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Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry
Nuclear Instruments and Methods in Physics Research A 791 (2015) 93–100
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Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry Riccardo Rossa a,b, Alessandro Borella a,n, Pierre-Etienne Labeau b, Nicolas Pauly b, Klaas van der Meer a a
SCK CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol, Belgium Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels, Belgium
b
art ic l e i nf o
a b s t r a c t
Article history: Received 18 December 2014 Received in revised form 13 April 2015 Accepted 13 April 2015 Available online 24 April 2015
The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235U fission chamber, with a 3He proportional counter, and with a 10B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239Pu and 235U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3He and 10B proportional counters to increase the sensitivity to 239Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies. & 2015 Elsevier B.V. All rights reserved.
Keywords: SINRD Non-destructive assay Neutron resonance densitometry Spent fuel verification 239 Pu
1. Introduction Several non-destructive assays (NDA) are currently under research for the safeguards verification of spent fuel [1]. The spent fuel is becoming of primary concern because of the increasing quantity of material that is stored worldwide, and because the first geological repository is scheduled to become operational in the next decade [2]. The opening of the repository requires the development of a precise NDA method for the verification of a spent fuel assembly before its final disposal. As part of the international effort on the development of advanced NDA methods, the Belgian nuclear research center n
Corresponding author. Tel.: þ 32 14 33 28 44. E-mail addresses: [email protected] (R. Rossa), [email protected] (A. Borella), [email protected] (P.-E. Labeau), [email protected] (N. Pauly), [email protected] (K. van der Meer). http://dx.doi.org/10.1016/j.nima.2015.04.032 0168-9002/& 2015 Elsevier B.V. All rights reserved.
SCK CEN is carrying out the investigation of the Self-Interrogation Neutron Resonance Densitometry (SINRD, [3]). SINRD is a passive neutron technique that aims at directly quantifying the 239Pu mass in a fuel assembly by measuring the attenuation of the neutron flux in the energy region close to 0.3 eV, because 239Pu has a strong resonance in the total cross-section around that energy [4]. The neutron flux close to 0.3 eV is estimated with SINRD by calculating the difference between the total neutron counts measured with detectors covered with a foil of Gd or Cd. These nuclides were chosen because they exhibit a cutoff in the total neutron crosssection slightly below and above 0.3 eV, and for this reason they are also referred to as neutron filters in this paper. At Los Alamos National Laboratory the SINRD technique was investigated for the measurement of spent fuel under water by placing a set of fission chambers on one side of the fuel assembly to measure the spontaneous neutron emission [5–9]. However, this configuration relied on the accurate positioning of the
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R. Rossa et al. / Nuclear Instruments and Methods in Physics Research A 791 (2015) 93–100
detectors [10] and was more sensitive to the first outer fuel pin rows [5]. The approach proposed in [3] considered the fuel assembly in air surrounded by a thick slab of polyethylene to ensure neutron moderation, while the neutron detector is placed in the central guide tube of the assembly. This configuration ensures better condition for SINRD in comparison with the case of fuel stored under water, because the neutron moderation occurring within the fuel pins when the fuel assembly is stored under water washes out the reduction of the neutron flux around 0.3 eV used for the SINRD technique [3]. In addition, the dry conditions used as reference in this study can be representative of the encapsulation plant foreseen to be built in Sweden over the coming years [11]. The definition of the SINRD signature and the approach used in this study for its determination are first described in the paper. Then the optimization of the filters thicknesses used for the selection of the 239Pu resonance region is presented, together with the method for the estimation of the total neutron counts and sensitivity to 239Pu. Several combinations of filters and detector types are compared and recommendations are given at the end of the paper on the selection of the measurement approach for the SINRD technique.
2. Determination of the SINRD signature 2.1. Definition of the SINRD signature To quantify the attenuation of the neutron flux around the 0.3 eV energy region the so-called SINRD signature, defined in [3] as the ratio between the energy-integrated values of the neutron flux in the fast region (FAST, i.e. 0.1–20 MeV) and in the region close to the resonance (RES, i.e. 0.2–0.4 eV), is used. SINRD ¼
FAST RES
ðaÞ
As shown in [3], the SINRD signature is correlated to the 239Pu amount in the fuel assembly model. To take into account the characteristics of detector types and neutron filters, the SINRD signature is estimated in this paper as the ratio between the neutron counts in the fast and in the resonance energy regions. Therefore, the response of neutron detectors in the central guide tube of the fuel assembly was modelled; the fast neutron counts were determined for a bare 238U fission chamber, while a 239Pu fission chamber covered with filters was the chosen detector to estimate the neutron counts close to the 0.3 eV resonance.
is valid for a parallel neutron beam (φin) perpendicular to a slab of material. The neutron beam that traverses the sample without any interaction (φtr) is given by: ! X nk σ D φtr ¼ φin U exp ðbÞ tot;k k
where σ D tot;k is the Doppler broadened total cross-section and nk is the number of atoms per unit area of nuclide k. Therefore the transmitted flux is directly related to the total cross section and the areal density of the nuclides present in a sample [13]. The integral neutron counts in the resonance region (RES) per atom of active material in the detector and per emitted neutron were then determined as: Z RES ¼ σ det Uφin U ½expð nGd σ Gd Þ expð nCd σ Cd ÞdE ðcÞ where σdet is the cross-section of the active material in the detector, E is the energy of the incoming neutron, nGd and nCd are the number of atoms per unit area of the neutron filters, while σGd and σCd are their corresponding total cross-sections. The ideal detector for the estimation of the RES integral is a 239Pu-loaded fission chamber [4] and therefore the σdet considered for this case was the fission cross-section for 239Pu. The RES integral value was calculated over the energy region between 10 9 and 20 MeV, with the values for the cross-sections of detectors and filters determined as explained in [14]. This approach for the calculation requires only a Monte Carlo simulation to determine the energy distribution of the neutron flux in the guide tube without filters. Then, by using the transmission factors for several filter thicknesses the influence of the neutron absorbers was calculated without running additional simulations. However, the formulas used for these calculations are valid only in the case of a parallel neutron beam and a perpendicular thin neutron absorber. Therefore, the effect of the neutron filters will be re-evaluated in the future if the measurement conditions will significantly differ from these assumptions. The difference calculated in the RES integral is motivated by the fact that the Gd filters have a lower cutoff energy compared to the Cd filters; therefore, the transmitted flux using a Gd filter is generally higher than the value calculated through a Cd filter. However, an effect is visible in Fig. 1, which plots the transmission factors of Formula (c) in the case of 0.5 mm Gd, 0.6 mm Cd filters and their corresponding difference. Due to the combination of thick Gd and thin Cd filters significant negative contributions to the integral difference can be obtained in the epithermal energy 1.0
2.2. Calculation of the SINRD signature
0.8 0.6
Transmission factor
The SINRD signature was calculated in this study using a Monte Carlo model of a PWR 17 17 fuel assembly in air surrounded by a slab of 12 cm of polyethylene on all sides. The Monte Carlo code MCNPX v.2.7.0 [12] was used to obtain the energy distribution of the neutron flux in the central guide tube of the fuel assembly, by simulating 108 particles and calculating the so-called F4 flux tally in 600 logarithmically interpolated energy bins from 10 9 to 20 MeV. The neutron flux was determined in an empty cavity that can host the selected detector type and it was normalized per starting particle. The influence of the filters and the response of each specific detector were then estimated according to the approach explained in Section 3. The SINRD signature requires the neutron counts in two different energy regions and the selection of the region close to the 0.3 eV resonance is achieved by using specific neutron absorbers. In order to determine the impact of a neutron absorber on the neutron flux we used the following analytical approach, which
0.4 0.2 0.0 -0.2 -0.4 -0.6
Gd 0.5 mm Cd 0.6 mm Difference
-0.8 -1.0 -9
10
-8
10
-7
10
-6
10
-5
10
-4
10
-3
10
-2
10
-1
10
0
10
1
10
Neutron energy (MeV) Fig. 1. Negative contributions due to the difference between the transmitted fluxes through Gd and Cd filters. The values shown in the plot are the transmission factors calculated with Formula (b).
R. Rossa et al. / Nuclear Instruments and Methods in Physics Research A 791 (2015) 93–100
region and this effect must be accounted for in the selection of the filters thickness. The total neutron counts in the fast energy region are the second quantity needed for the calculation of the SINRD signature. This value is calculated according to the following formula: Z FAST ¼ σ U Uφin dE ðdÞ where σU is the fission cross-section of 238U and φin is the neutron flux calculated in the guide tube of the fuel assembly.
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time (t), while the uncertainty associated to this quantity (σGd, σCd, σF) is estimated as the square root of the total count. The uncertainty of the total count in the resonance region (σR) takes into account the uncertainty propagation for the total counts estimated for the detectors covered with the neutron filters. Therefore, the relative uncertainty of the SINRD signature (σS,rel) due to counting statistics can be calculated as " # σ2 σ 2R σ 2R cF t 1 1 cGd þ cCd ¼ þ ¼ þ σ 2S;rel ¼ F2 þ t cF ðcGd cCd Þ2 C F ðC Gd C Cd Þ2 c2F t 2 ðcGd cCd Þ2 t 2 ðfÞ
3. Optimization of the filters and detectors for the SINRD technique 3.1. Selection criteria for thicknesses of the neutron absorbers The ideal configuration for the measurement of the neutron flux close to the 0.3 eV resonance is to have the RES integral sensitive to neutrons with energy within the 0.2–0.4 eV region and to minimize the contributions from neutrons with different energy. Therefore, two criteria based on the energy distribution of the transmitted fluxes were chosen. The first criterion was the ratio between RES integrated over the 0.2–0.4 eV energy window and RES integrated over the whole energy range. The second criterion was the ratio between the integral value of the negative contributions to RES and the absolute integral value, defined as the absolute sum of positive and negative contributions. In order to select the optimal configuration for SINRD, the first criterion should be maximized and the second criterion should be minimized. The energy distribution of the neutron flux calculated in the guide tube of the fuel assembly varies with the fuel composition because of the neutron absorption of the nuclides contained in the fuel composition. Therefore, also the values of the criteria used for the filters optimization are influenced by the 239Pu content because of the significant neutron absorption in the resonance region. In order to evaluate this effect the criteria for the selection of the optimal filter thickness were calculated for a 239Pu fission chamber in case of fuel without fissile content and with 1% of 239 Pu. This 239Pu content was chosen as extreme to cover the 239Pu content usually present in the LEU spent fuel assemblies [15]. Detailed fuel compositions were not included at this stage in the simulations to avoid influence from other nuclides apart from 239 Pu. 3.2. Assessment of the detectors total counts and associated uncertainties Based on the approach of [16], the count rate of a neutron detector (c) was estimated with the following formula: Z c ¼ N tot U n U σ det U φin U exp nf il σ f il dE ðeÞ where Ntot is the total neutron emission from the spent fuel assembly, n is the mass of active material in the detector, nfil is the number of atoms per unit area of the neutron filter, and σfil is its corresponding total cross-section. Since no filter is used for the estimation of the fast neutron flux, in that case the exponential term of Formula (e) is not considered. By using this approach, the energy distribution of the incoming flux φin is estimated by means of Monte Carlo simulations accounting for the angular and spatial distribution of the neutron flux, instead of using leakage probabilities for different energy regions as described in [7]. The total count (CGd, CCd, CF) of each measurement is the product between the count rate (cGd, cCd, cF) and the measurement
The detectors considered for this estimate contained 100 mg of active material (i.e. 239Pu, 235U, or 10B), or 31.63 mg of 3He (corresponding to a tube with 1.27 cm diameter and 50 cm length filled with 4 atm gas). The fissile amount in the fission chambers is similar to the equipment used for the Fork detector [16], while the pressure value of the 3He tube was taken from [17]. The fission chamber for the estimation of the fast neutron flux contained 1000 mg of 238U, in order to compensate for the small fission cross-section of 238U. The amount of active materials in the detectors was the only parameter of the detectors that was used to calculate the total neutron counts, since the detectors were not modelled in the simulations. Future work will refine this analysis and consider dimensions of detectors that can actually be inserted in a PWR guide tube. The total neutron emission of the spent fuel was calculated from the results of the reference spent fuel library developed at SCK CEN [18] and considered fuel with 3.5% initial enrichment, 5 years of cooling time, and discharge burnup ranging from 5 to 60 GWd/tU. The measurement time was set to 60 min for all calculations.
4. Results with a
239
Pu fission chamber
4.1. Selection criteria for the filter thickness A plutonium-loaded fission chamber is considered as the ideal detector for SINRD in order to increase the detection of neutrons in the energy region close to 0.3 eV [4]. The selection criteria for the filters thicknesses were evaluated for this detector type in the case of fuel without 239Pu and with 1% 239Pu content. Tables 1 and 2 show the two selection criteria calculated for several filter combinations. The share of the resonance region reported in Table 1 increases with thick Gd and thin Cd filters, while the opposite trend is shown in Table 2 for the second criterion. The significant negative contributions calculated in Table 2 are also the reason of the values higher than 100% shown in Table 1. The uncertainties reported in the tables take into account only the statistical uncertainty of the Monte Carlo simulations. A significant difference is observed in the values of both criteria when comparing the results from Tables 3 and 4 (fuel with 1% 239Pu) with the cases shown in Tables 1 and 2 (fuel without 239Pu). The Table 1 First criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel without fissile material. The criterion is the ratio between the RES integrated in the 0.2–0.4 eV energy window and the RES integral over the complete energy spectrum. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
73.2 7 0.6 72.7 7 0.5 72.6 7 0.5 72.6 7 0.5 72.5 7 0.5 72.4 7 0.5
87.0 7 0.8 85.6 7 0.7 85.0 7 0.7 84.7 7 0.7 84.4 7 0.7 84.2 7 0.6
93.5 70.9 91.2 70.8 90.2 70.8 89.6 70.8 89.1 70.7 88.7 70.7
97.9 7 1.1 94.4 7 0.9 93.0 7 0.9 92.2 7 0.8 91.5 7 0.8 90.9 7 0.8
101.9 7 1.3 97.0 7 1.1 95.0 7 1.0 93.8 7 0.9 92.9 7 0.9 92.2 7 0.8
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Table 2 Second criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel without fissile material. The criterion is the share of negative contributions to the RES integral over the absolute RES integral value. Share NEG (%) Gd 0.1 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
1.3 7 0.4 0.4 7 0.4 0.17 0.4 o 0.17 0.4 o 0.17 0.4 o 0.17 0.4
Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm
2.3 7 0.5 0.9 7 0.5 0.4 7 0.5 0.3 7 0.5 0.2 7 0.5 0.2 7 0.5
3.5 7 0.6 1.5 7 0.6 0.8 7 0.6 0.6 7 0.6 0.5 7 0.5 0.5 7 0.5
4.9 7 0.7 2.2 7 0.7 1.4 7 0.6 1.17 0.6 1.0 7 0.6 0.9 7 0.6
6.7 7 0.7 3.2 7 0.7 2.17 0.7 1.7 7 0.6 1.5 7 0.6 1.4 7 0.6
Table 3 First criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel containing 1% of 239Pu. The definition of the criterion is the same of Table 1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm 1% 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
43.4 7 0.8 42.7 7 0.7 42.4 7 0.7 42.17 0.7 41.8 7 0.7 41.5 7 0.6
61.8 7 1.4 58.9 7 1.2 57.3 7 1.1 56.0 7 1.1 54.9 7 1.0 53.9 7 1.0
74.5 7 2.1 68.37 1.7 65.17 1.5 62.8 7 1.4 60.9 7 1.3 59.2 7 1.2
88.2 73.3 76.5 72.3 71.0 71.9 67.4 71.7 64.5 71.5 62.2 71.4
110.6 75.8 87.1 73.3 77.8 72.5 72.1 72.1 68.0 71.8 64.7 71.6
difference in the values for the first criteria is due to the neutron absorption of 239Pu in the 0.3 eV resonance region. The presence of 239 Pu in the fuel composition leads to a decrease in the RES integral value because of resonant absorption, which finally leads to lower ratios in Table 3. The large share of negative contributions calculated with the second criterion (Table 4) is also related to the reduction of the total neutron flux due to the absorptions. Moreover, the reduced RES integral value obtained by adding 239Pu in the fuel composition leads to a higher sensitivity of the criteria to the Cd thickness compared to the results for fuel without fissile content. 4.2. SINRD signature Another parameter for the selection of the filters thickness is the value of the SINRD signature that can be achieved by each combination. Table 5 shows in the case of a 239Pu fission chamber the SINRD signature for several fuel compositions and filter combinations. The values are normalized to the SINRD signature calculated without fissile material in the fuel composition. The uncertainties of the SINRD signature deriving from the statistical uncertainty of the Monte Carlo simulations are smaller than 0.6%. The SINRD signature increases with the fuel burnup because of the increased quantity of 239Pu, and it increases with the thickness of the Gd filter. On the other hand, the SINRD signature decreases by increasing the thickness of the Cd filter, therefore the highest values are obtained in the case with Gd filter of 0.3 mm and Cd filter of 0.5 mm. However, the final choice of the filters must also take into account the total neutron counts achieved in a measurement as calculated in the next section.
Table 4 Second criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel containing 1% of 239Pu. The definition of the criterion is the same of Table 2. Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm 1% 239Pu Cd Cd Cd Cd Cd Cd
2.4 7 0.8 1.0 7 0.8 0.6 7 0.8 0.5 7 0.7 0.4 7 0.7 0.3 7 0.7
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
6.2 7 1.2 3.4 7 1.1 2.5 7 1.1 2.17 1.0 1.9 7 1.0 1.7 7 1.0
11.1 71.5 6.9 71.4 5.3 71.3 4.5 71.2 4.1 71.1 3.7 71.1
17.2 7 1.8 11.4 7 1.6 9.17 1.5 7.8 7 1.3 7.0 7 1.3 6.4 7 1.2
Table 5 SINRD signature for several filter combinations in the case of a chamber.
25.17 2.1 17.2 7 1.9 13.8 7 1.6 11.9 7 1.5 10.6 7 1.4 9.7 7 1.3
239
Pu fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 5 10 15 20 40 60
2.12 2.93 3.52 3.94 4.52 4.55
2.07 2.84 3.38 3.76 4.29 4.32
2.28 3.27 4.00 4.52 5.27 5.30
2.18 3.06 3.68 4.12 4.74 4.77
2.37 3.48 4.30 4.91 5.77 5.82
2.22 3.14 3.80 4.26 4.91 4.94
Table 6 Estimation of the total counts and relative uncertainties in the case of a 238U fission chamber. Fuel BU (GWd/tU)
Counts
Unc. (%) 1
5.84 10 2.58 102 1.09 103 4.03 103 9.12 104 4.47 105
5 10 15 20 40 60
13.1 6.2 3.0 1.6 0.3 0.1
The total counts for different filter combinations were calculated with the Formula (e) given in Section 3.2 and the results are shown in Table 7. As for the fast neutron flux, for all filters combinations the counts significantly increase in the case of fuel with high burnup because of its high neutron emission. Moreover, the total counts also increase by decreasing the filter thickness due to the reduced neutron absorption by the material. The relative uncertainty associated to the SINRD signature was estimated for different filter combinations according to Section 3.2. The results are included in Table 8 and show decreasing values by increasing fuel burnup, or by combining thin Gd filters with thick Cd filters. The results shown in this section are in line with previous work [3], where a Gd thickness of 0.3 mm was identified to maximize the SINRD signature. However, considering that all cases shown in Table 5 have high sensitivity to the 239Pu content, the combination of 0.1 mm Gd filter and 1.0 mm Cd filter is suggested to decrease the measurement uncertainty.
235
U fission chamber
4.3. Detector total counts
5. Results with a
The fast neutron counts together with the relative uncertainty were estimated according to Section 3.2 and the results are shown in Table 6. The total counts increase with the discharge burnup because of the increasing neutron emission of the fuel assembly. The values shown in Table 6 are used for the estimation of the overall measurement uncertainty for all detectors considered in this paper.
5.1. Selection criteria for the filter thickness Although a plutonium-loaded fission chamber is considered as the ideal detector for SINRD, previous work from [5] mentioned the difficulty to obtain such detector type. Therefore a 235U fission chamber was considered in this paper as alternative. Following the
R. Rossa et al. / Nuclear Instruments and Methods in Physics Research A 791 (2015) 93–100
Table 7 Estimation of the total counts and relative uncertainties in the case of a Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
239
97
Pu fission chamber.
Gd 0.2 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
5.79 102 2.06 103 7.80 103 2.69 104 5.62 105 2.74 106
4.2 2.2 1.1 0.6 0.1 0.1
4.38 102 1.58 103 6.08 103 2.11 104 4.46 105 2.18 106
4.8 2.5 1.3 0.7 0.1 0.1
3.72 102 1.37 103 5.32 103 1.86 104 3.97 105 1.93 106
5.2 2.7 1.4 0.7 0.2 0.1
2.18 102 9.06 102 3.72 103 1.35 104 2.98 105 1.46 106
6.8 3.3 1.6 0.9 0.2 0.1
1.75 102 7.53 102 3.14 103 1.15 104 2.57 105 1.26 106
7.6 3.6 1.8 0.9 0.2 0.1
Table 8 Estimation of the SINRD signature relative uncertainty in the case of a 239Pu fission chamber.
Table 10 Second criterion for the filter optimization considering 235U as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1.
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm
Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu
5 10 15 20 40 60
15.2% 7.8% 4.0% 2.2% 0.5% 0.2%
14.7% 7.4% 3.8% 2.0% 0.4% 0.2%
17.5% 9.7% 5.2% 2.9% 0.7% 0.3%
16.1% 8.5% 4.5% 2.5% 0.6% 0.3%
20.5% 12.0% 6.7% 3.8% 0.9% 0.4%
17.7% 9.7% 5.2% 2.9% 0.7% 0.3%
Table 9 First criterion for the filter optimization considering 235U as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
42.7 7 0.3 41.17 0.2 40.6 7 0.2 40.4 7 0.2 40.3 7 0.2 40.17 0.2
70.57 0.7 63.7 7 0.6 61.5 7 0.5 60.3 7 0.5 59.5 7 0.5 58.7 7 0.4
92.9 7 1.4 77.6 7 0.9 72.9 7 0.8 70.57 0.7 68.87 0.7 67.4 7 0.7
119.9 7 2.8 89.17 1.5 80.9 7 1.2 77.0 7 1.0 74.4 7 0.9 72.2 7 0.9
171.6 7 7.2 103.0 7 2.4 89.0 7 1.7 82.9 7 1.4 78.9 7 1.3 75.8 7 1.1
approach described in Section 2.2, the 235U fission cross-section was used to estimate the detector response. In order to compare the performance of different detectors, the same criteria used for the filter optimization were calculated for the filter thicknesses considered with the 239Pu fission chamber and the results are summarized in Tables 9 and 10. As a general trend the values of both criteria are inferior for the 235U fission chamber compared to the results shown in Section 4.1. 5.2. SINRD signature Table 11 shows the SINRD signature calculated in the case of a U fission chamber, and the values are normalized to the SINRD signature obtained without fissile material in the fuel composition. We observe the same trends of the signature with fuel burnup and filters thickness described in Section 4.2; however, there is a significant decrease in the values compared to the results with the 239Pu fission chamber. The uncertainties of the SINRD signature taking into account the statistics of the Monte Carlo simulations are smaller than 0.5%. 235
5.3. Detector total counts The performance of a 235U fission chamber was estimated also in terms of the total counts that can be achieved with this detector type (Table 12). The approach described in Section 3.2 was used for
Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
5.2 7 0.3 1.7 7 0.3 0.6 7 0.3 0.3 7 0.3 0.2 7 0.3 0.1 70.3
11.7 7 0.5 4.6 7 0.6 2.2 7 0.6 1.4 7 0.6 1.1 70.5 1.0 7 0.5
18.4 70.8 8.2 70.8 4.7 70.8 3.4 7 0.7 2.8 70.7 2.5 70.7
25.7 7 1.1 12.8 7 1.0 8.0 7 1.0 6.17 0.9 5.3 7 0.9 4.8 7 0.8
Table 11 SINRD signature for several filter combinations in the case of a chamber.
34.17 1.5 18.5 7 1.1 12.3 7 1.2 9.8 7 1.1 8.5 7 1.1 7.7 7 1.0
235
U fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 5 10 15 20 40 60
1.77 2.25 2.58 2.82 3.14 3.15
1.71 2.15 2.44 2.64 2.91 2.92
1.93 2.54 2.95 3.24 3.64 3.65
1.77 2.23 2.53 2.73 3.00 3.01
2.12 2.90 3.45 3.84 4.39 4.41
1.79 2.26 2.56 2.76 3.02 3.03
this calculation. The total counts are significantly lower than in the case of the 239Pu fission chamber, because of the lower fission cross-section of 235U compared to 239Pu. This leads also to an increase in the measurement uncertainty (Table 13). In the case of 235U fission chamber, the use of thin Gd filters is suggested in order to decrease the measurement uncertainty. Moreover, by choosing a thin Gd filter there is a reduction of the negative contributions to the absolute RES integral value (Table 10), while keeping still significant contributions from the resonance region (Table 9). Considering the low total counts shown for all cases in Table 12, the selection of 0.1 mm Gd filter and 1.0 mm Cd filter is suggested to increase the total counts while keeping a significant sensitivity to 239Pu.
6. Results with proportional counters 6.1. Selection criteria for the filter thickness Other detector types simulated for the SINRD technique were two proportional counters. As for the detector types considered before, the approach explained in Section 2.2 was used to determine the detector response, and either the absorption cross-section of 3He or the (n,α) reaction cross-section of 10B were chosen for the calculation. However, to limit the length of the paper, only results
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Table 12 Estimation of the total counts and relative uncertainties in the case of a Filter Fuel BU GWd/tU 5 10 15 20 40 60
Gd 0.1 mm
235
U fission chamber.
Gd 0.2 mm
Gd 0.3 mm
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
2.30 102 9.34 102 3.79 103 1.36 104 3.00 105 1.47 106
6.6 3.3 1.6 0.9 0.2 0.1
1.86 102 7.80 102 3.22 103 1.17 104 2.60 105 1.27 106
7.3 3.6 1.8 0.9 0.2 0.1
1.72 102 7.28 102 3.02 103 1.10 104 2.47 105 1.21 106
7.6 3.7 1.8 1.0 0.2 0.1
1.55 102 6.75 102 2.84 103 1.04 104 2.34 105 1.15 106
8.0 3.8 1.9 1.0 0.2 0.1
1.41 102 6.22 102 2.63 103 9.67 103 2.18 105 1.07 106
8.4 4.0 2.0 1.0 0.2 0.1
235
U fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 29.4% 16.7% 9.1% 5.1% 1.2% 0.5%
25.4% 14.1% 7.5% 4.2% 0.9% 0.4%
60.9% 37.0% 20.7% 11.8% 2.7% 1.2%
42.2% 24.5% 13.3% 7.4% 1.7% 0.8%
111.4% 71.4% 40.9% 23.7% 5.6% 2.6%
59.8% 35.1% 19.1% 10.7% 2.4% 1.1%
Table 14 First criterion for the filter optimization considering 3He as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
Cd 1.0 mm
Counts
Table 13 Estimation of the SINRD signature relative uncertainty in the case of a chamber.
5 10 15 20 40 60
Cd 0.5 mm
38.4 7 0.2 37.2 7 0.2 36.8 7 0.2 36.6 7 0.2 36.4 7 0.2 36.3 7 0.2
65.2 7 0.6 59.2 7 0.5 57.17 0.4 55.9 7 0.4 55.0 7 0.4 54.3 7 0.4
88.6 7 1.3 73.9 7 0.9 69.2 7 0.7 66.77 0.7 64.97 0.6 63.4 7 0.6
119.5 7 2.8 87.3 7 1.4 78.6 7 1.1 74.2 7 1.0 71.2 7 0.9 68.87 0.8
188.4 78.8 105.0 72.5 88.8 71.7 81.6 71.4 76.9 71.2 73.3 71.1
with 3He as active material are shown, since the performance of both detectors were almost identical but the total counts for 3He were higher than the values for 10B. Both detector types are not traditionally used for spent fuel measurements because of the sensitivity to γ radiation. Therefore the results contained in this section are meant only to offer an optimal performance of the detectors, while more detailed considerations on the effective applicability to field measurements are not tackled. The results obtained with the 3He detector confirm the reduction of the performance of other detector types compared to the 239 Pu fission chamber. In fact, the values of the criteria obtained with the 3He proportional counter (Tables 14 and 15) are similar to the case of 235U fission chamber (Tables 9 and 10). 6.2. SINRD signature The values of the SINRD signature for a 3He proportional counter are reported in Table 16 and show similar results to the case of a 235U fission chamber. As for the other detectors, the values are normalized to the case of fuel without fissile content. The statistical uncertainties of the Monte Carlo simulations are lower than 0.5%. 6.3. Detector total counts Following the approach described in Section 3.2, the total counts achievable with a 3He proportional counter were
Table 15 Second criterion for the filter optimization considering 3He as active material in the detector and fuel withot fissile material. The criterion is defined in Section 3.1. Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
4.9 7 0.3 1.8 7 0.3 0.8 7 0.3 0.5 7 0.3 0.4 7 0.3 0.4 7 0.3
11.8 7 0.5 5.2 7 0.5 3.0 7 0.5 2.3 7 0.5 2.0 7 0.5 1.8 7 0.5
19.17 0.8 9.6 7 0.7 6.2 7 0.7 5.0 7 0.7 4.4 7 0.7 4.17 0.7
27.3 7 1.0 15.0 7 0.9 10.4 7 0.9 8.5 7 0.9 7.6 7 0.8 7.17 0.8
36.5 7 1.9 21.7 7 1.1 15.6 7 1.1 13.0 7 1.0 11.6 7 1.0 10.7 7 0.9
estimated. By looking at Table 17, the advantage of using 3He as active material is visible on the increase of the total counts in the measurements compared to the previous detector types. Given the large cross-section of 3He, this detector type offers the best results in terms of measurement uncertainty (Table 18). The total counts obtained for a 10B proportional counter are lower than the case with the 3He proportional counter, but significantly higher than the cases with the fission chambers. This result is due to the fact that the (n,α) cross-section for 10B is slightly smaller than the absorption cross-section of 3He. Considering the low measurement uncertainty achieved by both proportional counters, a thick Gd filter (e.g. 0.3 mm) and a thin Cd filter (e.g. 0.5 mm) are proposed to increase the sensitivity to 239 Pu.
7. Discussion of the results The results presented in this work show the need for an optimization process taking into account total neutron counts, filters thickness, and detector size to develop a SINRD detector that can meet the requirements for an efficient verification of spent fuel assemblies. In order to increase the total counts of the measurements, several methods can be used. A straightforward approach is to increase the measurement time, but this is in contrast with the goal of obtaining an efficient detector that can be used during an inspection. An alternative solution would be to increase the size of the detector, but this is limited by the reduced size of the guide tube that hosts the instrument. Another method would be to select thin Gd filters and thick Cd filters, but this also is constrained by the considerations about the optimal range for the filters thickness. Therefore, for the 239Pu and 235U fission chambers the use of a thin Gd filter and a thick Cd filter (e.g. 0.1 mm and 1.0 mm respectively) is suggested in order to increase the total counts. On the contrary, taking advantage of the high neutron sensitivity of the 3He and 10B proportional counters, the combination of a thick Gd filter and a thin Cd filter (e.g. 0.3 mm and 0.5 mm respectively) is proposed to maximize the SINRD signature.
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Table 16 SINRD signature for several filter combinations in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU)
Gd 0.1 mm
Gd 0.1 mm
Gd 0.2 mm
Gd 0.2 mm
Gd 0.3 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
1.72 2.18 2.50 2.72 3.02 3.04
1.67 2.08 2.35 2.54 2.80 2.81
1.87 2.43 2.82 3.09 3.47 3.48
1.72 2.14 2.42 2.61 2.86 2.87
2.06 2.81 3.35 3.73 4.28 4.31
1.74 2.17 2.45 2.63 2.88 2.90
5 10 15 20 40 60
Table 17 Estimation of the total counts and relative uncertainties in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
Gd 0.2 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (I%)
Counts
Unc. (%)
Counts
Unc. (%)
4.66 104 1.85 105 7.41 105 2.64 106 5.74 107 2.81 108
0.5 0.2 0.1 0.1 o0.1 o0.1
3.49 104 1.43 105 5.85 105 2.11 106 4.46 107 2.28 108
0.5 0.3 0.1 0.1 o 0.1 o 0.1
3.09 104 1.29 105 5.32 105 1.93 106 4.28 107 2.10 108
0.6 0.3 0.1 0.1 o0.1 o0.1
2.68 104 1.16 105 4.85 105 1.78 106 3.98 107 1.95 108
0.6 0.3 0.1 0.1 o0.1 o0.1
2.32 104 1.02 105 4.30 105 1.58 106 3.56 107 1.74 108
0.7 0.3 0.2 0.1 o 0.1 o 0.1
Table 18 Estimation of the SINRD signature relative uncertainty in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
Gd 0.1 mm
Gd 0.2 mm
Gd 0.2 mm
Gd 0.3 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
13.2% 6.3% 3.1% 1.6% 0.3% 0.2%
13.1% 6.3% 3.0% 1.6% 0.3% 0.2%
13.4% 6.5% 3.2% 1.7% 0.4% 0.2%
13.2% 6.3% 3.1% 1.6% 0.3% 0.2%
14.4% 7.3% 3.7% 2.0% 0.4% 0.2%
13.4% 6.5% 3.2% 1.7% 0.4% 0.2%
Each detector type presented in this paper shows both positive and negative characteristics. The 239Pu fission chamber ensures the best sensitivity to 239Pu but its use is hindered by its unavailability on the market. On the other hand, the 235U fission chamber shows low SINRD signature but it benefits from extensive experience for spent fuel verification. Similar SINRD signatures are obtained with the proportional counters together with very high neutron sensitivity, but the influence of gamma radiation on these detectors must be evaluated. Therefore the development of the SINRD technique should take into account the results obtained from the optimization of the filter thicknesses, and it should be supported by considerations on detectors design, measurement time, and sensitivity to gamma radiation.
8. Conclusions The performance of several detector types for the SINRD technique was compared in this paper by calculating the SINRD signature and detector total counts for different filter thicknesses. In the case of the 239Pu fission chamber the influence of the neutron filters was assessed both in the case of fuel without fissile content and for fuel with high 239Pu concentration, and a general degradation of the results was observed in the second case due to the presence of 239Pu. Significant effort was devoted to the comparison of the 239Pu fission chamber with other technologies. By looking at the
selection criteria used for the definition of the optimal filter thickness, it was concluded that the alternative detectors do not ensure the same performance of the 239Pu fission chamber. This is also reflected in the sensitivity of each detector to the 239Pu content; the total 239Pu was correlated with the SINRD signature and the 239Pu fission chamber showed significantly better results. For all detector types it was found that the SINRD signature increases with the fuel burnup and with the Gd thickness, while it decreases with the Cd thickness. The comparison among the detectors also included an estimation of the total counts that can be achieved with each detector, and the 3He and 10B proportional counters showed the largest total counts because of their large neutron sensitivity. Therefore, by considering the SINRD signature and total neutron counts achievable for each detector type, thin Gd and thick Cd filters are suggested in case of 239Pu and 239U fission chambers to maximize the total neutron counts, while thick Gd and thin Cd filters are advised for 3He and 10B proportional counters to maximize the SINRD signature.
Acknowledgements This work is sponsored by GDF SUEZ in the framework of the cooperation agreement CO-90-07-2124 between SCK CEN and GDF SUEZ.
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Refeences [1] S.J. Tobin, et al., Prototype development and field trials under the next generation safeguards initiative spent fuel non-destructive assay project, in: Proceedings of the 35th ESARDA Symposium, 2013. [2] V. Ryhänen, Progress in the implementation of nuclear waste disposal in Finland, in: 8th European Nuclear Energy Forum, 2013. [3] R. Rossa, A. Borella, K. van der Meer, Annals of Nuclear Energy 75 (2015) 176. [4] H.O. Menlove, C.D. Tesche, M.M. Thorpe, R.B. Walton, Nuclear Applications 6 (1969) 401–408. [5] A.M. LaFleur, Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel, (Ph.D. dissertation)Nuclear Engineering, Texas A&M University (2011) , August. [6] A.M. LaFleur, W.S. Charlton, H.O. Menlove, M.T. Swinhoe, Nuclear Instruments and Methods Section A 680 (2012) 168–178. [7] A.M. LaFleur, W.S. Charlton, H.O. Menlove, M.T. Swinhoe, Nuclear Science Engineering 171 (2012) 175–191. [8] A.M. LaFleur, W.S. Charlton, H.O. Menlove, M.T. Swinhoe, Nuclear Technology 181 (2013) 354–370. [9] A.M. LaFleur, H.O. Menlove, Nuclear Instruments and Methods Section A 781 (2015) 86–95.
[10] J. Hu, H.R. Trellue, S.J. Tobin, T.J. Ulrich, A.M. LaFleur, C.R. Freeman, H.O. Menlove, M.T. Swinhoe, Journal of Nuclear Materials Management XL 3 (2012) 36–44. [11] A. Nyström, Encapsulation plant at ForsmarkSKB Rapport R 200707–36. [12] D. Pelowitz (Ed.), Los Alamos National Laboratory report2011 LA-CP-11-00438. [13] P. Schillebeeckx, A. Borella, F. Emiliani, G. Gorini, W. Kockelmann, S. Kopecky, C. Lampoudis, M. Moxon, E. Perelli Cippo, H. Postma, N.J. Rhodes, E.M. Schooneveld, C. Van Beveren, Neutron resonance spectroscopy for the characterization of materials and objects, in: Proceedings of the 2nd International Workshop on Fast Neutron Detectors and Applications, 2011. [14] R. Rossa, A. Borella, K. van der Meer, P.E. Labeau, N. Pauly, Optimization of the filters thickness for the SINRD technique applied to spent fuel verification, in: Proceedings of the 55th INMM Annual Meeting, 2014. [15] J.L. Conlin, et al., Spent Fuel Library Report, LA-UR10-07894Los Alamos National Laboratory 2010. [16] A. Borella, R. Rossa, K. van der Meer, Annals of Nuclear. Energy 62 (2013) 224–230. [17] D. Reilly, N. Ensslin, H. Smith Jr., S. Kreiner (Eds.), Los Alamos National Laboratory Report1991. [18] R. Rossa, A. Borella, K. van der Meer, ESARDA Bulletin 49 (2013) 27–39.
Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
Neutron absorbers and detector types for spent fuel verification using the self-interrogation neutron resonance densitometry Riccardo Rossa a,b, Alessandro Borella a,n, Pierre-Etienne Labeau b, Nicolas Pauly b, Klaas van der Meer a a
SCK CEN, Belgian Nuclear Research Centre, Boeretang, 200, B2400 Mol, Belgium Université libre de Bruxelles, Ecole polytechnique de Bruxelles, Service de Métrologie Nucléaire (CP 165/84), Avenue F.D. Roosevelt, 50, B1050 Brussels, Belgium
b
art ic l e i nf o
a b s t r a c t
Article history: Received 18 December 2014 Received in revised form 13 April 2015 Accepted 13 April 2015 Available online 24 April 2015
The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive non-destructive assay (NDA) technique that is proposed for the direct measurement of 239Pu in a spent fuel assembly. The insertion of neutron detectors wrapped with different neutron absorbing materials, or neutron filters, in the central guide tube of a PWR fuel assembly is envisaged to measure the neutron flux in the energy region close to the 0.3 eV resonance of 239Pu. In addition, the measurement of the fast neutron flux is foreseen. This paper is focused on the determination of the Gd and Cd neutron filters thickness to maximize the detection of neutrons within the resonance region. Moreover, several detector types are compared to identify the optimal condition and to assess the expected total neutron counts that can be obtained with the SINRD measurements. Results from Monte Carlo simulations showed that ranges between 0.1–0.3 mm and 0.5–1.0 mm ensure the optimal conditions for the Gd and Cd filters, respectively. Moreover, a 239Pu fission chamber is better suited to measure neutrons close to the 0.3 eV resonance and it has the highest sensitivity to 239 Pu, in comparison with a 235U fission chamber, with a 3He proportional counter, and with a 10B proportional counter. The use of a thin Gd filter and a thick Cd filter is suggested for the 239Pu and 235U fission chambers to increase the total counts achieved in a measurement, while a thick Gd filter and a thin Cd filter are envisaged for the 3He and 10B proportional counters to increase the sensitivity to 239Pu. We concluded that an optimization process that takes into account measurement time, filters thickness, and detector size is needed to develop a SINRD detector that can meet the requirement for an efficient verification of spent fuel assemblies. & 2015 Elsevier B.V. All rights reserved.
Keywords: SINRD Non-destructive assay Neutron resonance densitometry Spent fuel verification 239 Pu
1. Introduction Several non-destructive assays (NDA) are currently under research for the safeguards verification of spent fuel [1]. The spent fuel is becoming of primary concern because of the increasing quantity of material that is stored worldwide, and because the first geological repository is scheduled to become operational in the next decade [2]. The opening of the repository requires the development of a precise NDA method for the verification of a spent fuel assembly before its final disposal. As part of the international effort on the development of advanced NDA methods, the Belgian nuclear research center n
Corresponding author. Tel.: þ 32 14 33 28 44. E-mail addresses: [email protected] (R. Rossa), [email protected] (A. Borella), [email protected] (P.-E. Labeau), [email protected] (N. Pauly), [email protected] (K. van der Meer). http://dx.doi.org/10.1016/j.nima.2015.04.032 0168-9002/& 2015 Elsevier B.V. All rights reserved.
SCK CEN is carrying out the investigation of the Self-Interrogation Neutron Resonance Densitometry (SINRD, [3]). SINRD is a passive neutron technique that aims at directly quantifying the 239Pu mass in a fuel assembly by measuring the attenuation of the neutron flux in the energy region close to 0.3 eV, because 239Pu has a strong resonance in the total cross-section around that energy [4]. The neutron flux close to 0.3 eV is estimated with SINRD by calculating the difference between the total neutron counts measured with detectors covered with a foil of Gd or Cd. These nuclides were chosen because they exhibit a cutoff in the total neutron crosssection slightly below and above 0.3 eV, and for this reason they are also referred to as neutron filters in this paper. At Los Alamos National Laboratory the SINRD technique was investigated for the measurement of spent fuel under water by placing a set of fission chambers on one side of the fuel assembly to measure the spontaneous neutron emission [5–9]. However, this configuration relied on the accurate positioning of the
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detectors [10] and was more sensitive to the first outer fuel pin rows [5]. The approach proposed in [3] considered the fuel assembly in air surrounded by a thick slab of polyethylene to ensure neutron moderation, while the neutron detector is placed in the central guide tube of the assembly. This configuration ensures better condition for SINRD in comparison with the case of fuel stored under water, because the neutron moderation occurring within the fuel pins when the fuel assembly is stored under water washes out the reduction of the neutron flux around 0.3 eV used for the SINRD technique [3]. In addition, the dry conditions used as reference in this study can be representative of the encapsulation plant foreseen to be built in Sweden over the coming years [11]. The definition of the SINRD signature and the approach used in this study for its determination are first described in the paper. Then the optimization of the filters thicknesses used for the selection of the 239Pu resonance region is presented, together with the method for the estimation of the total neutron counts and sensitivity to 239Pu. Several combinations of filters and detector types are compared and recommendations are given at the end of the paper on the selection of the measurement approach for the SINRD technique.
2. Determination of the SINRD signature 2.1. Definition of the SINRD signature To quantify the attenuation of the neutron flux around the 0.3 eV energy region the so-called SINRD signature, defined in [3] as the ratio between the energy-integrated values of the neutron flux in the fast region (FAST, i.e. 0.1–20 MeV) and in the region close to the resonance (RES, i.e. 0.2–0.4 eV), is used. SINRD ¼
FAST RES
ðaÞ
As shown in [3], the SINRD signature is correlated to the 239Pu amount in the fuel assembly model. To take into account the characteristics of detector types and neutron filters, the SINRD signature is estimated in this paper as the ratio between the neutron counts in the fast and in the resonance energy regions. Therefore, the response of neutron detectors in the central guide tube of the fuel assembly was modelled; the fast neutron counts were determined for a bare 238U fission chamber, while a 239Pu fission chamber covered with filters was the chosen detector to estimate the neutron counts close to the 0.3 eV resonance.
is valid for a parallel neutron beam (φin) perpendicular to a slab of material. The neutron beam that traverses the sample without any interaction (φtr) is given by: ! X nk σ D φtr ¼ φin U exp ðbÞ tot;k k
where σ D tot;k is the Doppler broadened total cross-section and nk is the number of atoms per unit area of nuclide k. Therefore the transmitted flux is directly related to the total cross section and the areal density of the nuclides present in a sample [13]. The integral neutron counts in the resonance region (RES) per atom of active material in the detector and per emitted neutron were then determined as: Z RES ¼ σ det Uφin U ½expð nGd σ Gd Þ expð nCd σ Cd ÞdE ðcÞ where σdet is the cross-section of the active material in the detector, E is the energy of the incoming neutron, nGd and nCd are the number of atoms per unit area of the neutron filters, while σGd and σCd are their corresponding total cross-sections. The ideal detector for the estimation of the RES integral is a 239Pu-loaded fission chamber [4] and therefore the σdet considered for this case was the fission cross-section for 239Pu. The RES integral value was calculated over the energy region between 10 9 and 20 MeV, with the values for the cross-sections of detectors and filters determined as explained in [14]. This approach for the calculation requires only a Monte Carlo simulation to determine the energy distribution of the neutron flux in the guide tube without filters. Then, by using the transmission factors for several filter thicknesses the influence of the neutron absorbers was calculated without running additional simulations. However, the formulas used for these calculations are valid only in the case of a parallel neutron beam and a perpendicular thin neutron absorber. Therefore, the effect of the neutron filters will be re-evaluated in the future if the measurement conditions will significantly differ from these assumptions. The difference calculated in the RES integral is motivated by the fact that the Gd filters have a lower cutoff energy compared to the Cd filters; therefore, the transmitted flux using a Gd filter is generally higher than the value calculated through a Cd filter. However, an effect is visible in Fig. 1, which plots the transmission factors of Formula (c) in the case of 0.5 mm Gd, 0.6 mm Cd filters and their corresponding difference. Due to the combination of thick Gd and thin Cd filters significant negative contributions to the integral difference can be obtained in the epithermal energy 1.0
2.2. Calculation of the SINRD signature
0.8 0.6
Transmission factor
The SINRD signature was calculated in this study using a Monte Carlo model of a PWR 17 17 fuel assembly in air surrounded by a slab of 12 cm of polyethylene on all sides. The Monte Carlo code MCNPX v.2.7.0 [12] was used to obtain the energy distribution of the neutron flux in the central guide tube of the fuel assembly, by simulating 108 particles and calculating the so-called F4 flux tally in 600 logarithmically interpolated energy bins from 10 9 to 20 MeV. The neutron flux was determined in an empty cavity that can host the selected detector type and it was normalized per starting particle. The influence of the filters and the response of each specific detector were then estimated according to the approach explained in Section 3. The SINRD signature requires the neutron counts in two different energy regions and the selection of the region close to the 0.3 eV resonance is achieved by using specific neutron absorbers. In order to determine the impact of a neutron absorber on the neutron flux we used the following analytical approach, which
0.4 0.2 0.0 -0.2 -0.4 -0.6
Gd 0.5 mm Cd 0.6 mm Difference
-0.8 -1.0 -9
10
-8
10
-7
10
-6
10
-5
10
-4
10
-3
10
-2
10
-1
10
0
10
1
10
Neutron energy (MeV) Fig. 1. Negative contributions due to the difference between the transmitted fluxes through Gd and Cd filters. The values shown in the plot are the transmission factors calculated with Formula (b).
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region and this effect must be accounted for in the selection of the filters thickness. The total neutron counts in the fast energy region are the second quantity needed for the calculation of the SINRD signature. This value is calculated according to the following formula: Z FAST ¼ σ U Uφin dE ðdÞ where σU is the fission cross-section of 238U and φin is the neutron flux calculated in the guide tube of the fuel assembly.
95
time (t), while the uncertainty associated to this quantity (σGd, σCd, σF) is estimated as the square root of the total count. The uncertainty of the total count in the resonance region (σR) takes into account the uncertainty propagation for the total counts estimated for the detectors covered with the neutron filters. Therefore, the relative uncertainty of the SINRD signature (σS,rel) due to counting statistics can be calculated as " # σ2 σ 2R σ 2R cF t 1 1 cGd þ cCd ¼ þ ¼ þ σ 2S;rel ¼ F2 þ t cF ðcGd cCd Þ2 C F ðC Gd C Cd Þ2 c2F t 2 ðcGd cCd Þ2 t 2 ðfÞ
3. Optimization of the filters and detectors for the SINRD technique 3.1. Selection criteria for thicknesses of the neutron absorbers The ideal configuration for the measurement of the neutron flux close to the 0.3 eV resonance is to have the RES integral sensitive to neutrons with energy within the 0.2–0.4 eV region and to minimize the contributions from neutrons with different energy. Therefore, two criteria based on the energy distribution of the transmitted fluxes were chosen. The first criterion was the ratio between RES integrated over the 0.2–0.4 eV energy window and RES integrated over the whole energy range. The second criterion was the ratio between the integral value of the negative contributions to RES and the absolute integral value, defined as the absolute sum of positive and negative contributions. In order to select the optimal configuration for SINRD, the first criterion should be maximized and the second criterion should be minimized. The energy distribution of the neutron flux calculated in the guide tube of the fuel assembly varies with the fuel composition because of the neutron absorption of the nuclides contained in the fuel composition. Therefore, also the values of the criteria used for the filters optimization are influenced by the 239Pu content because of the significant neutron absorption in the resonance region. In order to evaluate this effect the criteria for the selection of the optimal filter thickness were calculated for a 239Pu fission chamber in case of fuel without fissile content and with 1% of 239 Pu. This 239Pu content was chosen as extreme to cover the 239Pu content usually present in the LEU spent fuel assemblies [15]. Detailed fuel compositions were not included at this stage in the simulations to avoid influence from other nuclides apart from 239 Pu. 3.2. Assessment of the detectors total counts and associated uncertainties Based on the approach of [16], the count rate of a neutron detector (c) was estimated with the following formula: Z c ¼ N tot U n U σ det U φin U exp nf il σ f il dE ðeÞ where Ntot is the total neutron emission from the spent fuel assembly, n is the mass of active material in the detector, nfil is the number of atoms per unit area of the neutron filter, and σfil is its corresponding total cross-section. Since no filter is used for the estimation of the fast neutron flux, in that case the exponential term of Formula (e) is not considered. By using this approach, the energy distribution of the incoming flux φin is estimated by means of Monte Carlo simulations accounting for the angular and spatial distribution of the neutron flux, instead of using leakage probabilities for different energy regions as described in [7]. The total count (CGd, CCd, CF) of each measurement is the product between the count rate (cGd, cCd, cF) and the measurement
The detectors considered for this estimate contained 100 mg of active material (i.e. 239Pu, 235U, or 10B), or 31.63 mg of 3He (corresponding to a tube with 1.27 cm diameter and 50 cm length filled with 4 atm gas). The fissile amount in the fission chambers is similar to the equipment used for the Fork detector [16], while the pressure value of the 3He tube was taken from [17]. The fission chamber for the estimation of the fast neutron flux contained 1000 mg of 238U, in order to compensate for the small fission cross-section of 238U. The amount of active materials in the detectors was the only parameter of the detectors that was used to calculate the total neutron counts, since the detectors were not modelled in the simulations. Future work will refine this analysis and consider dimensions of detectors that can actually be inserted in a PWR guide tube. The total neutron emission of the spent fuel was calculated from the results of the reference spent fuel library developed at SCK CEN [18] and considered fuel with 3.5% initial enrichment, 5 years of cooling time, and discharge burnup ranging from 5 to 60 GWd/tU. The measurement time was set to 60 min for all calculations.
4. Results with a
239
Pu fission chamber
4.1. Selection criteria for the filter thickness A plutonium-loaded fission chamber is considered as the ideal detector for SINRD in order to increase the detection of neutrons in the energy region close to 0.3 eV [4]. The selection criteria for the filters thicknesses were evaluated for this detector type in the case of fuel without 239Pu and with 1% 239Pu content. Tables 1 and 2 show the two selection criteria calculated for several filter combinations. The share of the resonance region reported in Table 1 increases with thick Gd and thin Cd filters, while the opposite trend is shown in Table 2 for the second criterion. The significant negative contributions calculated in Table 2 are also the reason of the values higher than 100% shown in Table 1. The uncertainties reported in the tables take into account only the statistical uncertainty of the Monte Carlo simulations. A significant difference is observed in the values of both criteria when comparing the results from Tables 3 and 4 (fuel with 1% 239Pu) with the cases shown in Tables 1 and 2 (fuel without 239Pu). The Table 1 First criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel without fissile material. The criterion is the ratio between the RES integrated in the 0.2–0.4 eV energy window and the RES integral over the complete energy spectrum. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
73.2 7 0.6 72.7 7 0.5 72.6 7 0.5 72.6 7 0.5 72.5 7 0.5 72.4 7 0.5
87.0 7 0.8 85.6 7 0.7 85.0 7 0.7 84.7 7 0.7 84.4 7 0.7 84.2 7 0.6
93.5 70.9 91.2 70.8 90.2 70.8 89.6 70.8 89.1 70.7 88.7 70.7
97.9 7 1.1 94.4 7 0.9 93.0 7 0.9 92.2 7 0.8 91.5 7 0.8 90.9 7 0.8
101.9 7 1.3 97.0 7 1.1 95.0 7 1.0 93.8 7 0.9 92.9 7 0.9 92.2 7 0.8
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Table 2 Second criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel without fissile material. The criterion is the share of negative contributions to the RES integral over the absolute RES integral value. Share NEG (%) Gd 0.1 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
1.3 7 0.4 0.4 7 0.4 0.17 0.4 o 0.17 0.4 o 0.17 0.4 o 0.17 0.4
Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm
2.3 7 0.5 0.9 7 0.5 0.4 7 0.5 0.3 7 0.5 0.2 7 0.5 0.2 7 0.5
3.5 7 0.6 1.5 7 0.6 0.8 7 0.6 0.6 7 0.6 0.5 7 0.5 0.5 7 0.5
4.9 7 0.7 2.2 7 0.7 1.4 7 0.6 1.17 0.6 1.0 7 0.6 0.9 7 0.6
6.7 7 0.7 3.2 7 0.7 2.17 0.7 1.7 7 0.6 1.5 7 0.6 1.4 7 0.6
Table 3 First criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel containing 1% of 239Pu. The definition of the criterion is the same of Table 1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm 1% 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
43.4 7 0.8 42.7 7 0.7 42.4 7 0.7 42.17 0.7 41.8 7 0.7 41.5 7 0.6
61.8 7 1.4 58.9 7 1.2 57.3 7 1.1 56.0 7 1.1 54.9 7 1.0 53.9 7 1.0
74.5 7 2.1 68.37 1.7 65.17 1.5 62.8 7 1.4 60.9 7 1.3 59.2 7 1.2
88.2 73.3 76.5 72.3 71.0 71.9 67.4 71.7 64.5 71.5 62.2 71.4
110.6 75.8 87.1 73.3 77.8 72.5 72.1 72.1 68.0 71.8 64.7 71.6
difference in the values for the first criteria is due to the neutron absorption of 239Pu in the 0.3 eV resonance region. The presence of 239 Pu in the fuel composition leads to a decrease in the RES integral value because of resonant absorption, which finally leads to lower ratios in Table 3. The large share of negative contributions calculated with the second criterion (Table 4) is also related to the reduction of the total neutron flux due to the absorptions. Moreover, the reduced RES integral value obtained by adding 239Pu in the fuel composition leads to a higher sensitivity of the criteria to the Cd thickness compared to the results for fuel without fissile content. 4.2. SINRD signature Another parameter for the selection of the filters thickness is the value of the SINRD signature that can be achieved by each combination. Table 5 shows in the case of a 239Pu fission chamber the SINRD signature for several fuel compositions and filter combinations. The values are normalized to the SINRD signature calculated without fissile material in the fuel composition. The uncertainties of the SINRD signature deriving from the statistical uncertainty of the Monte Carlo simulations are smaller than 0.6%. The SINRD signature increases with the fuel burnup because of the increased quantity of 239Pu, and it increases with the thickness of the Gd filter. On the other hand, the SINRD signature decreases by increasing the thickness of the Cd filter, therefore the highest values are obtained in the case with Gd filter of 0.3 mm and Cd filter of 0.5 mm. However, the final choice of the filters must also take into account the total neutron counts achieved in a measurement as calculated in the next section.
Table 4 Second criterion used for the selection of the filters thickness considering 239Pu as active material in the detector and fuel containing 1% of 239Pu. The definition of the criterion is the same of Table 2. Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm 1% 239Pu Cd Cd Cd Cd Cd Cd
2.4 7 0.8 1.0 7 0.8 0.6 7 0.8 0.5 7 0.7 0.4 7 0.7 0.3 7 0.7
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
6.2 7 1.2 3.4 7 1.1 2.5 7 1.1 2.17 1.0 1.9 7 1.0 1.7 7 1.0
11.1 71.5 6.9 71.4 5.3 71.3 4.5 71.2 4.1 71.1 3.7 71.1
17.2 7 1.8 11.4 7 1.6 9.17 1.5 7.8 7 1.3 7.0 7 1.3 6.4 7 1.2
Table 5 SINRD signature for several filter combinations in the case of a chamber.
25.17 2.1 17.2 7 1.9 13.8 7 1.6 11.9 7 1.5 10.6 7 1.4 9.7 7 1.3
239
Pu fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 5 10 15 20 40 60
2.12 2.93 3.52 3.94 4.52 4.55
2.07 2.84 3.38 3.76 4.29 4.32
2.28 3.27 4.00 4.52 5.27 5.30
2.18 3.06 3.68 4.12 4.74 4.77
2.37 3.48 4.30 4.91 5.77 5.82
2.22 3.14 3.80 4.26 4.91 4.94
Table 6 Estimation of the total counts and relative uncertainties in the case of a 238U fission chamber. Fuel BU (GWd/tU)
Counts
Unc. (%) 1
5.84 10 2.58 102 1.09 103 4.03 103 9.12 104 4.47 105
5 10 15 20 40 60
13.1 6.2 3.0 1.6 0.3 0.1
The total counts for different filter combinations were calculated with the Formula (e) given in Section 3.2 and the results are shown in Table 7. As for the fast neutron flux, for all filters combinations the counts significantly increase in the case of fuel with high burnup because of its high neutron emission. Moreover, the total counts also increase by decreasing the filter thickness due to the reduced neutron absorption by the material. The relative uncertainty associated to the SINRD signature was estimated for different filter combinations according to Section 3.2. The results are included in Table 8 and show decreasing values by increasing fuel burnup, or by combining thin Gd filters with thick Cd filters. The results shown in this section are in line with previous work [3], where a Gd thickness of 0.3 mm was identified to maximize the SINRD signature. However, considering that all cases shown in Table 5 have high sensitivity to the 239Pu content, the combination of 0.1 mm Gd filter and 1.0 mm Cd filter is suggested to decrease the measurement uncertainty.
235
U fission chamber
4.3. Detector total counts
5. Results with a
The fast neutron counts together with the relative uncertainty were estimated according to Section 3.2 and the results are shown in Table 6. The total counts increase with the discharge burnup because of the increasing neutron emission of the fuel assembly. The values shown in Table 6 are used for the estimation of the overall measurement uncertainty for all detectors considered in this paper.
5.1. Selection criteria for the filter thickness Although a plutonium-loaded fission chamber is considered as the ideal detector for SINRD, previous work from [5] mentioned the difficulty to obtain such detector type. Therefore a 235U fission chamber was considered in this paper as alternative. Following the
R. Rossa et al. / Nuclear Instruments and Methods in Physics Research A 791 (2015) 93–100
Table 7 Estimation of the total counts and relative uncertainties in the case of a Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
239
97
Pu fission chamber.
Gd 0.2 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
5.79 102 2.06 103 7.80 103 2.69 104 5.62 105 2.74 106
4.2 2.2 1.1 0.6 0.1 0.1
4.38 102 1.58 103 6.08 103 2.11 104 4.46 105 2.18 106
4.8 2.5 1.3 0.7 0.1 0.1
3.72 102 1.37 103 5.32 103 1.86 104 3.97 105 1.93 106
5.2 2.7 1.4 0.7 0.2 0.1
2.18 102 9.06 102 3.72 103 1.35 104 2.98 105 1.46 106
6.8 3.3 1.6 0.9 0.2 0.1
1.75 102 7.53 102 3.14 103 1.15 104 2.57 105 1.26 106
7.6 3.6 1.8 0.9 0.2 0.1
Table 8 Estimation of the SINRD signature relative uncertainty in the case of a 239Pu fission chamber.
Table 10 Second criterion for the filter optimization considering 235U as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1.
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm
Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu
5 10 15 20 40 60
15.2% 7.8% 4.0% 2.2% 0.5% 0.2%
14.7% 7.4% 3.8% 2.0% 0.4% 0.2%
17.5% 9.7% 5.2% 2.9% 0.7% 0.3%
16.1% 8.5% 4.5% 2.5% 0.6% 0.3%
20.5% 12.0% 6.7% 3.8% 0.9% 0.4%
17.7% 9.7% 5.2% 2.9% 0.7% 0.3%
Table 9 First criterion for the filter optimization considering 235U as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
42.7 7 0.3 41.17 0.2 40.6 7 0.2 40.4 7 0.2 40.3 7 0.2 40.17 0.2
70.57 0.7 63.7 7 0.6 61.5 7 0.5 60.3 7 0.5 59.5 7 0.5 58.7 7 0.4
92.9 7 1.4 77.6 7 0.9 72.9 7 0.8 70.57 0.7 68.87 0.7 67.4 7 0.7
119.9 7 2.8 89.17 1.5 80.9 7 1.2 77.0 7 1.0 74.4 7 0.9 72.2 7 0.9
171.6 7 7.2 103.0 7 2.4 89.0 7 1.7 82.9 7 1.4 78.9 7 1.3 75.8 7 1.1
approach described in Section 2.2, the 235U fission cross-section was used to estimate the detector response. In order to compare the performance of different detectors, the same criteria used for the filter optimization were calculated for the filter thicknesses considered with the 239Pu fission chamber and the results are summarized in Tables 9 and 10. As a general trend the values of both criteria are inferior for the 235U fission chamber compared to the results shown in Section 4.1. 5.2. SINRD signature Table 11 shows the SINRD signature calculated in the case of a U fission chamber, and the values are normalized to the SINRD signature obtained without fissile material in the fuel composition. We observe the same trends of the signature with fuel burnup and filters thickness described in Section 4.2; however, there is a significant decrease in the values compared to the results with the 239Pu fission chamber. The uncertainties of the SINRD signature taking into account the statistics of the Monte Carlo simulations are smaller than 0.5%. 235
5.3. Detector total counts The performance of a 235U fission chamber was estimated also in terms of the total counts that can be achieved with this detector type (Table 12). The approach described in Section 3.2 was used for
Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
5.2 7 0.3 1.7 7 0.3 0.6 7 0.3 0.3 7 0.3 0.2 7 0.3 0.1 70.3
11.7 7 0.5 4.6 7 0.6 2.2 7 0.6 1.4 7 0.6 1.1 70.5 1.0 7 0.5
18.4 70.8 8.2 70.8 4.7 70.8 3.4 7 0.7 2.8 70.7 2.5 70.7
25.7 7 1.1 12.8 7 1.0 8.0 7 1.0 6.17 0.9 5.3 7 0.9 4.8 7 0.8
Table 11 SINRD signature for several filter combinations in the case of a chamber.
34.17 1.5 18.5 7 1.1 12.3 7 1.2 9.8 7 1.1 8.5 7 1.1 7.7 7 1.0
235
U fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 5 10 15 20 40 60
1.77 2.25 2.58 2.82 3.14 3.15
1.71 2.15 2.44 2.64 2.91 2.92
1.93 2.54 2.95 3.24 3.64 3.65
1.77 2.23 2.53 2.73 3.00 3.01
2.12 2.90 3.45 3.84 4.39 4.41
1.79 2.26 2.56 2.76 3.02 3.03
this calculation. The total counts are significantly lower than in the case of the 239Pu fission chamber, because of the lower fission cross-section of 235U compared to 239Pu. This leads also to an increase in the measurement uncertainty (Table 13). In the case of 235U fission chamber, the use of thin Gd filters is suggested in order to decrease the measurement uncertainty. Moreover, by choosing a thin Gd filter there is a reduction of the negative contributions to the absolute RES integral value (Table 10), while keeping still significant contributions from the resonance region (Table 9). Considering the low total counts shown for all cases in Table 12, the selection of 0.1 mm Gd filter and 1.0 mm Cd filter is suggested to increase the total counts while keeping a significant sensitivity to 239Pu.
6. Results with proportional counters 6.1. Selection criteria for the filter thickness Other detector types simulated for the SINRD technique were two proportional counters. As for the detector types considered before, the approach explained in Section 2.2 was used to determine the detector response, and either the absorption cross-section of 3He or the (n,α) reaction cross-section of 10B were chosen for the calculation. However, to limit the length of the paper, only results
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Table 12 Estimation of the total counts and relative uncertainties in the case of a Filter Fuel BU GWd/tU 5 10 15 20 40 60
Gd 0.1 mm
235
U fission chamber.
Gd 0.2 mm
Gd 0.3 mm
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (%)
2.30 102 9.34 102 3.79 103 1.36 104 3.00 105 1.47 106
6.6 3.3 1.6 0.9 0.2 0.1
1.86 102 7.80 102 3.22 103 1.17 104 2.60 105 1.27 106
7.3 3.6 1.8 0.9 0.2 0.1
1.72 102 7.28 102 3.02 103 1.10 104 2.47 105 1.21 106
7.6 3.7 1.8 1.0 0.2 0.1
1.55 102 6.75 102 2.84 103 1.04 104 2.34 105 1.15 106
8.0 3.8 1.9 1.0 0.2 0.1
1.41 102 6.22 102 2.63 103 9.67 103 2.18 105 1.07 106
8.4 4.0 2.0 1.0 0.2 0.1
235
U fission
Filter Gd 0.1 mm Gd 0.1 mm Gd 0.2 mm Gd 0.2 mm Gd 0.3 mm Gd 0.3 mm Fuel BU (GWd/tU) Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm Cd 0.5 mm Cd 1.0 mm 29.4% 16.7% 9.1% 5.1% 1.2% 0.5%
25.4% 14.1% 7.5% 4.2% 0.9% 0.4%
60.9% 37.0% 20.7% 11.8% 2.7% 1.2%
42.2% 24.5% 13.3% 7.4% 1.7% 0.8%
111.4% 71.4% 40.9% 23.7% 5.6% 2.6%
59.8% 35.1% 19.1% 10.7% 2.4% 1.1%
Table 14 First criterion for the filter optimization considering 3He as active material in the detector and fuel without fissile material. The criterion is defined in Section 3.1. Share RES (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5 mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
Cd 1.0 mm
Counts
Table 13 Estimation of the SINRD signature relative uncertainty in the case of a chamber.
5 10 15 20 40 60
Cd 0.5 mm
38.4 7 0.2 37.2 7 0.2 36.8 7 0.2 36.6 7 0.2 36.4 7 0.2 36.3 7 0.2
65.2 7 0.6 59.2 7 0.5 57.17 0.4 55.9 7 0.4 55.0 7 0.4 54.3 7 0.4
88.6 7 1.3 73.9 7 0.9 69.2 7 0.7 66.77 0.7 64.97 0.6 63.4 7 0.6
119.5 7 2.8 87.3 7 1.4 78.6 7 1.1 74.2 7 1.0 71.2 7 0.9 68.87 0.8
188.4 78.8 105.0 72.5 88.8 71.7 81.6 71.4 76.9 71.2 73.3 71.1
with 3He as active material are shown, since the performance of both detectors were almost identical but the total counts for 3He were higher than the values for 10B. Both detector types are not traditionally used for spent fuel measurements because of the sensitivity to γ radiation. Therefore the results contained in this section are meant only to offer an optimal performance of the detectors, while more detailed considerations on the effective applicability to field measurements are not tackled. The results obtained with the 3He detector confirm the reduction of the performance of other detector types compared to the 239 Pu fission chamber. In fact, the values of the criteria obtained with the 3He proportional counter (Tables 14 and 15) are similar to the case of 235U fission chamber (Tables 9 and 10). 6.2. SINRD signature The values of the SINRD signature for a 3He proportional counter are reported in Table 16 and show similar results to the case of a 235U fission chamber. As for the other detectors, the values are normalized to the case of fuel without fissile content. The statistical uncertainties of the Monte Carlo simulations are lower than 0.5%. 6.3. Detector total counts Following the approach described in Section 3.2, the total counts achievable with a 3He proportional counter were
Table 15 Second criterion for the filter optimization considering 3He as active material in the detector and fuel withot fissile material. The criterion is defined in Section 3.1. Share NEG (%) Gd 0.1 mm Gd 0.2 mm Gd 0.3 mm Gd 0.4 mm Gd 0.5mm No 239Pu Cd Cd Cd Cd Cd Cd
0.5 mm 0.6 mm 0.7 mm 0.8 mm 0.9 mm 1.0 mm
4.9 7 0.3 1.8 7 0.3 0.8 7 0.3 0.5 7 0.3 0.4 7 0.3 0.4 7 0.3
11.8 7 0.5 5.2 7 0.5 3.0 7 0.5 2.3 7 0.5 2.0 7 0.5 1.8 7 0.5
19.17 0.8 9.6 7 0.7 6.2 7 0.7 5.0 7 0.7 4.4 7 0.7 4.17 0.7
27.3 7 1.0 15.0 7 0.9 10.4 7 0.9 8.5 7 0.9 7.6 7 0.8 7.17 0.8
36.5 7 1.9 21.7 7 1.1 15.6 7 1.1 13.0 7 1.0 11.6 7 1.0 10.7 7 0.9
estimated. By looking at Table 17, the advantage of using 3He as active material is visible on the increase of the total counts in the measurements compared to the previous detector types. Given the large cross-section of 3He, this detector type offers the best results in terms of measurement uncertainty (Table 18). The total counts obtained for a 10B proportional counter are lower than the case with the 3He proportional counter, but significantly higher than the cases with the fission chambers. This result is due to the fact that the (n,α) cross-section for 10B is slightly smaller than the absorption cross-section of 3He. Considering the low measurement uncertainty achieved by both proportional counters, a thick Gd filter (e.g. 0.3 mm) and a thin Cd filter (e.g. 0.5 mm) are proposed to increase the sensitivity to 239 Pu.
7. Discussion of the results The results presented in this work show the need for an optimization process taking into account total neutron counts, filters thickness, and detector size to develop a SINRD detector that can meet the requirements for an efficient verification of spent fuel assemblies. In order to increase the total counts of the measurements, several methods can be used. A straightforward approach is to increase the measurement time, but this is in contrast with the goal of obtaining an efficient detector that can be used during an inspection. An alternative solution would be to increase the size of the detector, but this is limited by the reduced size of the guide tube that hosts the instrument. Another method would be to select thin Gd filters and thick Cd filters, but this also is constrained by the considerations about the optimal range for the filters thickness. Therefore, for the 239Pu and 235U fission chambers the use of a thin Gd filter and a thick Cd filter (e.g. 0.1 mm and 1.0 mm respectively) is suggested in order to increase the total counts. On the contrary, taking advantage of the high neutron sensitivity of the 3He and 10B proportional counters, the combination of a thick Gd filter and a thin Cd filter (e.g. 0.3 mm and 0.5 mm respectively) is proposed to maximize the SINRD signature.
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Table 16 SINRD signature for several filter combinations in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU)
Gd 0.1 mm
Gd 0.1 mm
Gd 0.2 mm
Gd 0.2 mm
Gd 0.3 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
1.72 2.18 2.50 2.72 3.02 3.04
1.67 2.08 2.35 2.54 2.80 2.81
1.87 2.43 2.82 3.09 3.47 3.48
1.72 2.14 2.42 2.61 2.86 2.87
2.06 2.81 3.35 3.73 4.28 4.31
1.74 2.17 2.45 2.63 2.88 2.90
5 10 15 20 40 60
Table 17 Estimation of the total counts and relative uncertainties in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
Gd 0.2 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Counts
Unc. (%)
Counts
Unc. (%)
Counts
Unc. (I%)
Counts
Unc. (%)
Counts
Unc. (%)
4.66 104 1.85 105 7.41 105 2.64 106 5.74 107 2.81 108
0.5 0.2 0.1 0.1 o0.1 o0.1
3.49 104 1.43 105 5.85 105 2.11 106 4.46 107 2.28 108
0.5 0.3 0.1 0.1 o 0.1 o 0.1
3.09 104 1.29 105 5.32 105 1.93 106 4.28 107 2.10 108
0.6 0.3 0.1 0.1 o0.1 o0.1
2.68 104 1.16 105 4.85 105 1.78 106 3.98 107 1.95 108
0.6 0.3 0.1 0.1 o0.1 o0.1
2.32 104 1.02 105 4.30 105 1.58 106 3.56 107 1.74 108
0.7 0.3 0.2 0.1 o 0.1 o 0.1
Table 18 Estimation of the SINRD signature relative uncertainty in the case of a 3He proportional counter. Filter Fuel BU (GWd/tU) 5 10 15 20 40 60
Gd 0.1 mm
Gd 0.1 mm
Gd 0.2 mm
Gd 0.2 mm
Gd 0.3 mm
Gd 0.3 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
Cd 0.5 mm
Cd 1.0 mm
13.2% 6.3% 3.1% 1.6% 0.3% 0.2%
13.1% 6.3% 3.0% 1.6% 0.3% 0.2%
13.4% 6.5% 3.2% 1.7% 0.4% 0.2%
13.2% 6.3% 3.1% 1.6% 0.3% 0.2%
14.4% 7.3% 3.7% 2.0% 0.4% 0.2%
13.4% 6.5% 3.2% 1.7% 0.4% 0.2%
Each detector type presented in this paper shows both positive and negative characteristics. The 239Pu fission chamber ensures the best sensitivity to 239Pu but its use is hindered by its unavailability on the market. On the other hand, the 235U fission chamber shows low SINRD signature but it benefits from extensive experience for spent fuel verification. Similar SINRD signatures are obtained with the proportional counters together with very high neutron sensitivity, but the influence of gamma radiation on these detectors must be evaluated. Therefore the development of the SINRD technique should take into account the results obtained from the optimization of the filter thicknesses, and it should be supported by considerations on detectors design, measurement time, and sensitivity to gamma radiation.
8. Conclusions The performance of several detector types for the SINRD technique was compared in this paper by calculating the SINRD signature and detector total counts for different filter thicknesses. In the case of the 239Pu fission chamber the influence of the neutron filters was assessed both in the case of fuel without fissile content and for fuel with high 239Pu concentration, and a general degradation of the results was observed in the second case due to the presence of 239Pu. Significant effort was devoted to the comparison of the 239Pu fission chamber with other technologies. By looking at the
selection criteria used for the definition of the optimal filter thickness, it was concluded that the alternative detectors do not ensure the same performance of the 239Pu fission chamber. This is also reflected in the sensitivity of each detector to the 239Pu content; the total 239Pu was correlated with the SINRD signature and the 239Pu fission chamber showed significantly better results. For all detector types it was found that the SINRD signature increases with the fuel burnup and with the Gd thickness, while it decreases with the Cd thickness. The comparison among the detectors also included an estimation of the total counts that can be achieved with each detector, and the 3He and 10B proportional counters showed the largest total counts because of their large neutron sensitivity. Therefore, by considering the SINRD signature and total neutron counts achievable for each detector type, thin Gd and thick Cd filters are suggested in case of 239Pu and 239U fission chambers to maximize the total neutron counts, while thick Gd and thin Cd filters are advised for 3He and 10B proportional counters to maximize the SINRD signature.
Acknowledgements This work is sponsored by GDF SUEZ in the framework of the cooperation agreement CO-90-07-2124 between SCK CEN and GDF SUEZ.
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