91
compositions of spent fuels [11]. For increased Pb strength and improved processibility, the
92
shield actually consists of a melted mixture of 96.8% Pb and 3.2% tin (Sn). The housing, in
93
consideration of the weight of the shield, was fabricated of 4-mm-thick stainless steel (STS304)
94
to dimensions of 14.2 cm inner diameter (ID), 25.6 cm outer diameter (OD), and 82.2 cm height
95
(H), for a total weight of 302 kg.
96 97
2.2. External gamma-ray shield
98 99
In light of the frequent malfunctioning of electronics as the result of the long-term
100
irradiation of gamma-rays in a hot-cell [13,14], the role of the external gamma-ray shield is to
101
absorb gamma-rays emitted from other sources in a hot-cell. Additionally, the external gamma-
102
ray shield can further reduce the pile-up effect in the neutron detector. It was fabricated of the
103
same material as used for the internal shield to dimensions of 49.8 cm (ID), 56.4 cm (OD), and
104
82.2 cm (H), for a total weight of 448 kg.
105 106
2.3. Neutron moderator
107 108
The role of the neutron moderator is to thermalize neutrons emitted from sources to be
109
measured, thereby enhancing detection efficiency for a 3He neutron detector. The moderator is
110
made of high-density polyethylene (HDPE, ρ=0.96 g/cm3) for superior moderation power and
111
processibility. The thickness was selected in order to maximize detection efficiency through
112
MCNP simulations. A 1-mm-thick cadmium (Cd) sheet (99.95% purity, Advent Research
113
Materials Ltd., UK) neutron absorber was installed at the middle level of the neutron detectors,
114
covering 10 cm and 120 degrees of those detectors, in order to flatten the axial detection
115
efficiency. An additional Cd sheet (1-mm-thick, 30-cm-high) was wrapped around the outside of
5
NIMA–D–17–00892.Rev1
116
the moderator to absorb thermal neutrons coming from the outside, which were thermalized in
117
the external neutron shield. The moderator was fabricated to overall dimensions of 25.7 cm (ID),
118
49.7 cm (OD), and 64.0 cm (H), for a total weight of 103 kg.
119 120
2.4. External neutron shield
121 122
The role of the external neutron shield, made of HDPE, is to reduce the background of
123
neutrons emitted from other sources in a hot-cell. Incident neutrons are thermalized and
124
absorbed by the external neutron shield and/or Cd sheet installed around the outside of the
125
moderator, as described above. The neutron shield was fabricated to dimensions of 56.4 cm (ID),
126
64.5 cm (OD), and 82.2 cm (H), for a total weight of 57.1 kg.
127 128
2.5. Cable enclosure and LED-based monitoring box
129 130
The role of the cable enclosure is to cover the cables that transfer signals generated from
131
detectors and supply high-voltage (HV) and low-voltage (LV) power to the detectors and
132
electronics, respectively. LEMO connectors were used in consideration of their remote-handling
133
capability. Via the LED monitoring box, the condition of each detector could be visually
134
monitored. The materials used in the fabrication of the cable enclosure and LED-based
135
monitoring box were aluminum (A6061) and stainless steel (STS304), respectively. The
136
combined weight is 9.1 kg.
137 138
2.6. Sample container
139 140
The role of the sample container is to facilitate sample handling in a hot-cell environment
6
NIMA–D–17–00892.Rev1
141
via a manipulator. It has a cavity for sample placement (cavity size: 13.4 cm (D) × 26.0 cm (H)).
142
The container was fabricated of aluminum (A6061) in consideration of that material’s low
143
neutron-absorption cross-section. The overall weight, including the neutron reflectors (see
144
below), is 7.8 kg.
145 146
2.7. Neutron reflectors
147 148
The role of the two neutron reflectors located at the top and bottom of the cavity,
149
respectively, is to increase detection efficiency by reducing neutron leakage and to flatten the
150
axial efficiency profile. The material used for the bottom reflector was nickel (99.5% purity,
151
Carpenter Technology Corporation, USA), and its housing material was STS304. Nickel has
152
positive properties as a neutron reflector, because it has a high-elastic-scattering cross-section
153
with a low-absorption cross-section; however, its heavy weight is a disadvantage in terms of
154
remote-handling and mechanical stability. Thus, nickel was used only for the bottom reflector,
155
whereas the top reflector was fabricated of graphite. The sizes of the top and bottom reflector
156
are 12.8 cm (D) × 17.0 cm (H) and 13.5 cm (D) × 10.0 cm (H), and their weights, 3.78 and 12.7
157
kg, respectively.
158 159
2.8. 3He neutron detectors and associated electronics
160 161
The 3He neutron detectors (RS-P4-0820-116, RE Reuter Stokes, USA) previously used in
162
the ASNC were reused in order to save costs. The 3He-filling pressure and the size of an active
163
volume are 4 atm and 1 in (D) × 20 in (H), respectively. The quenching gas, inserted into the
164
detector to prevent unnecessary avalanches by UV light, is nitrogen gas (N2). The detector body
165
is 0.8-mm-thick aluminum.
7
NIMA–D–17–00892.Rev1
166
In the former ASNC, there was a problem in extracting the 3He detectors from the HDPE
167
moderator using a manipulator, because the inner structure had been deformed over the course
168
of many years of use. As a solution, a guide tube has been installed around the 3He detector.
169
Aluminum (A6061) was used as the guide tube material owing to its neutron-absorption cross-
170
section and mechanical strength. PDT 110A electronics (Precision Data Technology, USA),
171
designed for high-radiation environments (>1 Mrad), an integrated circuit suitable for charge-
172
sensitive pre-amplification, signal shaping, leading-edge discrimination, LED driver, HV
173
filtering, and OR logic applications (summation of the logic pulses) was utilized for signal
174
processing. Overall a total of 24 neutron detectors were installed in a circular-shape pattern.
175 176
Figure 2 shows the inner structure of the upgraded ASNC, and Table 1 summarizes the components’ properties.
177
178 179
Fig. 2. Inner structure of ASNC: (a) schematic side-view and (b) photograph.
180
8
NIMA–D–17–00892.Rev1
181 182
Table 1. Properties of components. Component
Materials
Size
Weight (kg)
Housing
ID: 14.2 cm OD: 25.6 cm H: 82.2 cm ID: 49.8 cm OD: 56.4 cm H: 82.2 cm
302
Stainless steel (STS304) 4-mm-thick
448
Stainless steel (STS304) 4-mm-thick
HDPE
ID: 25.7 cm OD: 49.7 cm H: 64.0 cm
103
None
neutron HDPE
ID: 56.4 cm OD: 64.5 cm H: 82.2 cm ID: 41.4 cm OD: 49.7 cm H: 9.6 cm
57.1
None
9.1
Aluminum (A6061) 3-mm-thick
Internal gamma-ray Pb 96.8% Sn 3.2% shield External gamma-ray Pb 96.8% Sn 3.2% shield Neutron moderator
External shield
Cable enclosure
Al (A6061)
LED-based monitoring box
Stainless steel W: 16.0 cm (STS304) D: 10.0 cm H: 30.0 cm Al (A6061) D: 14.0 cm H: 48.1 cm
Sample container Neutron reflectors
Top
Graphite
Bottom
Ni
3
He detectors
4 atm of 3He (+ N2 gas)
D: 12.8 cm H: 17.0 cm D: 13.5 cm H: 10.0 cm D: 2.54 cm H: 50.8 cm
Stainless steel (STS304) 1.5-mm-thick 7.8
Aluminum (A6061) 3-mm-thick
3.78
Aluminum (A6061) 3-mm-thick 12.7 Stainless steel (STS304) 3-mm-thick Model: RS-P4-0820-116 Body: 0.8-mm-thick aluminum
183 184
2.9. Configuration
185 186
Figure 3 shows the configuration for signals, LV and HV transmission between the inside of
187
the hot-cell and the operation area. The shift register (JSR-14, Canberra, USA), DC power
188
supply (E3648A, Agilent, USA), 4-channel derandomizer (LANL, USA) [15] and PC are
189
located in the operation area. The HV is supplied to the detector from the shift register, while the
190
+12 V DC power is supplied to the electronics from the DC power supply. Although the
191
electronics requires +5 V for operation, we applied +12 V and used the DC voltage regulator
9
NIMA–D–17–00892.Rev1
192
(Aquila Technologies, USA), located immediately in front of the electronics, to convert the +12
193
V into +5 V in consideration of the voltage drop as the current travels through the long cable
194
(~20 m) between the inside and outside of the hot-cell. The signals produced in each detector are
195
summed using the integrated OR circuit in the PDT 110A electronics. In order to reduce the
196
dead time, a total of 24 detector signals were configured into four groups (instead of one). Six
197
detector signals are summed for each group; hence, the four summed signals are transmitted to
198
the derandomizer for reduction of input synchronization loss. The four group signals are
199
summed once again in the derandomizer to produce one summed signal, which is finally fed into
200
the shift register for analysis. The LED signal produced in each detector is directly connected to
201
the LED installed in the monitoring box to enable monitoring of the status of each detector in
202
the operation area through the shielded window. The JSR-14 shift register is controlled by INCC
203
(IAEA Neutron Coincidence Counting) software [16].
204
Figure 4 shows the upgraded ASNC installed in the hot-cell of the ACPF. There are only
205
three operations that need to be handled remotely by a manipulator for measurement: (1)
206
opening/closing of the cover of the sample container, (2) moving of the sample container
207
upwards and downwards, and (3) moving of the sample container leftwards and rightwards.
208
These operations were all tested successfully with the manipulators. By these remote operations,
209
the sample to be measured can be safely inserted into the sample container and moved to the
210
measurement location (i.e., inside the cavity).
10
NIMA–D–17–00892.Rev1
211 212
Fig. 3. Configuration for signals and high- and low-voltage transmission.
213
214 215
Fig. 4. ASNC installed in hot-cell with indications of remote-handling parts.
216 217
11
NIMA–D–17–00892.Rev1 218 219
3. Detector characterization
220
Various detector parameters need to be determined in order to characterize the measurement
221
system and to correct the measurement data. The parameters determined in this study were the
222
HV plateau, dead time, efficiency profile, die-away time, gate length, doubles gate fraction, and
223
stability.
224 225
3.1. High-voltage (HV) plateau
226 227
The HV plateau for each detector as well as each group was measured after applying a gain-
228
matching procedure for each detector [17]. As shown in Fig. 5, the gains of the 24 neutron
229
detectors were well matched to each other in order to have the same shape for each detector as
230
well as each group. From these HV plateaus, the operating voltage was determined to be 1720 V.
231
Additionally, the HV plateau of each group was compared with that of the summed signal finally
232
fed into the shift register from the derandomizer. It was found that the averaged count loss at the
233
operating voltage of 1720 V was ~0.21% at the count rate of ~4.33×104 cps, due to the process
234
of summing the four group signals into one in the derandomizer. The HV plateaus were
235
measured for 30 s at 20 V intervals from 1400 to 1900 V. A 252Cf calibration source (L3-693,
236
Eckert & Ziegler Isotope Products, USA) with a neutron emission rate of ~1.93×105 n/s was
237
utilized.
12
NIMA–D–17–00892.Rev1
238 239
Fig. 5. High-voltage plateaus for each detector and group after normalization (top) and for each
240
group and summed signal (bottom).
241 242
3.2. Dead time
243 244
As the count rate increases, the signal loss increases due to the detection system (i.e.,
245
detector and electronics)’s dead time. Various factors can affect the degree of dead time of a
246
3
247
diameter, shaping time, and configuration of preamps. In the ASNC, each 3He proportional
248
counter has a preamp for reduction of the dead-time effect. The signal loss due to the dead time
249
can be corrected empirically by (1) the doubles-to-singles-ratio method, (2) the paired-source
He-based neutron detector, including the 3He-filling pressure, type of quenching gas, detector
13
NIMA–D–17–00892.Rev1
250
method, or (3) the source-intensity-ratio method [18]. In this study, the dead-time correction
251
factors were determined using the source-intensity-ratio method, due to the availability of
252
calibration sources. By this method, a strong and a weak 252Cf source is measured separately; the
253
ratio of the dead-time-corrected count rates should be equal to the ratio of the source intensities.
254
The relationship between the ratios can be expressed by
255 256
(1)
257
(2)
258 259
where S and D are the dead-time-corrected singles and doubles count rates, respectively, Y is the
260
source intensity, and the subscripts s and w represent the weak and strong sources, respectively.
261
Two
252
Cf sources with neutron emission rates of ~1.92×105 n/s (L3-693) and 642.4 n/s
262
(9594NC, AEA Technology, UK) for the strong and weak sources, respectively, were used;
263
accordingly, the source intensity ratio was 299.2. In order to sufficiently reduce the random error,
264
the measurements were performed over 7952 cycles × 30 s/cycle (=66.3 h) and 240 cycles × 30
265
s/cycle (=2 h) for the weak and strong sources, respectively. The dead-time-corrected singles (Sc)
266
and doubles (Dc) count rates [19] can be calculated using the equations
267 268
(3)
269
(4)
270 271
where Sm and Dm are the measured singles and doubles rates, respectively, and δ is given by δ =
272
(A + B·10-6×Sm) μs; where A and B are constants with the relationship B=A2/4 [20]. As a result,
273
the dead-time coefficient A for singles can be calculated from Eqs. (1) and (3), and for doubles,
14
NIMA–D–17–00892.Rev1 274
from Eqs. (2) and (4). The solutions for coefficient A for singles (AS) and doubles (AD) are
275
276
(5)
277
(6)
278 279
where S and D are the singles and doubles rates, respectively, the subscripts s and w represent
280
the weak and strong sources, respectively, and k is the source intensity ratio (i.e., Ys/Yw). From
281
the measurement data and Eqs. (5) and (6), the constants A and B were determined to be 7.44 μs
282
and 13.84 μs2 for the singles rate, respectively, and 2.13 μs and 1.13 μs2 for the doubles rate,
283
respectively. These constants A and B were applied to the INCC software in order to
284
automatically correct the measurement data from the dead-time loss. The fractional count loss
285
for singles was determined to be 7.71% at the count rate of ~4.7×104 cps while 8.78% at the
286
doubles rate of ~9.2×103 cps. Table 2 summarizes the measurement data before and after dead-
287
time correction as well as the fractional count loss.
288 289
Table 2. Measured singles and doubles count rates before and after dead-time correction. Source intensity (n/s)
Singles ± 1σ
Doubles ± 1σ
Before correction (cps)
After correction (cps)
Fractional Before count loss correction (%) (cps)
After correction (cps)
Fractional count loss (%)
Weak source
642.4
156.23 ±0.03
156.28 ±0.03
0.03 ±0.03
30.87 ±0.02
30.88 ±0.02
0.03 ±0.09
Strong source
1.92×105
43149.8 ±3.09
46756.2 ±3.09
7.71 ±0.01
8427.1 ±8.35
9237.8 ±8.35
8.78 ±0.13
Ratio
299.2
276.2 ±0.06
299.2 ±0.06
273.0 ±0.32
299.2 ±0.33
290
15
NIMA–D–17–00892.Rev1 291
3.3. Efficiency profile
292 293
In order to reduce the systematic error due to the sample location within the cavity, the
294
detection efficiency profile should be as flat as possible in both the axial and radial directions.
295
To this end, we designed the ASNC for a flat efficiency profile by MCNP simulations and
296
compared the measured efficiency with the calculated data. The efficiency profiles were
297
measured using the same
298
cm (2-cm intervals) in the axial direction and from 0 to 4.5 cm (1.5-cm intervals) in the radial
299
direction. The source location (0,0) was considered to be the center of the cavity. The
300
measurements for each source location were performed over 10 cycles × 10 s/cycle. The
301
measured efficiency was determined from the dead-time corrected singles rate as divided by the
302
source intensity. The average detection efficiency for the axial and radial source positions was
303
determined to be 24.4±0.07 and 24.3±0.05%, respectively (Fig. 6). The measurement precision
304
errors were smaller than the plot symbols. Indeed, in terms of the efficiency profiles in both the
305
axial and radial directions, the ASNC showed a flat response regardless of source location;
306
therefore, the entire cavity volume can be considered to be a flat efficiency zone. Additionally,
307
the measured efficiency profiles showed excellent agreement with the calculated data, which
308
represents the accuracy of the MCNP model.
252
Cf calibration source (L3-693) at locations ranging from -10 to 10
16
NIMA–D–17–00892.Rev1
309 310
Fig. 6. Measured and calculated efficiency profiles in axial (top) and radial (bottom) directions.
311 312
3.4. Die-away time
313 314
The die-away time is the mean lifetime of a neutron within the detection system [19]. At any
315
moment, a neutron emitted from the sample can disappear from the detection system, either by
316
being absorbed in the moderator or in the neutron detector, or by escaping. It depends on a
317
variety of factors including the structure (size, shape, and presence of cadmium) and efficiency
318
of the detection system as well as the physical/chemical properties of the nuclear material to be
319
measured (scattering, moderation, and production of additional neutrons by induced fissions).
17
NIMA–D–17–00892.Rev1
320
Let us assume that the neutron population in the detection system as a function of time could be
321
expressed by the single exponential
322 323
(7)
324 325
where N(t) and N(0) are the neutron population at time t and 0, respectively, and τ is the die-
326
away time. First, the die-away time of the ASNC was evaluated by MCNP simulations for which
327
the detection probability as a function of time was calculated. After fitting to the exponential
328
decay function with a single time constant, the calculated die-away time was determined to be
329
59.0±0.28 μs (Fig. 7).
330
331 332
Fig. 7. Calculated detection probability as a function of time with fitted curve by exponential
333
decay function.
334 335
The die-away time also can be determined from the measured doubles rates with different
336
gate lengths, followed by exponential growth function curve fitting [22] or by calculation using
337
the equation [19]
18
NIMA–D–17–00892.Rev1
338
(8)
339 340
where G1 and G2 are the gate lengths with the correlation of G2=G1×2, and D1 and D 2 are the
341
doubles count rate with the gate lengths of G1 and G2, respectively. The measured doubles rates
342
with different gate lengths for a fixed pre-delay of 4.5 μs are summarized in Table 3. The
343
measurements were performed over 40 cycles × 30 s/cycle using the
344
~1.92×105 n/s). The die-away time, with curve fitting with the exponential growth function, was
345
determined to be 65.2±1.1 μs (Fig. 8). The die-away times determined from Eq. (8) are
346
summarized in Table 4. In general, the die-away time was increased as the gate interval
347
increased. It is worthwhile to note that a neutron coincidence counter might have more than one
348
die-away time according to the gate interval [18]. Considering the die-away times determined
349
from the simulation (τsimulation=59.0 μs) and from the curve fitting of the measurement data (τcurve
350
fitting=65.2
351
which validates the model. For practical purposes, we will use the value of 62.4 μs determined
352
from the doubles gate method using Eq. (8). This was calculated from the data on the gate
353
interval 45–90 μs, which fully covers the die-away times of 59.0 and 65.2 μs.
252
Cf source (L3-693,
μs), the die-away times of the ASNC from simulation and measurement are similar,
354
355 356 357
Fig. 8. Measured doubles count rates as function of gate length with curve fitting by exponential growth function. 19
NIMA–D–17–00892.Rev1 358
Table 3. Measured doubles rate and its relative error with different gate lengths. Gate length (μs)
Doubles rate ± 1σ (cps)
Relative error (%)
16
3293.1±9.1
0.276
32
5850.1±14.0
0.240
45
7432.4±17.5
0.236
64
9270.5±22.2
0.239
90
11045.8±27.9
0.252
128
12653.0±35.7
0.282
180
13888.3±43.9
0.316
256
14638.3±54.0
0.369
359 360
Table 4. Calculated die-away times from measured doubles count rates with different gate
361
intervals. Gate interval (μs)
Die-away time (μs)
16–32
63.2
32–64
59.6
45–90
62.4
64–128
63.5
90–180
66.3
128–256
69.1
362 363
3.5. Gate length
364 365
The gate length in neutron coincidence counting is the time window applied to the shift
366
register in order to record the doubles rate if two neutron pulses are received within the
367
predefined time window. The accidentals (random coincidences) are increased as the gate length
20
NIMA–D–17–00892.Rev1
368
increases, while some of the reals (true coincidences) can be lost if the gate length defined is too
369
short. The optimal gate length for coincidence counting can be determined from the measured
370
relative error on the doubles rate at different gate lengths. Figure 9 shows the relative doubles
371
errors ranging from 16 to 256 μs, as listed in Table 3. The error of the doubles rates is relatively
372
flat in the region 32–64 μs. We have decided to use the gate length of 64 μs because it is close to
373
the theoretical optimum of 1.27×die-away time (i.e., Goptimum ≈ 1.27×τ) [19].
374
375 376
Fig. 9. Measured doubles errors at different gate lengths.
377 378
3.6. Doubles gate fraction
379 380
The doubles gate fraction is the ratio between the doubles rate with the finite gate length and
381
that with the infinite gate length. It can be calculated by the equation [19], under the assumption
382
that the neutron population can be expressed by a single exponential decay,
383 384
(9)
21
NIMA–D–17–00892.Rev1
385 386
where fd is the doubles gate fraction, P is the pre-delay, τ is the die-away time, and G is the gate
387
length. Using the values of P, τ, and G determined from the previous sections, the fd was
388
determined to be 0.60. However, as mentioned above, there are more than one die-away time for
389
the coincidence counter; therefore, in order to remove the dependency on the die-away time in
390
calculation, the fd was calculated using the equation [21]
391 392
(10)
393 252
394
where νs1 and νs2 are the first and second factorial moments for the
395
(νs1 = 3.757 and νs2 = 11.962 [19]); S and D are the dead-time-corrected singles and doubles
396
rates, respectively; and ε is the detection efficiency for the 252Cf source at the center location. In
397
order to use the Eq. (10), there are requirements for the source condition: no (α,n) neutrons and
398
no multiplication, which are usually satisfied for a
399
measurement data (S = 46756.2 cps, D = 9237.8 cps, and ε=24.3%), the double gate fraction fd
400
of the ASNC was determined to be 0.51. It is worthwhile to note that the decay of the neutron
401
population in the ASNC may have multiple components to their die-away curve, instead of
402
following a single exponential decay, due to the 5-cm-thick inner gamma-ray shield and the Cd
403
sheet installed around the 3He detector for flattening the axial detector efficiency. This could be
404
one of the reasons to cause the difference between the values determined from Eq. (9) and (10).
252
Cf source, respectively,
Cf source. From this equation and the
405 406
3.7. Stability
407 408
Because the ASNC will be operated in the hot-cell environment, the stability is one of the
409
major factors in considering the cost and time required for maintenance. In order to evaluate the
22
NIMA–D–17–00892.Rev1
410
stability of the ASNC, it was tested by measuring the background for 20,000 cycles × 30 s/cycle
411
(~7 days). Figure 10 shows the measured singles and doubles backgrounds for 16,000 cycles.
412
Although more measurements than these were performed, the INCC recorded the raw data of
413
each cycle for only 16,000 cycles. The measured singles and doubles backgrounds were 0.704
414
and 0.037 cps, respectively. The relative standard deviations for singles and doubles were 0.14
415
and 0.27%, respectively. There were some spikes due to spallation events caused by cosmic rays;
416
however, these were removed from the data by the INCC software’s quality control tests.
417
The detector parameters determined in this study are summarized in Table 5.
418
419 420
Fig. 10. Background count rate of ASNC for singles (top) and doubles (bottom).
421
23
NIMA–D–17–00892.Rev1 422
Table 5. Summary of ASNC detector parameters. Parameter
Value
Cavity size (Flat efficiency zone)
13.4 cm (D)×26.0 cm (H)
Operating voltage
1720 V for Singles
A = 7.44 μs B = 13.84 μs2
for Doubles
A = 2.13 μs B = 1.13 μs2
Dead-time coefficient
Detection efficiency (for 252Cf at center of cavity)
24.3%
Die-away time
62.4 μs
Gate length
64 μs
Pre-delay
4.5 μs
Doubles gate fraction
0.51
423 424 425
4. Summary
426 427
The ACPF, a hot-cell facility in KAERI, has been refurbished for oxide-reduction-process
428
testing using spent fuels. Several process- and safeguards-related instruments were installed and
429
are being tested. In the present study, the ASNC, a safeguards neutron coincidence counter, was
430
upgraded to improve its performance in terms of remote-handling, maintenance, and neutron-
431
detection capabilities. Specifically in the latter case, detection efficiency was improved with
432
better efficiency profiles in both the axial and radial directions. Various parameters including the
433
high-voltage plateau, efficiency profile, dead time, die-away time, gate length, doubles gate
434
fraction, and stability were measured for detector characterization. Some measured parameters
435
were compared with MCNP simulation results. In general, the measurement data showed good
436
agreement with the simulation data. To the best of the authors’ knowledge, the ASNC is the only 24
NIMA–D–17–00892.Rev1
437
safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell.
438
In the near future, the upgraded ASNC will be further tested for calibration (with spent-fuel rod
439
cuts) as well as for overall performance (with input and output materials from the oxide-
440
reduction process).
441 442
Acknowledgments
443
This work was supported by a National Research Foundation of Korea (NRF) grant funded by
444
the Korean government (MSIP) (No. NRF-2017M2A8A5015084).
445 446 447
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